Since 18 of December 2019 conferences.iaea.org uses Nucleus credentials. Visit our help pages for information on how to Register and Sign-in using Nucleus.

# International Conference on Fast Reactors and Related Fuel Cycles FR22: Sustainable Clean Energy for the Future (CN-291)

Europe/Vienna
Vienna, Austria

#### Vienna, Austria

Description

The International Atomic Energy Agency (IAEA) once again brings together the fast reactor and related fuel cycle community by organizing the International Conference on Fast Reactors and Related Fuel Cycles: Sustainable Clean Energy for the Future (FR22). The IAEA welcomes and encourages the participation of women, early career professionals and individuals from developing countries.

### FR22 in Vienna 19-22 April 2022, Starts at 9:30

Designated participants register here: InTouch+  (online access)

Observers register here and access online

Full Conference Programme here: Programme

## Attend online streaming here

### Due to COVID-19 crisis the IAEA, in agreement with the host country, the International Conference on Fast Reactors and Related Fuel Cycles will be held in Vienna from 19 to 22 April 2022 with a virtual attendance also possible.

To take advantage of the additional time to prepare the conference, the full paper submission deadline will be also extended to 20 January 2022. All submitted manuscripts will go through the peer-review process organized by the FR22 International Scientific Programme Committee (ISPC).

The accepted papers will be available online during the Conference and will be included in FR22 Proceedings that the IAEA plans to publish shortly after the Conference.

Purpose and Objective

The conference is aimed at providing a forum to exchange information on national and international programmes, and more generally new developments and experiences, in the field of fast reactors and related fuel and fuel cycle technologies. The conference will have the following general objectives:

• To identify and discuss strategic and technical options, including potential capabilities in mitigation of the climate change while reducing the burden of generated high level waste.
• To promote the development of fast reactors and related fuel cycle technologies to enhance nuclear energy development in a safe, proliferation resistant, environmentally friendly and cost-effective manner.
• To identify research and development gaps and key issues that need to be addressed in relation to the industrial deployment of these nuclear power technologies.
• To engage young scientists and engineers in this field, in particular with regards to the development of innovative fast reactor concepts

History

The first International Conference on Fast Reactors and Related Fuel Cycles (FR09) was held in Kyoto, Japan, in 2009 and was subtitled “Challenges and Opportunities”. The second conference (FR13) was held in Paris, France, in 2013 with the theme “Safe Technologies and Sustainable Scenarios”. The third conference (FR17) was held in Yekaterinburg, Russian Federation, in 2017, celebrating the sodium cooled fast reactor BN-800's connection to the grid in December 2015 at the Beloyarsk nuclear power plant (NPP).

The nuclear industry has from its inception recognized the important role of fast reactors and related fuel cycles in ensuring the long term sustainability of nuclear power. Fast reactors operated in a closed fuel cycle help to improve the utilization of resources — both fissile and fertile materials — used in nuclear fuels. This improvement is possible because fast reactors can breed fissile materials and, using modern fuel cycle technologies, recycle materials bred in these reactors. In this way, fast reactors and related fuel cycle technologies can make an enormous contribution to the sustainability of nuclear energy production. They have the potential to produce a hundred times more energy from natural uranium resources. At the same time, fast neutrons favour fission of heavy atoms, instead of capture, so they can also be used to transmute minor actinides, thereby reducing the demands on geological repositories for the final disposal of nuclear waste. Many countries are actively developing reactor, coolant, fuel and fuel cycle technologies. Reactor technologies under development include sodium- , lead- , gas-, molten salt- and even supercritical water-cooled systems and technologies and accelerator-driven systems. In parallel, several demonstration projects, ranging from small to large scale, are under study or construction. For such nuclear energy systems to become viable for industrial deployment in the coming decades, designers will have to increase their level of safety in order to gain public acceptance. Harmonization of safety standards at the international level could play a leading role in achieving these goals.

Contact
• Tuesday, April 19
• Opening Session

Chair: Ms. Aline des Cloizeaux

• 1
Speaker: Mr Rafael Grossi
• 2
• 3
Speaker: Ms Amparo Gonzalez-Espartero (Scientific Secretary)
• Coffee Break
• Plenary 1. Keynotes from Member States

Chair: Ms Aline des Cloizeaux

• 4
China Key Note
Speaker: Mr Hongyi Yang
• 5
France Key Note
Speaker: Mr Frederic Serre
• 6
India Key Note
Speaker: Mr Balasubramaniam Venkataraman
• 7
Japan Key Note
Speaker: Mr Hideki Kamide
• 8
Republic of Korea Key Note
Speaker: Mr Lim Chae Young
• 9
Russian Federation Key Note
Speaker: Mr Vyacheslav Pershukov
• 10
United States Key Note
Speaker: Ms Alice Caponiti
• Lunch Break
• 1.1 Overviews and Fundamentals of Fast Reactors

Chairs: Didier De Bruyn and Bob Hill

• 11
DEVELOPMENT OF BN REACTOR TECHNOLOGY IN RUSSIA
Speaker: Sergey Shepelev (JSC «Afrikantov OKBM»)
• 12
Progress in the Design and R&D for future FBRs
Speaker: RAGHUPATHY S. (Indira Gandhi Centre for Atomic Research, Kalpakkam)
• 13
Overview of U.S. Fast Reactor Technology R&D Program
Speaker: Robert Hill (Argonne National Laboratory)
• 14
Progress in conceptual design of a pool-type sodium-cooled fast reactor in Japan
Speaker: Mr Atsushi Kato (JAEA)
• 15
Status of Generation-IV Lead Fast Reactor Activities
Speaker: Mr Alessandro Alemberti (Ansaldo Nucleare)
• 16
The Status of the ALFRED Project
Speaker: Mr Marco Caramello (Ansaldo Nucleare)
• 17
MYRRHA, the Belgian prototype that fascinates the world
Speaker: Dr Didier De Bruyn (SCK CEN)
• 18
GFR Research and Development Programme in V4 countries
Speaker: Mr Petr Vácha (UJV Řež, a.s.)
• 4.1 Advanced Reactor Cladding and Core Material, Coolants, and Related Chemistry

Chairs: Christian Latgé and Ananthasivan K.

• 19
Creep and Tensile Properties of Indian Advanced Fast Reactor Clad tubes (IFAC-1) for Future FBRs
Speaker: Dr Prasad Reddy G. V. (Head, Creep Studies Section & Assistant Prof., HBNI )
• 20
Thermally conductive liquid-metal sublayer in fuel element
Speaker: Michael Orlov (Private institution «Innovation and technology center for the «PRORYV» project»)
• 21
FABRICATION AND PERFORMANCE ASSESMENT OF ODS FECRAL CLADDING TUBE
Speaker: Prof. Shi Liu (Institute of Metal Research, China Academy of Sciences)
• 22
Gear Test Assembly: First Liquid Metal Component Testing in METL
Speaker: Edward Kent (Argonne National Laboratory)
• 23
Influence of Low Dose Irradiation on Permanent Core Structural Materials of PFBR
Speaker: Mr Ran Vijay Kumar (Indira Gandhi Centre for Atomic Research)
• 24
Sodium coolant: interaction with its environment and coolant processing
Speaker: Dr Christian LATGE (CEA )
• 25
Investigation of sodium purification
Speaker: Dr Viktor Alekseev (SSC IPPE)
• 26
Development and Demonstration of Diffusion-type Hydrogen Meters for Sodium-cooled Fast Reactors
Speaker: Dr Hual-Te Chien (Argonne National Laboratory)
• Poster Session
• 27
CFD ANALYSES OF THE ALFRED HOT PLENUM
Speaker: Dirk Visser (NRG)
• 28
DESIGN & DEVELOPMENT OF CUSTOM SHAPED BACK-UP SEAL IN SILICONE FOR PFBR
Speaker: Sriramachandra Aithal (Indira Gandhi Centre for Atomic Research, Kalpakkam)
• 29
Development of Burnup Analysis System for rotational and Spiral Fuel Shuffling scheme in Breed-and-Burn Fast Rectors
Speaker: Dr Khanh Hoang Van (aInstitute for Nuclear Science and Technology, Vietnam Atomic Energy Institute)
• 30
Hybrid high power fast breeder reactor with metallic fuel and additives consisting with lightweight atoms
Speaker: IURII Drobyshev (Russian Federation)
• 31
INDUSTRIAL ENERGY COMPLEX WITH FAST NEUTRON REACTOR
Speaker: Andrey Petrenko
• 32
Investigation of characteristics of fast power reactor with an additional function of large-scale production of plutonium-238

A dual-purpose fast reactor with a light / heavy liquid metal coolant and a large-scale production of Pu-238 is considered. A universal target complex for large-scale production of Pu-238 is located inside the reactor. Np-237 is considered as a starting material for the production of Pu-238. The target complex has a heterogeneous structure, including Np-237 and a moderator with a high atomic weight. In the target complex, an area is formed with favorable conditions for the production of plutonium-238 of the required isotopic purity:
- High neutron flux density for intensive irradiation of the starting material;
- Resonance spectrum of neutrons, to intensify the radiative capture on Np-237 and avoid its useless fission, as well as reduce the radiative capture / fission of plutonium-238;
- Use of a moderator with a high atomic weight and low absorption of neutrons around the Np-237 target: Pb, Bi, Pb-Bi eutectic, radiogenic lead, Pb-208. Computational studies have shown that when using such moderators, it becomes possible to form a vast region with the required resonance neutron spectrum. This circumstance determines the large scale of the production of Pu-238;
- Slow neutron absorbers in the area of Pu-238 production, such as cadmium-113 and gadolinium-157. This will make it possible to weaken the radiative capture of neutrons and fission of plutonium-238, i.e. increase its isotopic purity and reduce burnup;
- High temperature of the target material. This increases the width of the resonances of radiative capture of neutrons by neptunium-237, i.e. increases the rate of plutonium-238 production.
Placing the target complex in a fast reactor opens up the possibility of increasing the nuclear safety of the reactor:
- in the center of the core - the positive void effect of reactivity decreases;
- above the core - the ability to remotely change the reactivity of the reactor without inserting the rods into the core.
The use of Np-237 for the production of Pu-238 serves also to reduce the risk of its unauthorized use, because it is a well-fissionable isotope in the fast neutron spectrum with practically no spontaneous fission.

Speaker: Dr Evgeny Kulikov (National Research Nuclear University MEPhI)
• 33
LARGE-SCALE HYDROGEN PRODUCTION; Fast-neutron Reactors Coupled to Thermochemical Copper-Chlorine Hydrogen Plant

Nuclear energy utilization for hydrogen production is experiencing a growing momentum worldwide, associated with the unprecedented interest in building large-scale hydrogen production plants to support national and international decarbonization and climate change mitigation plans. Thermochemical water splitting cycles coupled to nuclear power plants are one of the sustainable solutions to replace the conventional polluting methane steam reforming and gasification processes of hydrogen production. Thermochemical hydrogen production cycles have been considered as an economic pathway for hydrogen production since 1960s. The paper discusses the use of fast reactor Small Modular Reactors coupled to the Copper-Chlorine thermochemical cycle for hydrogen production. This cycle is being developed at Ontario Tech University and the Canadian Nuclear Laboratories. It is also being investigated by several other countries.

Speaker: Rami El-Emam (Ontario Tech University)
• 34
Leak-Before-Break Design of Double-Walled Once-Through Steam Generators for Lead Cooled Fast Reactor

Among the various types of liquid metal cooled fast reactor, lead cooled fast reactor (LFR) can take a closed fuel cycle to manage fertile fuel and actinide efficiently. LFR can adapt an External Boiling Bayonet Steam Generator (EBBSG) system instead of conventional once-through high pressure steam generator. LFR with the EBBSG has advantages in terms of thermal efficiency and safety during operation. The EBSSG designs have proven extensive operating experiences in submarine reactors of the Former Soviet Union. However, their complex tube geometries invite life-limiting disadvantages ranging from vibration-led fatigue, oxide particles sedimentation and increased resistance to natural circulation under accident conditions.
The EBBSG system has thermal vibration issues at the nozzle because of a huge heat exchanger inside once-through steam boiler which consists of additional steam blowers. Furthermore, this vibration can lower the heat transfer efficiency of heat exchanger due to the adhesion phenomenon of impurities from lead coolant. And then, fatigue failure as well as environmental corrosion may occur at the shell side of heat exchanger, which leads the coolant leakage accident. To assure the overall performance of steam generator, those structural material degradation issues on the thermal vibration should be solved.
In this paper, we present innovative design concepts for durable, maintainable and accident-tolerant double-walled once-through steam generators (DWOTSG) that can fulfill requirements for non-refueling and hermetically-sealed 40 years life micro-modular LFR: Micro-Uranus. The newly designed tube for steam generator is consisted of double walled layers. In order to assure Leak-Before-Break (LBB), the inter-tubular gap is filled with engineered materials for leak detection and conductivity enhancement. Tube materials are also designed to assure prolonged corrosion resistance and oxide deposition control on heavy-liquid metal side. Exceptional corrosion resistance over 40 year-life can be delivered by Functionally Graded Composite (FGC) tubes with substrate made of internationally certified fast reactor fuel cladding materials. Oxide deposition and thermal degradation can be controlled by advanced flow channel structure and maintenance designs. The LBB characteristics, corrosion resistance, thermal-hydraulic performance of the innovative designs will be presented using fracture mechanics, oxidation and heat transfer models.

Speaker: Taeyong Kim (UNIST)
• 35
New Finite Element Neutron Kinetics Code System FENNECS/ATHLET for Coupled Safety Assessment of (Very) Small and Micro Reactors

In recent years, interest in (very) small and medium size reactor– (v)SMR – concepts for specific purposes has grown. They are characterized by irregular geometries. Though Monte Carlo methods for safety assessment are becoming more and more standard for steady state simulations, they are not yet mature enough for transient applications. To perform future coupled transient safety assessments of (v)SMRs, the Finite ElemeNt NEutroniCS (FENNECS) code is being developed at GRS. It solves the time-dependent and steady-state 3-d few-group diffusion equation in Galerkin finite element representation using upright triangular prisms with linear basis functions as spatial elements. FENNECS is also coupled with the GRS thermal-hydraulic system code ATHLET. For spatial meshing, FENNECS includes an internal meshing module for regular Cartesian and hexagonal lattices. It also allows to read node lists and element connectivities from external ASCII files. For the meshing of the irregular geometries of (v)SMRs, an external meshing tool is being developed as a Python software module which generates the data required by FENNECS. This paper considers the Heat Pipe Micro Reactor core proposed in the publicly available document “Multi-Physics Simulations of Heat Pipe Micro Reactor” (ANL/NSE-19/25). The specification provides a detailed description of all geometries. The core layout is composed of a hexagonal fuel pin lattice surrounded by six control rod drums. However, no mass density of any material is given and must be gathered from opened literature. Two Monte Carlo Serpent models are developed for comparing with the results presented in the benchmark and for the generation of homogenized few-groups cross-sections to be used in FENNECS. Two configurations are considered, one where the absorber face of all control drums are turned out of the core (named All Rods Out – ARO – configuration) and the other one where all absorber faces are turned in (All Rods In – ARI – configuration). The multiplication factors obtained in these two configurations are in good agreement with the benchmark. The challenge of modelling this reactor in FENNECS is to properly mesh the circular shape of control drums into a hexagonal lattice. They are approximated by a polygon, either with 12 or 24 edges, which conserves the area of the circle. The preliminary FENNECS results on the multiplication factor are in good agreements with Serpent reference results for both ARO and ARI configurations. This demonstrates the applicability of FENNECS, the features of which will be presented in this paper, to (v)SMRs.

Speaker: Jeremy Bousquet (GRS gGmbH)
• 36
On substantial slowing down of the kinetics of fast transient processes in fast reactor

The neutron lifetime is an important parameter of the reactor kinetics. When the inserted reactivity is more than the effective fraction of delayed neutrons, the reactor kinetics becomes very rapid. The fast reactor kinetics can be slowed down by increasing the neutron lifetime. The possibility of using lead isotope 208Pb as a neutron reflector with specific properties in the lead-cooled fast reactor is considered. A point kinetics model has been chosen to assess the emerging effects. The model takes into account the effects produced by neutrons returning from 208Pb-reflector to the reactor core.
Such specific properties of 208Pb as large atomic weight, weak neutron absorption allow neutrons from the reactor core to penetrate deeply into 208Pb-reflector, slow down there and have a noticeable probability to return to the reactor core and affect the chain fission reaction. The neutrons coming back from 208Pb-reflector have a long “dead-time” which represents the sum of times when neutrons leave the reactor core entering 208Pb-reflector and then diffuse back into the reactor core. During the “dead-time” these neutrons can’t affect the chain fission reaction. The neutrons returning from deep layers of 208Pb-reflector are close to the delayed neutrons in the terms of time delay. Moreover, the number of the neutrons coming back from 208Pb-reflector considerably exceeds the number of the delayed neutrons.
As a result, the neutron lifetime formed by the prompt neutron lifetime and the “dead-time” of the neutrons from 208Pb-reflector can be substantially increased. This can lead to the longer reactor period, which mitigates the effects of prompt super-criticality. To conclude, the use of lead isotope 208Pb as a neutron reflector can improve significantly safety of the fast reactor operation.

Speaker: Dr Evgeny Kulikov (National Research Nuclear University MEPhI)
• 37
Simple Design Comparison of uranium nitride pin cell assembly and matrix fuel assembly for a Lithium Cooled Fast Reactor

This report presents the results of a numerical simulation of thermal hydraulics processes in a liquid lithium cooled fast reactor core, for a Thermal Propulsion Engine, combined with a simple neutron population computing for an infinite cell lattice. Two types of fuel configuration were studied: pin cell assembly and matrix fuel assembly, with all requirements regarding safety conditions observed. Temperature distributions along the cooling channel and distributions in the radial direction were prepared, in the search of the location of the fuel maximum temperature, then criticality calculations were performed for highly enriched Uranium Nitride fuel using Serpent code.

Speaker: Raul Pineda (IAEA FELLOW)
• 38
The "ALFRED White Book": a business card of the project

The European-wide research on the Lead-cooled Fast Reactor (LFR) has been steadily advancing the technology readiness level at a point where, for further targeting industrial maturity, the need of a demonstrator has risen. The Advanced Lead-cooled Fast Reactor European Demonstrator (ALFRED), along with its support research infrastructure, is the missing link of the innovation chain being conceived for demonstrating the viability of the LFR technology propelling it beyond the research frontier which has been hitherto.
Given the elevated level of maturity the ALFRED Project has currently reached, and foreseeing its inclusion in the Romanian strategy for competitiveness as a Major Project, the Fostering ALfred CONstruction (FALCON) consortium, in charge of all activities advancing the ALFRED Project, has identified the need to organize and refine the presently available material in an organic, harmonized and complete way. The results of this effort brough to the drafting of a comprehensive document presenting an exhaustive and effective overview of the entire Project: the ALFRED White Book.
A first step of knowledge of the Project – including the ALFRED demonstrator and its associated research infrastructure – is presented in the White Book, addressing its main elements under all the relevant points of view to provide an exhaustive (although qualitative) picture of the proposal and the expected benefits underneath. This includes mostly: the Project vision and mission, the scientific and technological relevance, the technical description, the supporting R&D programme and the operation strategy, the key safety and licensing elements, the socio-economic benefits, and the proposed deployment strategy and project implementation (including financial and managerial aspects).
The White Book is thus conceived as a “business card” of the ALFRED Project, meant for distribution to different stakeholders with a potential interest for involvement, including research institutes, universities, industries, safety authorities, policy makers as well as national and European Administrations and local communities.

Speakers: Francesco Lodi (ENEA) , Giacomo Grasso (Italian National Agency for New Technology, Energy and Sustainable Economic Development (ENEA)) , Gabriele Firpo (Ansaldo Nucleare SpA) , Ms Daniela Diaconu (RATEN/ICN)
• 39

We have investigated main neutronic parameters and core performance parameters of the reference China Initiated Accelerator Driven System (CiADS) core after loading uranium nitride instead of uranium oxide. Based on neutronic calculations performed with the Serpent code, only 15.2% enrichment is needed to provide similar sub-criticality as the reference design when using UN instead of oxide fuel. Moreover, if we applied the same enrichment level as the one in the reference design, the core size can be reduced by 25%, which eventually permits around 30% higher spallation source efficiency.

Speaker: Dr Youpeng Zhang (Fudan University)
• 40
A Study on the Development of a Procedure Complexity Evaluation and Optimization for Operating Procedures of China Experimental Fast Reactor

If the procedure can be decomposed and the complexity of the evaluation procedure can be quantified, it is an effective way to reduce the operator's workload and reduce the human error rate. In this study, a measure is developed, based on the task complexity (TACOM) evaluation method, combined with the characteristics of the China Experimental Fast Reactor. The TACOM measure consists of five sub-measures that can cover remarkable complexity factors, includes 1) task logic complexity, 2) operation step complexity, 3) information complexity, 4) knowledge complexity, 5) decision complexity. The method established a procedure complexity calculation process including complexity entropy calculation, questionnaire survey, and procedure optimization. To verify the method, the initial complexity scores are compared with optimization scores, from this comparison, the research proposes the following optimization suggestions for the human-machine interface: increase the "power monitoring device" indicator to reduce the information acquisition complexity.

Speaker: Mrs Qi Zhou
• 41
CORROSION HYDROGEN MASS TRANSFER IN FAST REACTOR STEAM GENERATORS OF THE SODIUM-WATER TYPE

The operation experience of sodium-water steam generators has demonstrated that the main source of hydrogen ingress into the secondary sodium is the process of hydrogen depolarization (equation 2) of the cathodic electrochemical corrosion process (equation 1) on the steam-generating steel surfaces from the water side:
3Fe + 4H2O = Fe3O4 + 4H2 (1)
2H2O+2e- = 2Hkt + 2OH- (2)
Hydrogen generation and its concentration in water and sodium increases in the course of chemical washing and when a steam generator is started up, as with them the metal corrosion rate goes significantly up. Determination of the predominant direction of hydrogen transfer is required, for example, for estimation of steam generator tubes’ 10X2M steel hydrogenation.
In this work, the specific rates of hydrogen supply to sodium were determined with analyzing the data of systematic monitoring of the hydrogen content in the third (steam-water) and secondary (sodium) circuits. The mechanism of diffusion of hydrogen formed in the cathode sections of the circuit through steel was considered in terms of two processes, i.e. recombination of H atoms adsorbed into H2 molecules with their subsequent release into the solution, and dissolution of atomic hydrogen in the metal itself with diffusion into sodium of the secondary circuit, where hydrogen is bound, thus forming sodium hydride. Based on the assumptions made and the calculations performed, as well as with consideration of the BN-600 reactor operational data, the following results were obtained:
- The content of hydrogen dissolved in the third circuit water does not influence the flow.
- Hydrogen diffusing through a steel wall of steam generating tubes is of electrolytic (corrosion) origin.
- At a low corrosion rate typical of the nominal operating mode of the sodium-water steam generator, most of the corrosion hydrogen (about 90%) diffuses into the secondary sodium, and at a significant corrosion rate (typical of chemical washings) it goes into water. Therefore, it is not correct to control the metal corrosion of the steam-generating surfaces in the nominal operating mode of the sodium-water steam generator only by the difference in the hydrogen concentration in the superheated steam and in the feed water.
- It was found that at all the real rates of metal corrosion in heat-exchange tubes in a steam-water environment, hydrogenation of 10X2M steel is insignificant. Thus, hydrogen embrittlement of steel is practically ruled out in the course of steam generator operation and chemical washing of its pearlitic surfaces.

Speaker: Vladimir Smykov (SSC IPPE, JSC)
• 42
Design of experimental scheme for activation method of China demonstration fast reactor

Aiming at the 600MW demonstration fast reactor（CFR600）activation method experiment, this article introduces the comprehensive experimental measurement system and the basic flow of the experiment, describes the experimental principle and the expected experimental content, and focuses on the design of the CFR600 activation method experiment. First, according to the project schedule, the characteristics of the CFR600 core and the experience of the activation method of the China Experimental Fast Reactor have preliminarily determined the number of experimental components, the batch of radiation into the reactor, the irradiation power, and the location of the measurement point of the activation method, and secondly, based on the NAS code for fast reactor, through theoretical calculation and analysis, the radial and axial measurement points of the core nuclear reaction rate the position of the power monitoring component are obtained. Finally, it is concluded that the introduction of activation method experimental reactivity has little influence on the reactor core, and a set of activation method experiment scheme suitable for 600MW demonstration fast reactor is formed, which provides theoretical and technical guidance for subsequent CFR600 to carry out activation method experiment.

Speaker: Ms XIAO HU (China Institute of Atomic Energy)
• 43
Development of a 15 kg servo manipulator for remote handling applications

Remote Handling is one of the prominent areas of research and development for the nuclear sector for the remote manipulation of irradiated nuclear fuels and structural materials. The standard coding for the Remote handling devices for radioactive materials is covered as part 1: General requirements under ISO: 17874-1:2004 & part 3: Electrical master-slave manipulators (Servo Manipulator) under ISO:17874-3:2019. Though the master-slave mechanical manipulators have found wide applications in the nuclear domain especially in the hot cells, but recently the electric manipulators find its application in both nuclear fuel fabrication and in carrying out complex tasks inside the hot cells. This also finds its application in the decommissioning operations of nuclear reactors. Under the mobile manipulators it is further classified into power manipulator and servo manipulator. The servo manipulators are closed-loop feedback system, which enable the sensing of slave side forces and reflect it on the master end. This makes the operator feel the slave environment remotely as tangible forces on the slave end-effecter are directly sensible. The dual arm design makes it further more easy to perform complex tasks with delicate handling of the slave objects. A 15 kg payload capacity servo manipulator was designed, developed and qualified at IGCAR. The real time force synthesis is done using the Jacobian matrix linearizing method for the inverse kinematic and force synthesis and the Jacobian matrix is constructed from the Jacobian generating vectors by the slave side manipulator pose/orientation. This is done by directly networking all the servo drive controllers and reading the slave motor encoder positions. Further, the motor torque is evaluated from the motor currents from all the slave joints and the torque vector is constructed. By calculating the work done by the manipulator the end effecter Cartesian forces and the direction vectors are computed and the same is applied on the master end thereby simulating the slave environment. This paper deals with the design of the servo manipulator for the remote handling application.

• 44
EXPORT OF RBN WITH SNCD AND NUCLEAR PROLIFERATION RISKS

Currently, the world society recognizes serious environmental problems with anthropogenic impact on the environment and almost does not dispute the main source of its pollution. However, the hope that growing demand of world economies for electricity can be provided by so-called alternative energy sources - solar and wind installations is illusory.
At the same time, the world's existing nuclear power industry consisting of more than 80% of light-water reactors of various capacities also cannot solve this problem as well. The real solution to the energy problem while preserving the climate and achieving the UN Sustainable Development Goals can only be large-scale nuclear energy based on fast neutron reactors (FNR) with a closed nuclear fuel cycle (CNFC). However, the development of such large-scale nuclear energy could create potential risks to the international nuclear non-proliferation regime and may complicate the IAEA's safeguards implementation in non-nuclear-weapon States operating such nuclear power systems.
The report using the example of the BREST reactor as part of a pilot demonstration energy complex being built in Russia will show that these potential risks can be reduced due to the technical features incorporated in the design of such nuclear system. In particular, the absence of separation of uranium and plutonium during reprocessing leads to a high level of radioactivity of the new fuel obtained from the reprocessed one providing self-protection from unauthorized use.
The report will also discuss elements that can be included in the design of the facility that promote the application of IAEA safeguards, including the availability of an automated system for accounting for nuclear materials, as well as online measurements of flows of nuclear materials to ensure the necessary accuracy of measurements of nuclear materials in bulk form.
The report will show that the results of the analysis based on the developed models can make it possible to conclude that the inherent technological elements of such a nuclear system are sufficient to reduce the potential risk of nuclear proliferation.
Based on this, the report will conclude that the future export of FNR with CNFC from Russia with an initial load of fresh fuel and technological processes developed to the best levels will create minimal potential risks to nuclear proliferation, and the use of such nuclear power systems in peaceful nuclear activities of states will contribute to their economic and technological development without negative impact on the environment.

• 45
INTEGRATED RADIATION AND HYGIENIC APPROACH TO PRODUCTION SAFETY. ASSESSMENT OF THE IMPACT ON PUBLIC HEALTH

The development of uranium-plutonium compounds as nuclear fuel is a new nuclear technology using poorly studied highly active uranium compounds. The toxicity and consequences of the biological action of such compounds are the subject of this work.
It should be noted that during the production of MNUP fuel there is a possibility of impact on the population and the environment, which is mainly due to the involvement of plutonium in the nuclear fuel cycle, as it is more radiation hazardous than uranium.
The report devoted to this work will discuss the multifactorial effects of MNUP fuel components on public health: external photon and neutron irradiation and internal irradiation due to inhalation of 238U, 239Pu and their decay products.
Based on the presented data, a scientifically grounded approach to assessing the impact of the factors of MNUP fuel components on the health of various critical groups of the population will be proposed.
To develop this scientifically grounded approach, the results of the analysis of the available experimental estimates of the toxicity of radionuclides and their compounds that make up the MNUP fuel and the results of our own research at individual stages of the NFC will be used.
The method used in this work for assessing the health indicators of the population living in the immediate vicinity of enterprises working with MNUP-fuel will make it possible to offer recommendations for doctors monitoring the health of the population.

Speaker: Dr Evgeny Metlyaev (State Research Center – Burnasyan Federal Medical Biophysical Center of Federal Medical Biological Agency)
• 46
New Concepts and Methodologies for the Effective Deployment of Gen IV reactors

Public concerns about nuclear safety and environmental impact of the operations of nuclear power plants and accident scenarios, have spurred innovative research into the study of advanced reactor fuels with the fast breeder reactors belonging to the new age generation reactors.
These reactors termed Gen IV reactors are evangelized for their many positives specially to mitigate the issues of waste management and proliferation while maintaining the high safety standard known to the industry. Applied for their high burn-up capabilities, spent fuel and weapons grade nuclear materials consumption thus making them exceptional.
The technology has also presented the industry as one which places great importance on pioneering ways in tackling issues and challenges on the global front just like its collaborative efforts in the sustainable development goals and continually finding ways to mitigate those issues. However, one of the challenges being passed on is the ability for the industry to analyze and develop a technical document by way of regulation to aid the commercialization of the breeder and where necessary, a practical and achievable policy for its global deployment.
Additionally, with the extension of life of reactors, deployment of new builds, decommissioning of NPPs or even the utilization of different fuel matrix for a range of concept units, developing critical methodologies is imperative to serve as criterion for perfecting this technology and its’ eventual deployment and the intention of the writer to present some approaches to address these.

Speaker: Mr BOMA WILCOX (Nigeria Atomic Energy Commission)
• 47

Currently the largescale nuclear power system, based on the two-component technology platform has been developed in Russia. The technology implies the buried spent fuel to be safe when its and natural uranium impacts on human health to be equal.
To 2100 fast reactors (FR) and thermal reactors (TR) will consume 541.7 thousand tons of natural uranium, due to the reactors operation 7.523 thousand tons of long-lived radioactive waste (RW) will be produced. With the account of ratio of natural uranium mass to RW mass the equivalence of committed effective doses to the public from the natural uranuim and RW (radiation equivalence) will be achieved after 287 years of the waste disposal, radiation-associated lifetime attributable risks (LAR) will be equal (radiological equivalence) after 99 years of RW disposal.
If two-fold uncertainty in radiation doses assessment occurs, radiological equivalence will be achieved after about 270 years of the waste disposal, however, radiation equivalence will be practically unachievable. If uncertainty in radiation doses assessment is about 30%, radiological and radiation equivalences will be achieved after about 100 and 700 years of the disposal respectively. If RW disposal lasts longer than100-150 years 241Am becomes the dominant contributor to radiation dose and radiation risk. The fraction of all americium radiosotopes in RW mass is 0.23%. If the Am fraction in the waste mass increases by 10 times, radiation equivalence will not be achieved even after 1000 years of the disposal, while radiological equivalence will be achieved after 414 years, that is 315 years more than needed to achieve radiological equivalence for the RW of the original composition, 99 years. So, sepa¬ration of americium from RW will result in reducing of the time to achieve equivalence of effective doses and LARs from depleted fuel and natural uranium. Radiological equivalence principle may be used to justify substantial reduction of the RW disposal duration. If thermal reactors only are used in the nuclear energy system, radiological equivalence may be achieved after more than 20,000 years of RW disposal.

Speaker: Viktor Ivanov
• 48
Novel Electrical, Electronics and Instrumentation systems for Fast Reactor Fuel Reprocessing Plants

Measurement of gas flow rates in exhaust ducts for radioactive plants are essential from process as well as radiation safety requirements. Averaging Pitot Tube (APT) is mechanical device which generates differential pressure proportional to flow rate when inserted in the path of the flow. In this paper we discuss the reason for the selection of this particular type of flow measurement technology, design and testing methodology followed, challenges in installation & commissiong and implementation of advanced wireless communication for signal transmission from field to control room.

Criticality Alarm System (CAS) is mandatory in nuclear fuel fabrication and reprocessing facilities for round-the-clock monitoring of criticality accident. The CAS shall not fail to detect any criticality event or trigger a false criticality alarm. To meet this requirement, online fault diagnostics, surveillance for CAS electronics from the control room is mandatory. A fault diagnostic system is developed to detect the failures associated with AC mains, low voltage, high voltage, system on battery, battery isolation diode and single-channel alarm of each CAS. It is connected to the plant alarm annunciation based on fail-safe condition.

The selection of materials for each component of light fixture such as glass tube, terminal electrodes, holder etc. also very important to survive high level radiation (approximately 1.15E+04 R/h) inside the hot cell with minimum reduction in lux output from the light fixture. A 40 W Heat pipe based Cob LED lamp with copper fins was selected as suitable for application. This paper also discusses the experimental setup prepared for development and calculations made.

The objective of this paper is to discuss recent developments and our experience in testing, qualification & operation of RPM’s and compare various standards (ANSI, IEC & ASTM) associated with them.

Qualification of RPM’s as per specified Minimum Detectable Activity (MDA) is challenging owing to background intensity variation in operating areas. This paper also talks about acceptable false alarm rates, position of RPM’s in nuclear facilities, measurement time (Walk through vs Wait in Monitor), Optimal Testing & calibration interval and in- situ calibration of RPM’s after installation in site.

NDAS, a 2-tier architecture VME bus based system, consisting of a M68020 CPU card and 5 VME based A16/D16 double Euro 30-ch analog input (AI) cards was developed. VME application software is developed using C, compiled using tasking cross compiler and fused into EPROM. The GUI for NDAS is developed using Qt kit on Linux platform.

Speakers: Mr Prakash Bhanu, Mr Mathews Geo
• 49
Nuclear Hydrogen and Fast Reactors

Hydrogen and hydrogen technology are expected to have a key role as an energy carrier for the technic and economic systems. This are expected to be a new impulse for the nuclear’s integration in the grid.
Large-scale demonstration projects of the low carbon hydrogen production require major investments by countries and long term strategies. Taking into account the specific of nuclear industry, that also require long term strategies and substantial investments, both hydrogen production and nuclear power can complement each other in future development.
From technical point of view, this binder is thermochemical cycle. This hydrogen pathway comprises the oxidation of a metal oxide or an oxidable compound (e.g. iodine-sulphur) by a reaction with water and, second, the recycling of this compound, that takes place at higher temperatures by stripping off one oxygen atom.
This paper summarizes the current concerns of the hydrogen community in finding solutions based on scientific fundamentals for hydrogen production using nuclear energy.

Speaker: Dr Ioan Iordache (ICSI Rm. Valcea)
• 50
Numerical Investigation of Cellular Convection in the Cover Gas space of Fast Breeder Test Reactor

Tilting of reactor vessel and other components due to circumferential temperature gradient is a critical safety and operational issue in loop type Fast Reactors. Natural convection of reactor cover gas developed in the narrow component penetrations is the main cause for such a phenomenon. Numerical studies have been carried out using CFD code to investigate the possibility for development of circumferential temperature gradient in the reactor vessel of Fast Breeder Test Reactor (FBTR). The computational model which accounts for all the three modes of heat transfer has been validated against experimental data from a mock-up experimental study as well as measured data from the plant. The analysis of natural convection phenomenon in the annular region of cover gas space reveals the formation of cellular convection pattern with single convective cell. This causes circumferential temperature gradient along the reactor vessel. The temperature asymmetry is also found to vary with elevation, with a maximum value at the plug bottom. With respect to the computational model adopted, parametric studies have been carried out using various turbulence models and by varying the radiation emissivity of different surfaces present in the domain. With respect to the system design and operating conditions, parametric studies have been carried out to investigate the influence of anti-convection barrier incorporated in the design and the operating status of plug cooling and biological shield cooling systems. It is seen that the anti convection barrier and labyrinth help in significant reduction of temperature asymmetry in the reactor vessel. The operation of reactor plug cooling has been found to be very effective in reducing the temperature asymmetry as well as the average temperature of reactor vessel in the cover gas space, while the biological shield cooling system is not found to provide significant effects. Another important study that is carried out is the analysis of the natural convection behaviour as a function of the thermo-physical properties of the cover gas medium. The relative performance of Helium and Argon as cover gas medium has also been investigated. It is seen that when helium is selected as the cover gas medium, it causes lower temperature asymmetry in the system as compared to argon. This is attributed to higher thermal diffusivity of helium compared to that of argon. Hence, Helium is desirable as a cover gas medium from cellular convection considerations.
Key Words: Circumferential Temperature Gradient, Fast Reactors, Reactor vessel, CFD, Turbulence Modeling

Speaker: Mr Kanha Chaturvedi (Dept. of Atomic Energy, IGCAR)
• 51
RISK FACTORS OF COMPLEX RADIATION AND NON-RADIATION EFFECTS ON THE HEALTH OF PERSONNEL IN ASSESSING THE IMPACT OF THE PRODUCTION OF INNOVATIVE FUEL FOR FAST REACTORS

The report will present the results of assessing the risk of the impact of a complex of negative factors of production of MNUP-fuel of a radiation and non-radiation nature on the health of workers. Such factors include both the combined effect of external photon and neutron irradiation due to the inhalation of 238U, 239Pu and their decay products, and the toxic effect of the compounds included in the MNUP fuel and formed during its production.
The work used estimates of the state of primary and chronic morbidity of 50 workers at the experimental production of MNUP fuel for the observation period from 2007 to 2018.
Revealed relatively high levels of morbidity, which cannot be explained by the action of only the radiation factor.
A method for assessing the relative risk of the influence of a complex of factors of the radiation and non-radiation nature of the production of MNUP-fuel on the health (morbidity) of personnel, both as a whole and of individual organs and systems, is proposed.
The proposed assessment of the relative risk of the influence of a complex of production factors on the health of workers is an indicator of the safety of a complex radiochemical production with nuclear technology and is a “Health Passport” (safety) for the production of MNUP fuel. The results of the analysis of the health of workers in the production of MNUP-fuel will serve to improve the system for optimizing the control of the radiation-hygienic situation in production and will become the basis for solving medical problems when building a radiation safety system at all stages of handling a new type of fuel.

Speakers: Dr Oleg Parinov (Burnasyan Federal Medical Biophysical Center of Federal Medical Biological Agency) , Prof. Alexander Samoilov (Burnasyan Federal Medical Biophysical Center of Federal Medical Biological Agency)
• 52
Sustainability of nuclear and non-nuclear power generation options under Russian conditions: a comparative evaluation study

This article provides the results of a case study on a comparative evaluation and ranking of six power generation technologies including 2 nuclear, 2 fossil fuel and 2 renewable power generation options. The performed analysis was based on the combination of the approaches proposed within the project New Energy Externalities Development for Sustainability (the NEEDS project) and the INPRO/IAEA KIND collaborative project (so-called the integrated NEEDS & KIND framework). The set of key indicators (36 key indicators arranged in a four-level objectives tree) along with approaches for their assessment including relevant databases were borrowed from the NEEDS project and adapted for the present study. The approach for judgments aggregation is based on the IAEA/INPRO approach to carrying out comparative evaluations (the KIND approach). Also, the study includes the results of the advanced sensitivity/uncertainty treatment with respect to weighing factors and key indicators using relevant tools developed within the INPRO/IAEA CENESO collaborative project. The presented results demonstrate the outstanding performances of nuclear power technologies (especially fast reactor technologies) for meeting sustainable energy development goals in the case of Russia whose performance even better than the performance of renewable energy sources.
Keywords: power generation technologies, multi-criteria analysis, ranking, uncertainty / sensitivity analysis, project NEEDS, project KIND, IAEA, INPRO

Speaker: Stepan Kviatkovskii (SC Rosatom)
• 53
Two-Component Energy Industry under Conditions of Closed Nuclear Fuel Cycle: Economic Benefits

Transition to the system of the closed nuclear fuel cycle based on reactors on thermal and fast neutrons is an important stage of formation of two-component energy industry, which provides enhancement of efficiency of uranium resource efficiency. The closed nuclear fuel cycle at the international level can be presented both from the viewpoint of energy security and stability and from the viewpoint of increase of competitiveness of nuclear power solutions.
However, countries-owners of fast reactor technologies use different strategies of transfer to the closed fuel cycle.
By the example of fast sodium-cooled reactors and VVER, the paper will consider economic expediency of implementation of the closed fuel cycle for the purpose to provide stable ecologically clean energy.
The paper will present approaches to estimation of commercial advantages of commercial fast reactor application both within one system and in the international context, reveal drivers and limitations of commercialization from viewpoint of economic parameters over the medium term.

Speaker: Mariia Roslaya
• 54
ULTRASONOSCOPY SYSTEM “VIZUS” FOR SODIUM-COOLED BN-TYPE REACTORS

Development of ultrasonoscopy system “Vizus” is aimed at increase of operation safety of power units with sodium cooled BN-type reactors. “Vizus” system application permits to detect foreign objects in the space above the core and, as a result, to prevent failure of devices passing through the CPS column and core elements at rotary plug rotation.
The system is intended for registration and determination of coordinates of foreign objects being in the space between ends of top nozzles of core assemblies and lower altitudes of devices passing through the CPS column before refueling start and, if necessary, during reactor refueling.
The system is based on the principle of ultrasonoscopy of the space and registration of reflected signals, which permits to detect obstacles to refueling activities.
The system shall consist of one or two identical manipulators which are installed in special penetrations in the reactor vessel and spaced from one another, a control system and an operator work place. The manipulator is intended to carry an ultrasonic transducer to the area under control and to point an ultrasonic beam in the sodium environment at ultrasonic beam reflectors located over the core perimeter. Signals reflected from ultrasonic beam reflectors come to the control system and permit to make a conclusion on presence of foreign objects in the space under control.
Now, this system is at the stage of implementation at the BN-800 reactor.

Speaker: Mr Dmitrii Lesiukov (JSC “Afrikantov OKBM”)
• Special Session: IAEA Coordinated Research Projects

Chairs: Vladimir Kriventsev and Nikoleta Morelová

• 55
Neutronics Benchmark of CEFR Start-Up Tests: An IAEA coordinated research project
Speaker: Nikoleta MORELOVA (IAEA)
• 56
CEFR Physical Start-Up Tests: the Core Specifications and Experiments
Speaker: Mr Xingkai Huo
• 57
Verification and validation of neutronic codes using the start-up fuel load and criticality tests performed in the China Experimental Fast Reactor
Speaker: Dr Armando Miguel Gomez Torres (Instituto Nacional de Investigaciones Nucleares)
• 58
Neutronics Benchmark of CEFR Start-Up Tests: Temperature Coefficient, Sodium Void Worth, and Swap Reactivity
Speaker: Jiwon Choe (Ulsan National Institute of Science and Technology )
• 59
Neutronics Benchmark of CEFR Start-Up Tests: Reaction Rates and Reactivity Coefficients
Speaker: Taek Kyum Kim (Argonne National Laboratory)
• 60
BLIND PHASE RESULTS FOR TRANSIENT SIMULATIONS OF THE FFTF LOSS OF FLOW WITHOUT SCRAM TEST #13
Speaker: Anton Moisseytsev (Argonne National Laboratory)
• 61
Blind-Phase Results of the FFTF Neutronic Benchmark
Speaker: Nicolas Stauff (Argonne National Laboratory)
• Coffee Break
• 2.1 General Safety Approach

Chairs: Tanju Sofu and Xiaoyan Yang; Anton Moisseytsev

• 62
Basis for the Safety Approach (BSA) for Design & Assessment of Generation IV Nuclear Systems
Speaker: Paul Gauthé (CEA)
• 63
EXAMPLES OF AREAS OF NOVELTY IN LIQUID METAL FAST REACTORS TO CONSIDER IN THE REVIEW OF APPLICABILITY OF THE IAEA SAFETY STANDARDS: FISSION PRODUCT RETENTION BARRIERS: DIFFERENCES BETWEEN LIQUID METAL FAST REACTORS AND LIGHT WATER REACTORS
Speaker: Paula CALLE VIVES (IAEA)
• 64
System Safety Assessment of the Generation IV Lead Fast Reactor
Speaker: Kamil Tucek (European Commission, Joint Research Centre)
• 65
France-Japan Collaboration on the SFR Severe Accident Studies: Outcomes and future work program
Speaker: Mr SHIGENOBU KUBO (JAEA)
• 66
Safety Analysis of the ARC-100 Sodium-Cooled Fast Reactor
Speaker: Tyler Sumner (Argonne National Laboratory)
• 67
Approach for ALFRED licensing in Romania
Speaker: Giacomo Grasso (Italian National Agency for New Technology, Energy and Sustainable Economic Development (ENEA))
• 68
Activities of the GIF Safety and Operation Project of Sodium-Cooled Fast Reactor Systems
Speaker: Hidemasa Yamano (Japan Atomic Energy Agency)
• 69
Application of the practical elimination concept within the framework of the ESFR-SMART project to improve the intrinsic safety of the sodium-cooled fast reactor
Speaker: Mr Joel Guidez (CEA)
• 70
Application of a Risk-Informed Performance-Based Approach for the Authorization of the Versatile Test Reactor
Speaker: Jason Andrus (Idaho National Laboratory)
• 7.1 Sustainability: Economics, Environment, and Proliferation

Chairs: Brian Boyer and David Settimo

• 71
Modeling the optimal economic structure of a global deploying nuclear power system with fast and thermal reactors in a partially closed nuclear fuel cycle
Speaker: Alexander Yegorov (IPPE)
• 72
KEY ASPECTS OF COMPETITIVENESS FOR INDUSTRIAL ENERGY COMPLEX WITH FR AND CLOSED NFC
Speaker: Mr Dmitriy Tolstoukhov
• 73
TECHNICAL AND ECONOMICAL FEATURES OF COMMERCIAL SODIUM FAST REACTOR IN FRANCE
Speaker: Mr DAVID SETTIMO (EDF)
• 74
SPECIFIC FEATURES OF THE EXPORT OF RUSSIAN TECHNOLOGIES OF FAST REACTORS AND A CLOSED NUCLEAR FUEL CYCLE
Speaker: Dr Aleksandr CHEBESKOV (IPPE)
• 75
Development status of commercial SMRs and its experience to China
Speaker: Ping Li (China Institute of Atomic Energy)
• 76
Effect of Reactor Technology on Economics of SMR Projects
Speaker: Ilya Zhuravlev
• 77
Comparative multi-criteria analysis of scenarios of the Russian nuclear energy development in the context of uncertainty knowledge about the future
Speaker: Alexander Yegorov (IPPE)
• 78
EFFECTIVE FUEL SUPPLY OF TWO-COMPONENT NUCLEAR ENERGY SYSTEM WITH VVER-BN REACTORS
Speaker: Elena Marova (JSC “Afrikantov OKBM”)
• 79
Export Potential and Commercialization Conditions of Fast Reactors Considering Non-Proliferation Items
Speaker: Nadezhda Salnikova (JSC "Afrikantov OKBM")
• 80
TECHNOLOGICAL SUPPORT OF THE NON-PROLIFERATION FOR SVBR-100 FUEL CYCLES
Speaker: Dr Oleg Komlev (JSC “AKME-engineering”)
• 81
The INPRO project studies on the double-component nuclear power systems with the closed fuel cycle and fast reactors: past and future
Speaker: Alexander BYCHKOV (Senior Nuclear Engineering Expert - INPRO)
• 8.1 SFR Commissioning, Operation, and Decommissioning

Chairs: M.Thangamani and Mr. Masanobu ARAI

• 82
Reactor Core Viewing System for the pre-commissioning stage inspection of reactor core components of Prototype Fast Breeder Reactor
Speaker: Chellapandian Ramalingam (Reactor Design and Technology Group, Indira Gandhi Centre for Atomic Research)
• 83
Experience in Preheating of PFBR Reactor Assembly
Speaker: A JYOTHISHKUMAR (D A E)
• 84
Commissioning and Operating Experience for Secondary Sodium Systems and is Auxiliaries of PFBR
Speaker: Mr NISHANTH SAHU (DAE)
• 85
Advanced in-situ Calibration and Probe Release Mechanism for PFBR SG Inspection System (PSGIS)
Speaker: Joel Jose
• 86
Treatment of sodium of Superphenix Fast Breeder Reactor
Speaker: Dominique VILLANI (Framatome)
• 87
Operating Experience of FBTR
Speaker: THANGAMANI M (IGCAR,INDIA)
• 88
Fuel handling Experience of FBTR
Speaker: Mr MURALITHARAN G (IGCAR / DAE / KALPAKKAM)
• 89
EXPERIENCE OF OPERATIONAL CHEMICAL CLEANING OF BN-600 STEAM GENERATOR EVAPORATORS FROM CORROSION PRODUCT DEPOSITS
Speaker: Ms Kristina Legkikh
• 90
Design, Experimental trials and Qualification of explosive welding technique for plugging of degraded PFBR Steam Generator tubes
Speaker: Visweswaran Padmanabhan (Reactor Design and Technology Group, Indira Gendhi Centre for Atomic Research)
• 91
BOR-60 REACTOR OPERATING EXPERIENCE, WORK ON IMPROVING SAFETY AND EXTENDING LIFETIME
Speaker: Dr Alexey Izhutov (JSC SSC RIAR)
• 92
PROBLEMS OF DECOMISSIONING FAST REACTORS AND WAYS OF THEIR SOLUTION ON THE BASIS OF THE BR-10 RESEARCH REACTOR
Speaker: Dr Vladimir Smykov (SSC IPPE,JSC)
• Wednesday, April 20
• Plenary 2. International Organizations and YGE Winners

• 93
European Commission (EC) Key Note
Speaker: Ms Maria Betti (TBV)
• 94
Generation IV International Forum Key Note
Speaker: Mr Bob Hill
• 95
OECD/Nuclear Energy Agency Key Note
Speaker: Ms Tatiana Ivanova
• 96
International Atomic Energy Agency (IAEA) Key Note
Speaker: Ms Aline Des Cloizeaux
• 97
YGE Winner: Advanced Functional Materials for Next-Generation Fuel Reprocessing
Speaker: Mr Kuntal Kumar Pal
• 98
YGE Winner: Production of Mo-99 isotope in the BN reactor by beryllium blocks
Speaker: Ms Oksana Kucheryavykh
• 99
YGE Winner: Small Modular Fast Reactors for the ASEAN Region: Implementation of the TRISO Fuel Particle Concept as a Regional Variant of the Fast Reactor
Speaker: Mr Tan Zhe Chuan
• Coffee Break
• Poster Session
• 100
CURRENT STATE AND ISSUES OF THE HEAVY LIQUID METAL COOLANT TECHNOLOGY DEVELOPMENT (PB, PB-BI)
• 101
DEVELOPMENT OF SUBMERGED ELECTROMAGNETIC PUMP FOR LIQUID LEAD

Development results of submerged electromagnetic pump (EMP) on liquid lead for reactor BREST-OD-300 are presented. EMP is planned to use for liquid lead level regulation in the reactor during putting it into exploitation under partial or full EMP submerging in lead. Main EMP parameters: pressure head 1.0 MPa, nominal flow rate 2.0 m3/hr, lead temperature 390-420 °С. Required service life time under nominal conditions is 1000 hours. EMP optimum design was chosen – cylindrical linear induction pump (CLIP), calculations supporting its main characteristics were done, list of necessary tests for supporting design and main parameters was estimated.

Speaker: Mr Denis Obukhov (Joint Stock Company “D.V. Efremov Institute of Electrophysical Apparatus” (JSC «NIIEFA»))
• 102
Experimental study on sodium insulation interaction and its effect on structural material

Sodium, owing to its high heat transfer properties and excellent compatibility with structural materials is the preferred coolant for Liquid Metal cooled Fast Breeder Reactors (LMFBR). Apart from its favorable properties, sodium also poses a concern due to high chemical reactivity with air. Accidental sodium leaks from secondary circuit may result in fire due to sodium reaction with oxygen and moisture in ambient air. Usually, leak incidents with leak rates above 100 g/h are detected by leak sensors. However, if the leak origin is a pin hole or a hairline crack, sodium leak can go undetected and may not lead to fire. In fact, sodium reacts with thermal insulation mounted over the piping and oxygen available within leak vicinity. The solid reaction products can settle near the leak path and eventually plug the leak path, even before the leak detectors respond. The structural material at the leak zone can undergo localized degradation under the influence of sodium, oxides and other reaction products of sodium. A detailed experimental investigation was carried out to study sodium interaction with thermal insulation and effects of reaction products on the structural material. Sodium leak experiments were performed with rock wool insulation and specimens of SS-316 LN at a temperature of 300oC in air. Subsequently, post test analysis of the samples by using X-ray Diffraction (XRD) revealed presence of silicates and aluminates of sodium in the outer layers and complex ternary and quaternary oxides of sodium, iron, chromium and nickel in the inner layers. Field Emission Scanning Electron Microscopy (FESEM) analysis of the specimen revealed deposition of oxides on the exposed surface. Elemental analysis of samples was carried out by Energy Dispersive Spectroscopy for deducing the leaching of chromium, nickel and iron from the exposed surface. Fatigue crack growth test of the exposed specimen was done for assessment of reduction in fatigue life. The paper briefs about the interaction of sodium with thermal insulation, characterization methods for reaction product analysis and consequent degradation of structural material.

Speaker: Mr Avinash Ch SSS (Indira Gandhi Center for Atomic Research)
• 103
HEAT TRANSFER CALCULATION AND SERVICE LIFE TIME ESTIMATION OF SUBMERGED ELECTROMAGNETIC PUMP FOR LIQUID LEAD

Results of heat transfer analysis and service life time estimation are presented for submerged electromagnetic pump (EMP) on liquid lead for reactor BREST-OD-300. EMP is planned to use for liquid lead level regulation in the reactor during putting it into exploitation under partial or full EMP submerging in lead. Main EMP parameters: pressure head 1.0 MPa, nominal flow rate 2.0 m3/hr, lead temperature 390-420 °С. Required service life time under nominal conditions is 1000 hours. Cylindrical linear induction pump was chosen as a result of optimization. Heat transfer analysis was done with ANSYS. Maximum EMP winding temperature was found to be 471 °С. Life time estimation of EMP based on these results proved that required duration 1000 hours was provided. The design of the mock-ups was chosen for carrying out life tests and confirming the pump resource.

Speaker: Valeriia Federiaeva (JSC "NIIEFA")
• 104
Impact of Cladding Material on Neutronic Balance in Breed-and-Burn fast reactors

Breed and Burn fast reactors are very attractive reactor concept. It can achieve very high burnup without reprocessing facility using depleted uranium or natural uranium as the fuel for the reactor. To achieve criticality in the equilibrium burnup condition improvement of neutron economy is the essential issue. The feasibility of the reactor can be discussed based on the neutron balance of a fuel assembly from the loading to the discharge. There is a possibility that cladding materials give a lot of impact on the neutron balance. The purpose of study was to make clear the impact of cladding material on the neutron balance. In the previous study, HT-9 and ODS were assumed as the cladding material. The thickness was decided based on an example of previous fuel design. In the study, the impact of cladding material on the neutron balance was analyzed by changing the cladding material and the thickness. The analysis results showed the difference of neutron balance between HT-9 and ODS is negligible. Significant improvement of neutron balance can be expected by the reduction of thickness of cladding, which gives the same effect by the reduction of the density of cladding material. The change of burnup characteristics in whole core analysis was discussed also by the change of density of cladding.

Speaker: Toru Obara (Tokyo Institute of Technology)
• 105
Impact of Core Materials on The Fuel cladding Irradiation Damage in Breed-and-Burn Fast Reactors

The breed and burn fast reactor concept have been a familiar with nuclear engineers since the 1950s in which the depleted or not enriched fuel is loaded in a core and breed the fissile fuels while burning them. Development of Breed-and-Burn (B&B) reactor means of effective utilization of uranium resource and reduction of spent fuel. There are issues to be solved as of irradiation damage in high burnup and reducing the Displacement-Per-Atom (DPA) for the same burnup. However, there is no clear investigation on optimum combination of core material as fuel and coolant for bread and burn reactor possible and on impact on DPA by the use of different combinations of core material. The objective of the study is to make clear impact on DPA for several materials by performing burnup analysis using Monte Carlo SERPENT code. DPA calculation and neutron balance analyses for the infinite fuel cell in which different combination of fuel as metallic, natural and enriched nitride, and oxide ones and coolant as helium, sodium, lead and lead-bismuth were performed in order to find out the optimum combination to make the DPA minimum. The analysis results showed the DPA at the same burnup is the lowest for oxide fueled core and the highest for metallic fueled one. For metallic fueled core, the spectrum is the hardest while it is softer for oxide and nitride fueled cores. But, the minimum burnup required to sustain a B&B mode of operation in a large fast reactor core is the lowest for core with combination of metallic fuel and any type of coolant compared to others. As the results, the DPA at the minimum burnup to make the B&B fast reactor critical can be the smallest in metallic fuel. It means the use of metallic fuel has some advantage to reduce DPA in B&B fast reactors. In the case, the use of helium gas as coolant can make the DPA smallest. The DPA by using other material can be expected almost the same in B&B fast reactors.

Speaker: Dr Odmaa Sambuu (School of Engineering and Applied Sciences, National University of Mongolia)
• 106
Influence of preheating temperature on delta-ferrite formation and mechanical properties of 12%Cr steel weld metals

A 12% Cr ferritic/martensitic (F/M) steel, HT-9, has been used as a primary core material for nuclear reactors. Welding is inevitably used in the nuclear structure design. Fusion welding processes such as gas tungsten arc welding (GTAW) were broadly applied for the F/M steel components. Unfortunately, in the fusion zone of F/M steel weldment, delta (δ)-ferrite is frequently produced, which is considered to decay the creep strength and impact toughness.
In this work, the fusion-zone microstructure and mechanical properties of gas tungsten arc butt welded joints of HT-9 have been explored. The fusion zone solidified with the transformation from the liquid into δ-ferrite, followed by a solid-state transformation from the δ-ferrite to the γ phase. The cooled microstructure was formed by a fresh martensite matrix and retained δ-ferrite. Two kinds of retained δ-ferrite with were observed: one was rich in Cr and depleted in C; the other one was rich in Cr、C、Mo and V. The area fractions of retained δ-ferrite decreased when a preheating treatment at 95 ℃ was applied, moreover the number and mean size of the retained δ-ferrite decreased significantly due to the preheating effect. It indicated that a slower cooling rate (e.g. conducting a preheating treatment) provided more time for the δ → γ reaction and therefore reduced the retained δ-ferrite. After the welding process, a tempered treatment was conducted with 760℃/1h. The impact toughness test for the welds in the as-tempered condition showed that the ductile-brittle transaction temperature (DBTT) of the welds with preheating treatment was lower than that without preheating treatment. The improvement of impact toughness in the welds with preheating treatment was due to the less δ-ferrite in the matrix.

Speaker: Dr Dong Wu
• 107
Irradiation Effects of T91 Ferritic/Martensitic Steel

The harsh neutron irradiation environments in the core region of fast breeder reactors (FBRs) pose a unique challenge for cladding materials. Microchemistry and Microstructural changes resulting from displacement damage and creep rupture are anticipated for structural materials after extended neutron irradiation. Various irradiation effects on the service performance of cladding materials need to be understood. Because of their excellent thermal conductivity and irradiation resistance, ferritic/martensitic steels (FMS) with 9-12%Cr are considered the candidate cladding materials for the new generation FBRs. T91 with the nominal chemistry of 9%Cr-1%Mo-0.2%V-0.08%Nb-0.05%N-0.1%C has been used extensively for fossil power plants because of its excellent creep resistance up to approximately 600℃. As a candidate material for cladding, the fundamental of irradiation damage and its effects on T91 are reviewed in this paper. The objective is to provide a insight of neutron irradiation damage, microstructural and microchemical changes, mechanical properties and facture behavior of T91. The chemistry modification of T91 are also discussed for the improvement of fracture toughness before and after irradiation. This review also provide some suggestions of additional investigations for the application of T91 in FBRs.

Speaker: Mr Junhong LI (China Institute of Atomic Energy)
• 108
Irradiation-Thermo-Mechanical Coupling Analysis and Calculation of Fast Neutron Oxide Fuel Element

During their service life, fast neutron reactor oxide fuel elements operate at high neutron fluence rate, high linear power, and high temperature environment. The fuel elements will exhibit complex irradiation -thermo-mechanical coupling characteristics. The stress-strain analysis calculation is very important for the design of fast reactor fuel elements. This article focuses on the irradiation mechanical phenomenon of the oxide fuel element in the fast neutron reactor. The finite element method is used to consider irradiation creep, irradiation swelling, irradiation hardening, and short-term plastic strengthening. Nonlinear calculation of coupling irradiation-thermo-mechanical is achieved, and the correctness of the calculation method is verified through typical examples. The results show that the non-linear calculation method is correct, and can realize the analysis and prediction of the stress-strain evolution behavior of the fast neutron oxide fuel element during its entire life. The research results in this paper lay a good foundation for the design and evaluation of high linear power fast neutron oxide fuel elements.

Speaker: Mr Qidong CHEN (+86(010)69359493)
• 109
NON-DESTRUCTIVE METHOD FOR DETERMINING STEEL CORROSION COEFFICIENTS IN LEAD

Three methods are recommended for determining the corrosion coefficients of steels in liquid metal (LM).
The essence of the first method, the so-called metallographic method is to measure the thickness of oxides and study their structure and composition. The main advantage of the method is its relative simplicity, as there is no need to remove any residues of LM or oxides from the specimen, but only preparation of metallographic specimens is required. Shortcomings of the method are its location, the need to know the phase composition and density of oxides for the calculations to determine the corrosion depth in the metal and the inability to determine the removal of corrosion products (СP) of steels in LM. The method is applicable only if there is no shearing of the oxides from the sample or its dissociation during corrosion tests.
The second method, the so-called gravimetric method is based on determining the sample weight loss after corrosion tests. In this case, it is necessary to remove the residues of LM and oxides completely from the specimen. The main advantage of this method is the ability to determine all corrosion coefficients, such as total corrosion losses of metal, depth of corrosion, thickness of the oxides and СP removal to LM. Shortcomings of the method are the need to remove СP completely from the specimen and the need to select chemical pickle solutions that have minimal impact on the metal.
The third method is based on metal thickness measurement of samples before and after corrosion tests as well as the thickness of oxides. The method requires the preparation of metallographic specimens on the samples after testing. This method has the same advantages as the first method. In addition, the third method makes it possible to determine corrosion coefficients and measure the depth of internal oxidation layers and the depth of local types of corrosion.
All these methods are destructible, leading to irreversible destruction of samples under investigation.
This paper analyzes the methodical possibilities of non-destructive method for determining steel corrosion coefficients in Pb, based on the determination of weight and density of samples before and after corrosion tests, and allows the determination of all parameters of steels corrosion. Applicability of the method was tested on Fe12Cr1Si steel samples subjected to corrosion tests in a lead flow at a rate of 0.2–0.3 m/s, containing oxygen within (0.3–1.6)·10-6 weight %, at (490±10) оС during 895 h.

Speaker: Mr Oleg Golosov (Research institute of nuclear materials, (INM JSC))
• 110
ON MEASUREMENT OF OXYGEN CONCENTRATION IN SODIUM BY MEANS OF PLUG INDICATOR

Possibility of measurement of oxygen concentration in sodium discussed in the papier. Usually by means of the plug indicator “plugging temperature” is measured and supposed it is equal saturation temperature of sodium with impurities. Then, concentration of oxygen is determined using equation for solubility of oxygen in sodium. As a rule, Noden´s equation is used. But solubility of oxygen in sodium depends on presence of other impurities in sodium. So, many equations for solubility of oxygen in sodium exist which are proposed by different authors at different conditions. Now 19 equations are known. Difference of these equations arise ten times and more at the plugging temperature 250-110°C. Authors of the papier have compared these equations and found out that almost all equations comes together at the temperature 558K and gives logarithm of concentration 1,957. As soon as, this point satisfies all equations it can be used for calibration plug indicator and to get equation solubility of oxygen in sodium adequate to real facility and conditions. Using the equation allows to measure oxygen concentration correct.

• 111
The working capacity analysis of boron carbide after two-year operation as an emergency protection material of the fast reactor

The state of high enriched В4С (industrial and re-fabricated quality) consisting of emergency protection elements, irradiated up to neutron fluence 7x1022 cm-2 (Е>0.1 МeV), was investigated. The maximum burnup of an 10В isotope at two-year irradiation period has made 25 %. That burnup decreases on diameter and height of an absorber column. High local mechanical deformation of emergency protection element metal covers is found out in a maximum fluence zone. It is caused by swelling of neutron absorber material. At the same time, the metal keeps high values of durability and plasticity. Parameters of an irradiated industrial В4С crystal lattice ware amount to а=0.5540 nm and с=1.1835 nm. The halfwidth of X ray reflexes and level of micropressure have increased. According to hydrostatics weighing and electron microscopy, the swelling of industrial and re-fabricated В4С reaches 12 and 18 % respectively only in the bottom of absorbing elements at maximum fluence. Other parts of absorber column with smaller fluence and burnup swell slightly. Volumetric changes are caused by growth of closed porosity. The maximum swelling of irradiated industrial and re-manufactured materials was 25 and 30% respectively. The same initial values were about 19 and 23%. The value of closed porosity from industrial and re-fabricated B4C after irradiation was 14 and 21% versus the initial values of 4 and 9% respectively. Detected pores with a diameter more than 100 nm settle at the grain border. The size and concentration of pores increase with growth of neutron fluence and burnup. Pores cause the formation of microcracks between the boron carbide grains. The open porosity inside intensive swelling and mechanical interaction zone of absorber with the cladding is reduced. The forced fracture nature of the B4C changes from mixed nature with a predominance of intragranular splitting to a purely intergranular type. Functional working capacity of enriched В4С remains during reached burnup as well as after re-fabrication.

Speaker: Mr Evgenii Kinev (JSC «INM»)
• 112
Analysis of sodium fire accident after upgrade of ventilation system of primary loop’s corridor

CEFR is a fast neutron experimental reactor with the coolant of liquid metal sodium. The primary loop’s corridor of sodium purification system, as we know the name of room 309/1, is one of the sodium aerosol containment chambers, where laid out with many sodium pipes of high temperature, especially the pipe with high-level radioactive sodium inside which is come from reactor main vessel. In the event that this pipe is broken, the spray of hot sodium can burn and then produce sodium fire, under certain conditions, a kind of continuous spray sodium fire, which affects the pressure and gas temperature of room and the temperature of internal structure etc a lot. In order to lower the temperature of room 309/1, the ventilation system of this room is upgraded by means of increasing the air intake volume and exhaust air volume, and this kind of upgrade should affect thermodynamic results of sodium fire accident. FEUMIX is used to calculate the results of spray sodium fire accident of the primary loop’s corridor after upgrade of ventilation system, and the results is compared with the results of the same room before the upgrade of ventilation system. It appears that the upgrade of ventilation system of primary loop’s corridor can relieve the consequences of spray sodium fire partly.

Speaker: Mrs jiayin yang
• 113
Development of in-vessel source term evaluation method for ULOF events in sodium-cooled fast reactor

Based on the lessons learned from the TEPCO’s Fukushima Daiichi NPPs’ accidents, evaluation of fission products (FPs) transfer during the severe accidents is quite important in the regulation activities.
In core expansion phase caused by core disruptive accident (CDA) of sodium-cooled fast reactors (SFRs), a large bubble consisting of mixture of core materials and sodium vapor inside, so-called CDA bubble, expands rapidly and accelerates the sodium slug of the upper plenum to the shielding plug to threaten the reactor vessel integrity. If the reactor vessel integrity fails, FPs contained in the coolant and in the cover gas may release into the containment vessel.
To evaluate the FPs transfer in the reactor vessel during core expansion phase, it is important to simulate the FP transfer by coupling with CDA bubble behavior such as vaporization and condensation. Based on this background, S/NRA/R has started developing the method by utilizing the calculation codes: ACTOR and ASTERIA-SFR.
ACTOR [1] has been developed for evaluation of FPs transfer in the primary cooling system during Protected Loss-of-Heat-Sink accident. ACTOR has models to simulate major phenomena: FP release from the fuel pin, bubble transport including break-up and coalescence in sodium, FPs transfer from bubble to coolant, and FP diffusion into the sodium. These analytical models, however, are insufficient to simulate FP transfer of CDA bubble behavior.

ASTERIA-SFR [2] is a CDA analysis code which has capability to calculate interfacial area of CDA bubble considering vaporization and condensation of core materials and sodium, and to calculate transfer of the materials based on mechanistic fluid dynamics calculation. ASTERIA-SFR has potential to simulate not only thermal-hydraulics behavior of CDA bubble but also the FP transfer adequately.
In order to enhance the capability, S/NRA/R has started development of methodology on in-vessel source term evaluation during the ULOF event. The methodology utilizes two computer codes: ACTOR and ASTERIA-SFR. In this framework, ACTOR is incorporated to a plant dynamics analysis code in order to simulate FPs transfer behavior based on more precise thermo-hydraulics calculation. To evaluate FPs transfer behavior with CDA bubble, fluid dynamics calculation module of ASTERIA-SFR is improved to model cesium component.
This paper describes the plan and progress of the development of the in-vessel source term evaluation method. It is also discussed that calculation results of CDA analysis focusing on FPs transfer behavior during the ULOF event.

Speaker: Mr Hiroki Sonoda (Regulatory Standard and Research Department, Secretariat of Nuclear Regulation Authority (S/NRA/R))
• 114
Development of the Simplified Radionuclide Transport (SRT) Code Version 2.0 for Versatile Test Reactor (VTR) Mechanistic Source Term Calculations

The Versatile Test Reactor (VTR) is a fast spectrum test reactor currently being developed in the United States under the direction of the US Department of Energy (USDOE), Office of Nuclear Energy. The mission of the VTR is to enable accelerated testing of advanced reactor fuels and materials required for advanced reactor technologies. The conceptual design of the 300 MWth sodium-cooled metallic-fueled pool-type fast reactor has been led by US National Laboratories in collaboration with General Electric-Hitachi and Bechtel National Inc.

The VTR is utilizing a risk-informed performance-based approach for authorization by the USDOE. As part of this approach, the development of a mechanistic source term (MST), or a realistic evaluation of radionuclide transport and release for specific transient scenarios, is central to developing an accurate representation of reactor risk. A new version of the Argonne National Laboratory Simplified Radionuclide Transport (SRT) code has been developed to support VTR MST analyses.

The SRT code is an integral sodium fast reactor radionuclide transport analysis tool, which assesses radionuclide movement from the reactor fuel to the environment. The code includes models for phenomena associated with radionuclide behavior within the fuel pins, release from failed fuel, migration through the sodium pool, and behavior in the cover gas region and containment. SRT is purposefully designed to facilitate sensitivity and uncertainty analyses with offsite consequences as the metric of interest.

For VTR, version 2.0 of the code was established to fulfill quality assurance requirements associated with the project. These efforts included updating and expanding the suite of verification and validation test cases, extension of the code models for the demonstration of code accuracy, and revisions to code software quality assurance documentation. This paper provides a summary of these activities along with a sample of analysis cases utilizing the new version of the code. The work reported in this summary is the result of studies supporting a VTR conceptual design, cost, and schedule estimate for DOE-NE to make a decision on procurement. As such, it is preliminary.

Speaker: David Grabaskas (Argonne National Laboratory)
• 115
Experimental and Numerical Study on Temperature Fluctuation in The Upper Plenum of Fast Reactor

In the fast reactor, flow fields which have different temperature flow out of heads of assemblies and then mix intensely, which leads to temperature fluctuation. The temperature fluctuation in the fast reactor is very serious due to the thermal characteristic, especially in the upper plenum, which may cause thermal fatigue in the structure of fast reactor. In the published work, simplified parallel jet models were used to investigate the temperature fluctuation, which can not fully reflect the characteristic of temperature fluctuation in the upper plenum of fast reactor. This paper made a more realistic model that includes three heads of assemblies and part of the central column to do further study both numerically and experimentally. The experimental setup was designed and manufactured. The temperature at different heights was obtained while cool and hot fluid mixed in the setup. Meanwhile the same experimental conditions were simulated with Large Eddy Simulation(LES) turbulence model.

Speaker: Mr Yongqi Du (North China Electric Power University)
• 116
Fast Reactor Source Term Modeling and Simulation Functional Requirements and Gap Assessment

A vital part of the licensing process for advanced (non-LWR) nuclear reactor developers in the United States is the assessment of the reactor’s source term. The source term represents the potential release of radionuclides from the reactor system to the environment during normal operations or accident sequences. While historically source term assessment for LWRs has followed a bounding approach with conservative assumptions, a more mechanistic, or realistic, approach to modeling radionuclide transport is being used for advanced reactor systems. As the designs of advanced reactors increase in maturity and move towards licensing, there will be a need to develop modeling and simulation capabilities in analyzing the source term of a prospective reactor concept.
Two types of fast spectrum advanced reactors currently being pursued are liquid metal-cooled fast reactors and fast-spectrum molten salt reactors. The sodium fast reactor (SFR) is the most pursued variant of the liquid-metal cooled reactors, while there is also interest in lead-cooled fast reactors (LFR), which share much of the same source term modeling phenomena. Among salt-fueled molten salt reactors (MSR), there is interest in both chloride salt (Molten Chloride Fast Reactor, MCFR) or fluoride salt (Molten Salt Fast Reactor, MSFR) as the fuel-bearing salt system. There has been much work in the development of the mechanistic source term for SFRs due to the extensive operational experience in the U.S. MSR technology is generally less mature and there is more diversity in the designs being pursued, therefore the development of source term strategies is largely incomplete at present.
This work is meant to be a summary of the phenomena important to modeling the mechanistic source term of both SFRs and MSRs, the functional requirements needed to model those phenomena, and the current state of computational capabilities in fulfilling those requirements. In completing a survey of the current landscape of modeling capabilities with a gap analysis, this work aims to identify the future modeling and simulation development needs.

Speaker: Shayan Shahbazi (Argonne National Laboratory)
• 117
Increase of nuclear power plant hydrogen safety using zirconium accumulator

Ensuring hydrogen safety is one of the most important conditions of the general substantiation of NPP safety.
The greatest danger of the release of a huge amount of hydrogen arises in relation to nuclear stations in water and water stations during the beyond design basis steam-zirconium reaction and in case of leaks between the second (sodium) and third (water) circuits of fast reactors with a sodium coolant.
There are an increasing number of proposals for the use of nuclear power facilities for the production of hydrogen as an alternative source of electricity.
Currently, to prevent an explosive situation, a hydrogen afterburning system is used with catalytic oxidation (combustion) of hydrogen in a gas mixture with using a platinum catalyst.
The authors propose, along with post-combustion of hydrogen or instead, to separate hydrogen from the hydrogen vapor medium using a membrane consisting of several modules of nickel tubes, and to bind it in the form of zirconium hydride using a special structure made of zirconium foil.
The hydrogen storage device is regenerable. By thermal dissociation ZrHx the accumulated hydrogen can be separated and implemented in power devices acting on the basis of converting hydrogen and oxygen into water or steam.
The method is also effective in the field of tritium production for thermonuclear reactors with simultaneous increase in ecological purity of nuclear power plants.

Speaker: Ekaterina Orlova
• 118
Model validation of the ASTERIA-SFR code related to freezing phenomena of liquid and liquid/particle mixtures based on THEFIS experimental results

Mechanical consequence which might be caused by core disruptive accidents (CDAs) is one of the major concerns in fast reactor safety. Once core disruption occurs caused by severe re-criticality, core materials are dispersed azimuthally and radially. The dispersed materials, e.g., liquid/particle mixture of fuel and steel, penetrate into the pin bundles and control rod guide tubes (CRGTs), and then, freeze at the edge of penetration due to heat transfer with surrounding structures. Such freezing phenomena might cause to suppress the negative reactivity feedback of fuel dispersion. When tight blockages are created inside CRGTs due to freezing, discharge of core materials could be impeded and a molten core pool formation could be enhanced. A radial motion of the molten core pool, so-called pool sloshing, is one of major causes to induce a prompt critical condition. Thus, penetration and freezing phenomena of core materials play a key role to govern the mechanical consequence of CDA.
S/NRA/R has been developed ASTERIA-SFR, which was renamed from ASTERIA-FBR, to evaluate the mechanical consequences due to severe re-criticality during CDA. For the sake of adequate simulation of penetration and freezing behavior of molten materials into the pin bundles and/or CRGTs, two major models of CONCORD , thermo-fluid dynamics calculation module of ASTERIA-SFR, were improved: heat transfer model and momentum exchange function model. The former model was modified by implementing fuel caps freezing model considering thermal resistance at the interface between molten material or crust and structure. The latter was modified by implementing particle viscosity model.
This paper describes model validation study of ASTERIA-SFR related to freezing phenomena of liquid or liquid/particle mixtures through the THEFIS test simulation. To reproduce the test results, models of heat transfer and momentum exchange function were improved considering interfacial heat resistance and effective viscosity of liquid/particle mixture. As a result, it was found that the calculated penetration depths of liquid or liquid/particle mixtures were in good agreements with the experimental results. Freezing behavior and uncertainty shown in the calculation results were also discussed.

Speakers: Dr Tomoko Ishizu (Regulatory Standard and Research Department, Secretariat of Nuclear Regulation Authority (S/NRA/R)) , Mr Satoshi Fujita (Regulatory Standard and Research Department, Secretariat of Nuclear Regulation Authority (S/NRA/R))
• 119
Modelling of the in-vessel source term during a hypothetical severe accident in an SFR

The in-vessel source term problem is to determine the partitioning of radionuclide (RN) in fuel, coolant and cover gas following a hypothetical severe accident. The determination of the in-vessel source term is a challenging task, as there is significant uncertainty involved in RN transport. The RN transport mostly controlled by their diffusion characteristics in the fuel, chemical interactions between RNs and coolant, their physical form (for ex., molten, solid, vapor, and aerosol) to mention a few phenomena.

The mechanistic modelling of in-vessel source term considering all phenomenon is a numerically demanding task. In the present work, a thermo-chemical equilibrium based approach is adopted to study the release behaviour of RN to the cover gas. We would consider this approach as the first step towards mechanistic model development for oxide fuelled SFRs. To study the chemical behaviour of RN, a medium-sized oxide fuelled, pool type SFR (1250 MWt) is chosen as a reference design. For the analysis, unprotected loss of flow accident (ULOFA) initiated by loss of flow to both the coolant pumps resulting in whole core melt is considered. The analysis is performed for different temperatures with two mixing assumptions, i.e., no mixture assumption and real mixture assumption. The no mixture assumption essentially means that, during the calculation of the equilibrium species, the solubility properties are not considered. This assumption provides conservative estimates. Whereas, for the real mixture assumption, the excess functions and solubility are considered. With the help of the equilibrium species, the release fractions are evaluated.

From the calculations, we observed that the amount of the oxygen available for the reaction affects the release dynamics of lanthanides and barium-strontium group (For example, Eu and Sr). Additionally, various mixture assumptions do affect the release behaviour of the RN in the coolant. For example, it is found that with no mixture assumption, the release fraction for Cs is about 0.9, when mixing assumption is considered, the release fractions are of the order of 1E-04 – 1E-05.

Speaker: PARTHKUMAR RAJENDRABHAI PATEL (Indira Gandhi Centre for Atmoc Research, India)
• 120
Preliminary Shielding Analysis for the Versatile Test Reactor

The Versatile Test Reactor (VTR) is currently under development by the U.S. Department of Energy (DOE). It will provide very high fast neutron flux irradiation capabilities that are currently unavailable in the U.S. Given the increasingly large number of advanced reactor concepts being pursued in recent years, this irradiation testing capability will be essential to support maturation of these concepts. Radiation protection is an important part of the VTR design. High neutron fluxes can pose a challenge for radiation protection of the structures and equipment near the reactor core. This paper provides a summary on the status of the shielding considerations and analysis performed for VTR. The paper focuses on the shielding needs for the secondary sodium flowing through the Intermediate Heat Exchanger (IHX), the air flowing through the Reactor Vessel Auxiliary Cooling System (RVACS), the dose rate above the reactor head access area, the radiation dose and flux in the neutron detectors, and the radiation damage on the reactor vessel wall and the structures near the core. One key challenge for the shielding design of VTR is the activation of the secondary sodium when it flows through the IHX. The IHXs are placed in the primary sodium pool inside the reactor vessel. During operation, neutrons produced in the core can reach the IHXs due to the small absorption cross section of sodium and the large mean free path of fast neutrons. The secondary sodium gets activated inside the IHX and emits photons in the secondary loop outside of the reactor vessel. Activation of secondary sodium needs to be mitigated in order to meet the radiation dose limit sets for personnel working at the plant as secondary sodium circulates outside of the reactor building to the air-dump heat exchangers. Similarly, for the RVACS, the air flowing in the system gets close to the reactor core, and some argon would become activated and released into the atmosphere. Thus, proper shielding is also required to reduce argon activation in the RVACS. The VTR design and development is underway and shielding considerations discussed in this paper will progress with the rest of the reactor.

The work reported in this summary is the result of studies supporting a VTR conceptual design, cost, and schedule estimate for DOE-NE to make a decision on procurement. As such, it is preliminary.

Speaker: Tingzhou Fei (Argonne National Laboratory)
• 121
SIMULATION OF THE FAST FLUX TEST FACILITY LOSS-OF-FLOW WITHOUT SCRAM ACCIDENT SCENARIO USING THE SAM COMPUTER CODE

A major appeal of sodium fast reactors is their passive safety capabilities. To demonstrate this, the Fast Flux Test Facility (FFTF) conducted a series of Loss of Flow WithOut Scram (LOFWOS) tests at up to 50% power. Experimental results from this test were made available through an IAEA CRP for use in a code benchmarking activity. In this work, the System Analysis Module (SAM) was used to analyze the FFTF during the LOFWOS transient. Efforts were made first to develop a faithful steady state representation of the facility. In this model, flow rates and pump heads were matched to experimental values within 0.4% in each loop. Additionally, the core inlet temperature was matched within 0.02%. At this stage, the results of the transient experiment were not available to CRP participants, creating a blind simulation stage. During this blind stage, a point kinetics model SAM was used to predict the core power, as well as peak fuel temperatures, system flow rates, etc. throughout the LOFWOS transient. Simulation results were compared to experimental results at the end of the blind stage. Results show that the SAM model was able to capture the general trend of the transient, however, the magnitude of the results showed significant deviation from the experimental results. More recently, work has been ongoing to understand the cause of these deviations and improve the model to more accurately represent the transient results.

Speaker: Brent Hollrah (Texas A&M University)
• 122
Study on Sodium Fire PSA Methodology for Pool-Type Sodium cooled Fast Reactor

According to the sodium cooled fast reactor operation experience, about 100 sodium leakage accidents have happened in history. Sodium fire is a typical hazard in SFR, which is also one of the main reasons for its unavailability, and may be one of the main contributors to the total reactor risks. Study on sodium fire PSA methodology can not only quantitatively evaluate the sodium fire risk in SFR, but also identify the weakness of sodium fire prevention and improve the safety of SFR. Based on the specific technique and process of NUREG/CR-6850 LWR fire PSA methodology, considering the design characteristics and sodium fire characteristics of pool type SFR, a sodium fire probability safety assessment methodology for pool type sodium fast reactor is proposed. Then, an example of study on the sodium fire in secondary loop sodium pump room of China Demonstration Fast Reactor CFR600 is presented, which shows the sodium fire event sequence and its frequency of core damage. After that, several key issues which need to be further researched for sodium fire PSA in the future were discussed.

Speaker: Jing WANG (China Institute of Atomic Energy)
• 123
Thermal Hydraulic Simulation of Loss of Flow Without Scram Test in FFTF using DYANA-P code

System dynamic simulation of loss of flow without SCRAM test carried out in FFTF has been carried out using plant dynamics code DYANA-P. DYANA-P has one-dimensional models for various sub-systems of sodium cooled fast reactor. Thermal models are based on heat balance between various sections exchanging heat. Hydraulic model is based on momentum balance between various flow segments in sodium circuits. Pumps are modeled with characteristics derived from generalized homologous characteristics. Neutronic model for the core is based on point kinetics approximation. A similar modeling methodology is adopted in the formulation of computer codes such as DYNAM and SIFDYN developed for performing dynamic calculations in the Fast Breeder Test Reactor (FBTR). These codes have been validated through various tests carried out in FBTR.
In the hydraulics modeling of FFTF, flow evolutions in different coolant loops as well as core channels are modeled through integral momentum balance approach. However, level variations in coolant plenums are not modeled. The reactor core comprising of 80 fuel sub-assemblies is represented through six core channels and decay power in the core is evaluated using an empirical correlation. Various reactivity feedback effects due to (i) radial expansion of core support structure, (ii) control rod drive line expansion, (iii) volumetric expansion of sodium in the core, (iv) axial expansion of cladding, (v) axial expansion of fuel, (vi) Doppler Effect due to changes in fuel temperature, (vii) sodium level changes inside GEM sub-assemblies, etc. are considered.
As a result of primary pump speed reduction, the pump developed head decreases causing a decrease in pump flow rates. The reduction in primary loop flow rates causes GEM sodium levels to decrease resulting in large negative feedback reactivity addition leading to reactor shutdown. The natural circulation flow sets up in the primary circuit. Initially, because of the reduction in flow rate, the sodium outlet temperatures from core channels start increasing. Later, because of the power reduction due to the action of GEM, the temperature starts decreasing. The rate of power reduction is less than the rate of flow reduction. Hence, a second temperature peak appears. Further, the temperatures decrease continuously as the decay heat decreases continuously, and also the flow rate has become constant. No significant temperature rise is predicted in any reactor component (maximum clad temperature predicted is ~100 °C above its nominal value for a few seconds). Continuous decay heat removal from core to DHX is established through IHX.

Speaker: Vikram Govindarajan (IGCAR)
• 2.2 Safety Design and Analysis

Chairs: Marina Demeshko and Yamano Hidemasa

• 124
Overview of the Versatile Test Reactor Safety Analysis

The Versatile Test Reactor (VTR) is a fast spectrum test reactor currently being developed in the United States under the direction of the US Department of Energy, Office of Nuclear Energy. The mission of the VTR is to enable accelerated testing of advanced reactor fuels and materials required for advanced reactor technologies. The conceptual design of the 300 MWth sodium-cooled metallic-fueled pool-type fast reactor has been led by US National Laboratories in collaboration with General Electric-Hitachi and Bechtel National Inc.
Safety analysis of the conceptual VTR design has been performed using the SAS4A/SASSYS-1 fast reactor safety analysis code with a model representing the reactor core, primary and intermediate heat transport systems, reactor vessel auxiliary cooling system (RVACS), and reactor protection system (RPS). A number of protected transients have been evaluated to demonstrate the system’s response to various initiating events. The present analysis addresses several transients that are expected to be bounding accident scenarios representing the three key ways to perturb a reactor: through changes to the mass flow rate, reactivity, or core inlet temperature. These correspond to a loss of flow (LOF) or station blackout (SBO), transient overpower (TOP), loss of heat sink (LOHS), respectively.
At the current stage of design, transient simulation results for the Versatile Test Reactor indicate that large safety margins exist for many event initiators. The RPS response following the detection of elevated plant conditions dominates the transient behaviors in these transients. Because the primary heat transport system is able to transition quickly and effectively to natural circulation and because the RVACS, which is responsible for decay heat removal, provides sufficient heat rejection, large margins for all criteria were predicted for the transients.
The work reported in this summary is the result of studies supporting a VTR conceptual design, cost, and schedule estimate for DOE-NE to make a decision on procurement. As such, it is preliminary.

Speaker: Tyler Sumner (Argonne National Laboratory)
• 125
Integrating safety at the first design stages: a new methodology for safety-oriented SFR core design

Within the framework of the Generation IV Sodium-cooled Fast Reactor (SFR) R&D program of CEA (French Alternative Energies and Atomic Energy Commision), a methodology is proposed to enable the early consideration of safety in the undergoing design process. Before the use of mechanistic tools (CATHARE, SIMMER, APOLLO, etc.) that usually requires an advanced knowledge of the reactor design, this approach involves several physical tools dedicated to each physical phenomenon likely to govern the choice of design parameters. These tools are mostly based on simplified modelings (mostly 0D and 1D) and are carefully validated versus experimental results. They are gathered in a platform that covers all kind of accidental phenomenology, from the initiator (pump trip, reactivity insertion, local blockages, etc.) until the reach of a stable and coolable state after corium relocation. Thus, enabling a large number of simulations in a reasonable computational time, it makes possible the characterization of some major accident transient bifurcations (such as boiling onset, boiling stabilization, primary power excursion, molten fuel vaporization, corium axial relocation in transfer tubes, etc.), in terms of probability of occurrence and of consequences on the transient evolution. It also enables to identify the main physical parameters causing the bifurcations in order to allow a straight feedback on the core design and to give some orientation for future R&D studies. In this paper, a focus is firstly made on some scenario bifurcations to illustrate the platform abilities. The boiling onset and possible stabilization are studied. The possibility of power primary excursion depending on the core design is also addressed, and the fuel vaporization possibilities are finally assessed. Depending on these results, a global estimation of the core safety features is eventually provided for both prevention and mitigation features, at a quite early design stage. This allows a rapid comparison between core concepts or design options in order to facilitate the sketch of incoming safety-oriented SFRs.

Speaker: Dr Jean-Baptiste Droin (CEA Cadarache)
• 126
Thermal hydraulic assessment of the performance of secondary sodium system based decay heat removal circuit

To improve the safety of future sodium cooled fast reactor, alternative Decay Heat Removal (DHR) path is envisaged through secondary sodium main circuit. Secondary Sodium Decay Heat Removal System (SSDHRS) transfers heat from secondary sodium to ambient through Air Heat Exchanger (AHX). SSDHRS is planned to operate in forced circulation mode using Class 3 power supply. SSDHRS in addition to Safety Grade Decay Heat Removal System help in demonstrating failure of DHR function is highly unlikely, so as to be regarded as practically eliminated.

Analysis of SSDHR System is carried out using system dynamics code Flownex. Flownex is a general purpose one dimensional code for thermal hydraulic simulation of systems with component level modeling capabilities. Different components of the system, viz., Intermediate heat exchangers, AHX, stack, secondary sodium Pump (SSP), blower are modeled through appropriate component models. Single zone model of core is also developed to represent the decay heat source. Heat transfer capacity of each SSDHR system is found to be 15.17 MW at 544 °C temperature of hot pool sodium. SSP develops 26 % of nominal flow during SSDHRS operation, out of which 4% flows through AHX and the remaining flow is bypassed through a separate bypass path. Flow rates of sodium and air in AHX are estimated as 133 kg/s and 59.5 kg/s respectively. Parametric studies have been carried out by varying hot pool temperature in the range of 200 °C to 650 °C and primary sodium flow rate, and their effect on performance of SSDHRS is studied. Transient analysis of 'off-site power failure' event is carried out and the predicted hot pool and cold pool temperatures are found to be within the design safety limits. Further study is carried out to assess performance of SSDHRS during natural circulation of secondary sodium and air. Heat removal capacity at 544 °C and 650 °C primary sodium temperature is found to be 8.78 MW and 11.07 MW respectively. Study is also carried out to assess the performance of SSDHRS, when all the three circuits are under natural circulation mode. The maximum heat removal capacity is found to be 9.775 MW when the hot pool temperature reaches 650 °C.

Speaker: anurag samantara (Indira Gandhi Centre for Atomic Research)
• 127
PRE-DESIGN OF A PASSIVE DECAY HEAT REMOVAL SYSTEM WITH A PHASE CHANGE MATERIAL FOR SMR-SFR

The Fukushima accident highlighted the need to improve the design of the safety systems in order to cope the consequences of a full blackout accident. In particular, the decay heat removal function must always be ensured in order to guarantee the integrity of the first barrier and the second one confining the radioactivity. The paper shows the main features of the pre-design of an innovative ex-vessel Decay Heat Removal System (DHRS) designed for a SMR-SFR of a core power of 400 MWth. This safety system allows the heat removal in a fully passive way, only by the radiation of the outer surface of the primary vessel. The radiated heat is extracted by the DHRS through the enhanced natural circulation of a liquid metal flowing inside a bundle of tubes outside the vessel. The final heat sink of the DHRS is another important aspect of innovation. The aim is to diversify it, improving the safety function ensured by a standard water-cooled or air-cooled heat sink. For this purpose, the use of a Phase Change Material (PCM) has been chosen inside the heat sink in order to provide a supplementary thermal inertia, which contributes to improve the cooling capacity of the DHRS. Moreover, a PCM contributes to improve the compactness of the final heat sink, allowing its storage inside a dedicated building. This aspect allows this component to be protected by external aggressions, leading to a significant improvement the safety of the Nuclear Power Plant. The actual design of the ex-vessel DHRS has been applied to the new SFR sketch ATRIUM, in order to guarantee a fully passive cooling of the reactor core for a duration of three or more days.

Speaker: Mr Alessandro Pantano (CEA)
• 128
Development of the Versatile Test Reactor (VTR) Probabilistic Risk Assessment

The Versatile Test Reactor (VTR) is a fast spectrum test reactor currently being developed in the United States under the direction of the US Department of Energy (USDOE), Office of Nuclear Energy. The mission of the VTR is to enable accelerated testing of advanced reactor fuels and materials required for advanced reactor technologies. The conceptual design of the 300 MWth sodium-cooled metallic-fueled pool-type fast reactor has been led by US National Laboratories in collaboration with General Electric-Hitachi and Bechtel National Inc.

The VTR is utilizing a risk-informed performance-based approach for authorization by the USDOE, derived from recent efforts by the US industry led Licensing Modernization Project (LMP). As part of this methodology, the probabilistic risk assessment (PRA) is a key input into decisions regarding the identification and selection of safety basis events, the safety classification of structures, systems, and components, and the evaluation of the adequacy of defense-in-depth. This paper provides an overview of key factors in the development of the VTR PRA, including applicable USDOE and industry PRA standards, the risk metrics and criteria to be utilized for risk-informed decision-making, and the selected structure of the PRA technical elements. The work reported in this summary is the result of studies supporting a VTR conceptual design, cost, and schedule estimate for DOE-NE to make a decision on procurement. As such, it is preliminary.

Speaker: Mr Jason Andrus (Idaho National Laboratory)
• 129
ANALYSIS OF THE SGTR ACCIDENT FOR SAFETY JUSTIFICATION OF TWO-CIRCUIT LEAD COOLED REACTOR.

Lead cooled reactor BREST-OD-300 design is under development as a part of Russian federal project "PRORYV". Two circuits are used for heat removal from the reactor. The special feature of two-circuit heat removal system is the potential risk of steam ingress into the core in case of large leak in the steam generator as a result, for example, of steam generator tube rupture (SGTR). The main concern is caused by the possibility of positive reactivity insertion in case of steam ingress into the central section of the core because this can have an impact on nuclear safety.
The analysis of physical phenomena that are important for correct prediction of SGTR accident consequences was performed and it was concluded that for correct modeling of the transient in the core in the case of steam injection the following procedures are required:
a) Accurate modeling of neutronics and thermal-hydraulics taking into account their coupling;
b) 3D modeling of the accident considering strong non-symmetry of spatial distribution of concentration of steam entering the reactor core during the accident.
3D multi-physics (neutronics + thermal-hydraulics) UNICO-2F code was developed for studying SGTR accident. The code is capable of calculating transient 3D spatial distributions of coolant velocity, pressure and temperature, as well as steam concentration and power density in the core.
The analysis of BREST-OD-300 reactor parameters under SGTR accident conditions was performed and it was shown that even for the most conservative scenario of the accident maximum (during the transient) fuel pin cladding temperature was kept within permissible limits. Therefore, the self-protection of BREST-OD-300 reactor in case of SGTR accident was confirmed.

Speaker: Mr IURII SHVETSOV (JSC "PRORYV")
• 130
Design Studies Towards Raising FBTR to Full Power

Fast Breeder Test Reactor (FBTR) is a loop type sodium cooled fast reactor, operating at Kalpakkam. FBTR core is originally designed to operate at 40 MWt using mixed oxide (MOX) fuel with 30% PuO2 and 70% UO2 (85% enriched U). However, due to the non-availability of enriched uranium, mixed carbide fuel (Mark-I (70% PuC+30%UC)) was chosen for the initial core. After first criticality in 1985, the initial core had operated with a maximum power of 10.6 MWt. Subsequent operations have been carried out by adding many variants of carbide and oxide fuel sub-assemblies (FSAs) and increasing the operating linear heat rating (LHR) in stages, based on the encouraging post irradiation examination (PIE) results. The plant is operating at the power level of 32 MWt during its 29th campaign.
It is planned to raise the power of FBTR to its design target power of 40 MWt by using Mark-I subassemblies (SAs). The transition to 40 MWt core is proposed to be taken up during the 30th campaign. The envisaged core will have 70 fuel SAs and the peak LHR will remain restricted at 400 W/cm. In order to ensure a minimum shutdown margin of 4200 pcm as per the technical specification for operation, four poison SAs (with 50% B-10 enrichment) are added in the second ring along with existing 6 control rod SAs (B-10 enrichment of 90 %) provided in the 4th ring. The core design studies have been completed.
Safety studies of the 40 MWt core have been carried out and perturbation worths & kinetic parameters have been estimated. Hypothetical Core Disruptive Accidents (HCDA) have been analysed for ULOFA, ULOCA and UTOPA. For this core, 12 MJ can be considered as the maximum possible mechanical energy release under HCDA. In order to demonstrate the inherent safety characteristics and the capability of plant protection system with respect to various plant transients, analyses of various enveloping design basis events have been carried out using the plant dynamics code DYNAM and safety is demonstrated. Detectable and permissible flow reduction for different SAs has been estimated and found to be safe. Shielding analysis show that in general, there is an increase in neutron and gamma fluxes at various locations of core and shield regions with respect to the 32MWt core. This paper summarises the studies carried out towards raising the power of FBTR to 40 MWt.

Speaker: RAGHUPATHY S. (Indira Gandhi Centre for Atomic Research, Kalpakkam)
• 131
Modelling of postulated reactivity insertion in a Generation IV Molten Salt Reactor

The MSFR (Molten Salt Fast Reactor) consists in a concept of high power molten salt reactor. In molten salt reactors, the fissile and fertile nuclei are dissolved in a circulating salt that acts as fuel and coolant. The physical state of the fuel permits to consider draining as a way to mitigate hypothetical accidents. Contrary to solid fuel FNR (Fast Neutron Reactor) concepts, in the MSFR, the fuel is nearly in its most compact geometry. That is why a large-scale compaction cannot occur in these cores. These concepts can operate and are studied in the Th/U cycle with a fluoride salt or in the U/Pu cycle with a chloride salt. In the fluoride reference concept, the global temperature reactivity feedback is around -8pcm/K. This strong negative feedback, that guarenties an excellent intrinsic stability of the core, takes into account two effects: Doppler Effect and salt density effect. Each one being approximately half of the total value.

The goal of this work is to study the MSFR behaviour in case of a postulated reactivity insertion. In order to evaluate the consequences of extremes reactivity insertions, the first study concentrates on slow reactivity insertions to verify the efficiency of the draining of the core. Because of the presence of the expansion tank, the beginning of the draining has no impact on the reactivity in the core. Then, we also want to verify that there are wide safety margins between the consequences of plausible reactivity insertion and the consequences of extreme postulated reactivity insertions.

In the case of extreme reactivity insertions, during the early stages of the transient, the salt cannot expand freely and goes out of the core. The pressure increases and the bubbles inside the salt collapse.

To make these studies, we are developing two independent modelling. The first one is being developed in order to study slow reactivity insertion. The second code aims at calculating fast explosion of the vapour formed in the salt in case of fast reactivity insertions.

When the temperature of the salt rises, some vapour could be formed in the salt due to some fission products or because of the vaporization of the salt itself. The vaporisation of the salt could then lead to a quick expansion of the vaporized fluid. Our ultimate goal is the chaining of our codes to calculate any kind of accidental scenarios that could lead to successive recriticalities.

Speaker: Thibault LE MEUTE (CEA Cadarache - LPSC-IN2P3-CNRS)
• 3.1 Fuel Cycle Scenarios

Chairs: Alina Constantin

• 132
The initial stage of closing the NFC of two-component nuclear power. Challenges and solutions

When transferring to two-component nuclear energy system at the initial stage of closing the nuclear fuel cycle of Russia, the number of fast breeder reactors BN type fuelled with plutonium do exist to a small extent and stocks of separated plutonium already are sizable and continue to grow due to pilot reprocessing SNF of VVER. Therefore, for this period of time, it is proposed to abandon reprocessing the spent fuel from the BN reactors, i.e. to operate in the "open in plutonium" cycle with storage of this spent nuclear fuel. The results of scenario studies show that the amount of plutonium extracted from VVER SNF is sufficient for commissioning and continuing operation of a small series of BN reactors. In addition, it is possible to eliminate a radial reproducing blanket in a fast reactor. According to our estimates, it is worthwhile to start full-scale reprocessing of BN SNF closer to the middle of this century, when we can expect the quantity of BN SNF to be sufficient to reduce significantly the reprocessing unit cost. This approach will reduce the unit fuel cost of electricity production by ~25%, while reducing the total cost of production to 5%.

Speaker: Mr Andrey Gulevich (IPPE JSC)
• 133
Presentation of the new European project PUMMA devoted to Plutonium management in the whole fuel cycle

The European project PUMMA (Plutonium Management for More Agility) is dedicated to the different Pu management options in 4th generation systems to assess the impact on the entire fuel cycle. Fast neutron reactors with associated fuel cycle strategies have been chosen to cope with these options because they are flexible: they offer the possibility of isogeneration, burning or breeding of plutonium.
The fuel cycle scenarios associated with the different strategies will be evaluated. The behaviour of MOX fuel with Pu contents of 45% will be studied experimentally through the characterizations of fuels from three irradiations carried out under nominal conditions (in MTR and SFR) and incidental (in MTR). PUMMA will provide additional results on the ther-mo-mechanical properties of this fuel covering the full range of composition and effect of irradiation. These studies will be supplemented by dissolution tests on spent fuels with high Pu contents because to date, the studies have been lim-ited to concentrations below 30%.
The construction of this project was carried out with complementarity between the disciplines of the fuel cycle and with a close exchange between simulation and experimental verification for each of the fields: fuel behaviour under irradia-tion, material properties, spent fuel dissolution and partitioning. PUMMA will be the link between Europe and other international organizations: fuel cycle studies at IAEA and OECD, GEN-IV systems at ESNII and GIF, studies on fuel materials at OECD.
Another objective is to maintain the expertise and skills on the management of plutonium in Europe involving the young generation of researchers with experts who have contributed to these projects for over 20 years.
22 participants will contribute to this project with a total budget of around 7M €.
PUMMA starts in October 2020.

Speaker: nathalie chauvin (CEA)
• 134
Fuel cycle closure for high power fast neutron reactor

The Russian Federation is developing a number of technologies within the «Proryv» project for closing the nuclear fuel cycle utilizing mixed (U-Pu-MA) nitride fuel. Key objectives of the project include improving fast reactor nuclear safety by minimizing reactivity changes during fuel operating period and improving radiological and environmental fuel cycle safety through Pu multi-recycling and MA transmutation.
This advanced technology is expected to allow operating the reactor in an equilibrium cycle with a breeding ratio equaling approximately 1 with stable reactivity and fuel isotopic composition. Nevertheless, to reach this state the reactor must still operate in an initial transient state for a lengthy period (over 10 years) of time, which requires implementing special measures concerning reactivity control.
The results obtained from calculations show the possibility of achieving a synergetic effect from combining two objectives. To sustain the required reactivity margin on burnup the initial fuel loading includes MA from thermal reactor spent fuel. This should be combined with using reactivity compensators in the first fuel micro-campaigns. The work also considers various options of the core including with axial layer.
The work presents the findings obtained from modeling the entire lifecycle of a 1200 MWe fast reactor, the transition to an equilibrium state and the changes occurring in spent nuclear fuel nuclide and isotopic composition. The work also demonstrates the possibility of completely utilizing MA from thermal and fast reactor spent fuel in next generation FRs without the need of special actinide burners.

Speaker: Elena Rodina (JSC "Proryv")
• 135
Perspectives and discussions on the modes and development path of China's commercial closed nuclear fuel cycle

China already has 48 nuclear power reactors in operation, and 12 nuclear power reactors under construction till June，2020. The accumulated spent fuel in China has exceeded 8,500 tons. China implements the established policy of closed nuclear fuel cycle for the sustainable development of nuclear power. However, there seems no feasible development plan and road map to initiate and deploy a commercial closed fuel cycle in China up to now.
If the commercial closed nuclear fuel cycle would to be truly implemented in China, then how to initiate and deploy the commercial closed fuel cycle, and how to determine the feasible development path are the issues worthy of serious consideration for the Chinese nuclear industry. Unlike pure fuel cycle academic research, commercial closed nuclear fuel cycles need considering industrial and engineering feasibility as well as project economy in addition to technical feasibility aspects.
Large scale and commercial fast reactors and their fast fuel cycles are still far away, and the industrialization of the nuclear fuel cycle requires gradual and phased progress. Most of the operating nuclear power plants in China are PWR units. Since China implements and promotes a commercial closed nuclear fuel cycle, it cannot avoid the major issue of whether it is necessary to start a closed fuel cycle in the mature commercial PWRs. On the contrary, this issue needs to be seriously considered.
Different from the implementation of closed nuclear fuel cycle reactors in countries such as France and Russia, the operating status and modes of PWRs in our country are varied, and some have implemented different plant modifications such as reactor power, core design and fuel improvements, with different fuel types, different burn-ups, different cycle lengths, and different safety margins. These characteristics and differences bring challenges and difficulties to the implementation of a closed nuclear fuel cycle.
Based on international experiences and China’s situation, this paper discusses the necessity of initiating a closed nuclear fuel cycle from mature commercial nuclear power plants in China, as the initial stage of the closed fuel cycle to lay the foundation for the future advanced nuclear fuel cycle. Analyze and discuss the launch mode of China's commercial closed nuclear fuel cycle and spent fuel reprocessing plants, review the nuclear fuel types to be utilized in the closed nuclear fuel cycle, and discuss the possible configuration and development path of China's closed fuel cycle in the future.

Speaker: Mr MIN XIAO (China Nuclear Power Technology Research Institute)
• 136
Potential Role of Fast Reactors with Heterogeneous Fuel Assembly in Development Nuclear Power Structure

The current stage of nuclear power development is often characterized by directly opposite trends: on the one hand, there is a growing understanding of the need to transition to a closed fuel cycle, and on the other hand, there is a very large uncertainty about the growth rates of nuclear power capacity. In such conditions, the main driver of the transition to SNF is not the need for capacity building and accelerated commissioning of fast reactors for extended fuel reproduction, but the need to reduce the amount of SNF accumulated in storage. In other words, at the transition stage of NPS development as a system, the main problem is to reduce the volume of SNF in VVER reactors by reprocessing. In conditions where it is not necessary to immediately put into operation a large number of fast reactors, the problem of involving dense fuel in the fuel cycle can be solved in a simpler way. Certain changes can be made to the FA of fast reactors, which can provide sufficient operating conditions for poorly justified fuels and provide the necessary energy processing, as well as the reactivity reserve.
As shown by the research carried out at NRC KI, this problem can be solved by using a fast reactor fuel assembly with intra-cassette heterogeneity, which contains two types of fuel elements: MOX fuel with a high plutonium share, and reproducing fuel rods made of dense fuel, for example, depleted uranium metal. The challenge of ensuring the reactivity and energy producing is assigned to the fuel elements made of MOX fuel, and the problem of compensation of reactivity and breeding is assigned to the metal fuel rods with depleted uranium. At the end of the campaign the fall of energy production of MOX-fueled rods is compensated by the growth of energy production of the breeding rods. Thus, the dense fuel located in the breeding rods reaches the limits of its operability only before unloading from the core.
To assess the capabilities of a fast reactor with heterogeneous fuel assembly in a two-component system of the Russian nuclear power industry in present paper material balances of fissile nuclides were calculated. For this purpose, three scenarios were considered: one with the traditional BN-1200 core layout, and two scenarios with heterogeneous fuel assembly of BN-1200. One of them provided for joint reprocessing of spent MOX-fueled rods and raw fuel rods, and the second – separate.

Speaker: Yaroslav Kotov (National Recearch Centre "Kurchatov Institute")
• 137
Reference Fuel Options for Generation-IV Sodium-cooled Fast Reactors

Within the Generation IV International Forum, the partners of the Sodium-cooled Fast Reactor System Arrangement (China, Euratom, France, Japan, Korea, the Russian Federation, The United Kingdom and the United States of America) completed an evaluation of SFR advanced fuel options. This work was based on a preliminary work performed in the GIF SFR Advanced Fuel Project. It entailed a comparison of the oxide, metal, and nitride fuel types with respect to fuel fabrication processes and fuel performances, to identify advanced fuel candidates for different applications. Additional R&D efforts were also focused on the minor actinide bearing fuels and high burnup capability evaluation.
A brief history of the use of these fuel is first given in the paper, including their major advantages regarding their use for SFRs. This is followed by a description of the SFR fuels elements for each type of fuel, and design choices explained based on fuels properties.
In the second section of the paper, the roles of the fuels in Safety Cases are presented and discussed. To achieve the fundamental safety functions, the basic SFR fuel design requirements include prevention of fuel failures, maintaining a coolable geometry, and keeping the control/safety rod/element injection channels open.
In the third part of the paper, the challenges to Fuel Qualification are presented. Irradiation and safety testing experience and data exists for standard oxide and metal fuel types, with their performance and reliability proven for burnup levels much higher than the LWR fuel, and for nitride fuel extensively studied in the Russian Federation. However, this is not the case for all fuel and cladding systems. This section ends with a summary of the current status and future challenges for qualification of each fuel type and cladding.
To conclude, it can be stated that, in general, the acquisition of fuel performance data is the most common and effective method to qualify the integrity of fuel pins and fuel subassemblies. A variety of irradiation experiments for SFR oxide, metal, and nitride fuels were identified. Yet, fast spectrum irradiation capabilities are very limited internationally, and fuel testing campaigns can require a great deal of time, effort and expense. Thus, advanced fuel performance modeling techniques simulating the fuel irradiation behavior in the reactor may play a more significant role in future fuel qualification, with the main challenge being to validate the predictability of the complex fuel performance phenomena identified for each fuel type.

Speaker: Frédéric SERRE (Commissariat à l’Energie Atomique et aux Energies Alternatives (CEA))
• 138
The influence of isotopic composition of plutonium in fast reactor fuel on the reactivity margin

In Russia, it is planned to build a pilot demonstration fast lead-cooled reactor with an electric power of 300 MW (BREST-OD-300). Mixed nitride U-Pu fuel with minor actinides (MA) of energy composition will be used as nuclear fuel. Plutonium will be taken from the spent nuclear fuel (SNF) of a VVER-440. The isotopic composition of power plutonium can be different due to different storage periods and different burnup rates in spent nuclear fuel. It is practically unattainable to obtain a uniform isotopic composition.
The plutonium in the storage has a different content of the Pu239 isotope with a difference of more than 15%. This difference will affect the neutronic characteristics of the fuel in the reactor.
The paper considers the influence of fuel compositions with different composition of plutonium on the neutronic characteristics of the fuel and the equability of the BREST reactor operation.
In the BREST reactor, the equilibrium mode of the reactor operation requires a change in reactivity during fuel burnup within the effective fraction of delayed neutrons between refuelings, taking into account all processes accompanying the operation of the reactor.

Speaker: Yulia Karazhelevskaia
• 139
RADIATION AND HYGIENE ASSESSMENT OF EXTERNAL EXPOSURE FACTORS OF PERSONNEL WORKING AT EXPERIMENTAL FACILITIES IN THE PRODUCTION OF MIXED NITRIDE URANIUM-PLUTONIUM FUEL

The purpose of this work was to study and identify the features of the formation of external radiation doses to personnel during the production of a new mixed uranium-plutonium nitride fuel, which surpasses MOX fuel in almost all thermal and physical characteristics, the production of which is currently being developed in Russia.
Quite a few works have been devoted to the radiation characteristics of this fuel. This paper presents the results of a radiation-hygienic study of the factors of external exposure of personnel at the workplaces of experimental facility carried out in 2018-2019. During the research, dosimetric and spectrometric measurements of operational quantities were carried out at personnel workplaces, as well as the measurement of individual dose equivalents for personnel, including doses to the skin, using individual dosimeters. It has been shown that the neutron component forms about 20 percent of the external dose. The annual equivalent dose of irradiation to the skin of the hands was 190 ± 50 mSv, and with an increase in production volumes, this value will be a factor limiting the operating time. It was also determined that due to the configuration of workplaces, the equivalent dose of radiation to the lower abdomen is about 60% higher than the equivalent dose at the chest level, which is important for individual monitoring of female personnel. The radiation doses to the skin of the face and the lens of the eye turned out to be insignificant due to the sufficient protection provided by the protective glass of the sealed boxes.
The results obtained give an idea of the current levels of personnel exposure during the development of new technologies and make it possible to identify the main factors of the harmful effects of external exposure to which the personnel will be exposed during the industrial implementation of the mastered technology.

Speaker: Mr Pavel Gantsovsky (Burnasyan Federal Medical Biophysical Center of Federal Medical Biological Agency)
• 4.2 Structural, Novel, and Large Components Materials

Chairs: Massimo Angiolini

• 140
Creep and Creep-Fatigue Behavior of an Advanced Stainless Steel (Alloy 709) - Application to Sodium-Cooled Fast Reactors

Sodium-cooled Fast Reactor (SFR) possesses highest technology readiness level for deployment among six Gen-IV nuclear reactor designs intended to provide a low-carbon energy option and endure higher operating temperatures for longer service life (60-80 years). Thus, advanced materials developed for Gen-IV reactors should be able to withstand the harsh operating conditions allowing for safety improvement, efficiency enhancement and cost reduction. The advanced austenitic stainless-steel Alloy 709 (Fe-25wt.%Ni-20%Cr) is of current interest for structural applications in the SFRs owing to its desired set of properties including mechanical properties relative to conventional austenitic stainless steels. SFRs are subjected to on-load periods at elevated temperatures and thermal transients during startups and shutdowns, resulting in creep, fatigue, and creep-fatigue interaction as major considerations in the design of such high-temperature systems. In this work, high-temperature creep, fatigue and creep-fatigue behavior of the Alloy 709 are characterized along with microstructural examinations before and after mechanical testing. Creep tests were carried out at temperatures and stresses from 700 – 800 °C and 40 – 275 MPa, respectively, and the creep data were found to follow creep-power law with true stress exponent and activation energy of 4.9 ± 0.2 and 299 ± 15 kJ/mole, respectively. The microstructural observations of the crept specimens revealed different types of precipitates including Z-phases and the evidence of dislocation-precipitate interactions together with subgrain boundary formation. This suggests that high-temperature dislocation climb deformation is the rate-controlling creep mechanism in the alloy. Additionally, Larson-Miller Parameter (LMP) and Monkman–Grant relationships were developed using the creep rupture data.
Furthermore, strain-controlled cyclic loading tests were performed under constant strain rate of 2 × 10^-3 s^-1 at strain ranges varying from 0.3% to 1.2% at temperatures of 650 °C and 750 °C with tensile hold times of 0, 60, 600, 1,800, and 3,600 seconds. Introducing hold times at the maximum tensile strain resulted in creep regimes characterized by stress relaxation. The creep-fatigue life is found to decrease with increasing hold time, strain range, and temperature. According to ASME code, Linear Damage Summation rule was employed to predict the creep-fatigue life where creep damage was calculated using LMP and integration of the stress relaxation curve. Microstructural characterization of the deformed samples with no hold time indicated that fatigue is the dominant mode of deformation whereas with the hold times, both creep and fatigue contributed to the deformation of the alloy.
Financial support through DOE/NEUP program and KACST are gratefully acknowledged.

Speaker: Dr Abdullah Alomari (King Abdulaziz City for Science and Technology)
• 141
Development of Plasma Nitriding as alternate hardfacing technique for Large components of FBR and Assessment of static In-Sodium Stability of Plasma Nitrided Layer

Plasma nitriding is considered as a plausible alternate hard-facing technology for enhancing the wear and fatigue resistance of large size and intricate fast reactor components. Plasma nitriding is an environmentally clean process that can be adopted in principle, to produce high surface hardness in a controlled manner, with little distortion of finished components. In this regard, an attempt has been made to adopt plasma nitriding for the casing ring of the secondary sodium pump of PFBR.
A model with reduced thickness and another scaled-down version of the casing ring of the pump made of SS316L were fabricated. These rings were stress relieved, followed by chrome plating to a thickness of ~100 μm. Subsequently, the rings were plasma nitrided at 500°C for about 48 h under standardised process conditions (Working gas: H2:N2 ratio in 1:3 at 4mbar, 600V pulsed DC, 18A). No gross dimensional distortion of the component was found as a result of plasma nitriding. Since the rings have a step like discontinuity in geometry, the possible variation in microstructure and hardness profiles across the thin and thick sections of the scaled down version of the ring were characterized to assess the effectiveness of the plasma nitriding in providing a nitrided layer of reasonable case depth. It was found that a hardness of ~800VHN with a case depth of ~100 μm and 55 μm for thicker and thinner sections respectively could be achieved. X-ray diffraction analysis showed the formation of CrN and Cr2N phases in both the sections after plasma nitriding.
In order to assess the stability of plasma nitrided stainless steel component in liquid sodium, long term exposure to static sodium at 550°C up to 5000h was performed. Only a minor reduction in hardness, in the range of 560-700 VHN was observed due to thermal ageing induced from prolonged high temperature exposure in sodium. However, no significant change in microstructure or nitride layer case depth was observed. These results suggest the retention of long term integrity of the plasma nitrided stainless steel in static sodium.

Speaker: Dr R. Mythili (IGCAR)
• 142
A novel method of manufacturing a heavy integrated support ring in fast reactor

Support ring is a key heavy structural component of the sodium-cooled Fast Reactor (SFR), which supports the whole reactor vessel and the reactor internals, are subjected to high pressure and high temperature and other wind loads, seismic loads, dead loads etc. Therefore, the security and stability of the support ring is essential to nuclear reactor. As the support ring has a super large diameter up to 15.6 meters, it will theoretically consume more than 200 tons of stainless steel ingot. Unfortunately, it is basically impossible to make such heavy stainless steel ingot without segregation and shrinkage defects by using traditional manufacturing method. As an alternative, although the welding type of support ring could be manufactured successfully, the security and stability of the support ring will be significantly impaired due to several longitudinal welds. To tackle this grand challenge, we put forward a novel strategy to manufacture an integral weldless support ring (φ=15.6 m) by metal Additive Forging. In sharp contrast to the conventional wisdom that a heavy forging must be created by a larger roughcast, the core idea of this technique is to manufacture large high-quality component by building much smaller and cheaper metal slabs. The specific technological process is as follow. Before forging, the 316 stainless steel casting billets with cleaned surfaces are stacked in order and then vacuum-packaged by electron beam welder. Subsequently, the whole package is hot-compression bonded until the bonding interfaces were healed perfectly to form the initial billet. In order to obtain much larger billet, the initial billets are molded into the required billet shape and vacuum-packaged again by electron beam welder for secondary hot-compression bonding. After that, two initial billets are hot-compression bonded into a large billet required for ring rolling. Next, punching and broaching are performed on the cylindrical forging. Then, rolling process is carried out to acquire the required shape, size and microstructure. Finally, an integral support ring (φ=15.6 m) was manufactured successfully.

Speakers: Prof. Sun Mingyue (Institute of Metal Research, Chinese Academy of Sciences) , Dr Xu Bin (Institute of Metal Research, Chinese Academy of Sciences) , Dr Xie Bijun (Institute of Metal Research, Chinese Academy of Sciences) , Prof. Li Dianzhong (Institute of Metal Research, Chinese Academy of Sciences)
• 143
Material Data Acquisition Activities to Develop the Material Strength Standard for Sodium-cooled Fast Reactors

Adopting the 60-year design is regarded as one of the most effective means for the practical realization of Sodium-cooled Fast Reactor (SFR), which improves the economic efficiency and reduces the radioactive waste of SFR. In addition, since the happening of the severe accident (SA) at the Fukushima Daiichi Nuclear Power Station, the structural integrity evaluation of SA has been emphasized on SFR as well. As for the practical realization of SFR, it is indispensable to improve materials strength standards such as the extremely high temperature material properties which is required for the application of structural integrity evaluation during the happening of SA.
The material strength standards of SFR are stipulated by the Japan Society of Mechanical Engineers (JSME). However, in order to support the 60-year design, it is necessary to extend the time-dependent allowable value from 300,000 hours to 500,000 hours, as well as to increase the number of the fatigue evaluation cycles to up to 1×10^9 cycles. The Japan Atomic Energy Agency (JAEA) has developed a creep characteristic formula that is excellent for long-term creep evaluation based on the acquired long-term creep data, based on which a revised proposal to extend the evaluation time of material strength standard to 500,000 hours has been proposed. Furthermore, based on the high cycle fatigue data acquired for a maximum of 1×10^9 cycles, a high cycle fatigue evaluation method is under development. In order to evaluate the structural integrity during SA, the extremely high temperature material test data of SUS304 were acquired by JAEA and material strength standards were extended to 1000℃. In the future, to extend the material strength standards to high temperatures, it is scheduled to acquire extremely high temperature data for 316FR steels and Mod.9Cr-1Mo steels which are the main structural materials of SFR.
As described above, in order to make it possible to evaluate the structural integrity during 60-year design and SA, JAEA is working on the sophistication of the material strength standards. Moreover, material strength tests such as high temperature tensile tests, creep tests and fatigue tests are conducted systematically. In this presentation, the overall picture of material testing that we have acquired or plan to acquire in order to establish the JSME standard will be reported.

Speaker: Kodai Toyota (Japan Atomic Energy Agency)
• 144
The δ-ferrite transformation behavior and mechanical properties of 316H weld metal during high temperature service

316H austenitic stainless steel is widely used in the manufacture of nuclear reactor components owing to its excellent comprehensive properties, such as main vessel, support assembly, etc. During the fusion welding of austenitic stainless steel, the tendency of hot cracking tends to occur when the structural restraint is too large. For preventing the cracking, it is usually desirable to form a certain amount of δ-ferrite in the weld. However, the δ-ferrite is harmful to the mechanical and corrosion properties of the weld during high temperature service process, so the δ-ferrite content in weld metal must be strictly controlled. The microstructure evolution and mechanical properties of the 316H stainless steel weld metals with different C contents were studied at the aging temperature of 600 ℃ for different times. The results indicated that for the as-welded weld metal, with the increasing C content, the yield and tensile strengths increased, while the elongation decreased owing to the increasing C solid solution strengthening effect. During the aging process, the rapid precipitation of M23C6 carbide occurred in δ-ferrite firstly owing to the high diffusion rate of C. Once the carbon is depleted by precipitation of M23C6, the slow formation of σ phase occurred through eutectoid transformation (δ → σ + γ) depending on the diffusion of Cr and Mo. Furthermore, after a long enough aging time, a transformation from M23C6 to σ occurred. The C content has a significant influence on the δ-ferrite transformation behavior, δ-ferrite in the low and medium C weld metals transforms into M23C6 and σ phase successively, while δ-ferrite in high C weld metal only transforms into M23C6 carbides. The variations of mechanical properties with aging conditions depended mainly on the microstructures at different aging conditions. For the low C weld metal, as the aging time increased, fine M23C6 first precipitated, after that σ phase formed, the increasing σ phase content improved the strength obviously. For the medium and high C weld metals, as the aging time increased, first the depletion of the solid solution C as a result the M23C6 precipitation deteriorated the strength, then the formation of σ phase improved the strength. Furthermore, with the increasing aging time, the precipitation and coarsening of M23C6 and σ phase deteriorated the elongation and impact energy. This research provides theoretical and practical guidance for the control of the chemical composition and δ-ferrite content in the 316H weld metal for high temperature service.

Speaker: Shitong Wei
• 145
STATE OF DEVELOPMENT OF LEAD COOLANT TECHNOLOGY COMPONENTS FOR BREST-OD-300 REACTOR

The future of nuclear power lies in inherently safe fast reactors operating in a closed nuclear fuel cycle. The concept of inherent safety, which is based on deterministic exclusion of the most serious accidents due to the internal properties of the reactor, and not by creation of engineering barriers, is the foundation for ensuring the safety and economic efficiency of future nuclear power. “Proryv” is the leading Russian nuclear power project, which develops technologies of a fast lead-cooled reactor BREST-OD-300 and a closed nuclear fuel cycle based on the principle of inherent safety. The key element of ensuring the safety of the BREST-OD-300 fast reactor is the lead coolant.
Liquid lead was chosen as the main coolant for the nuclear power plant with inherent safety because of its physicochemical properties, which provide certain advantages over other types of coolants considered for inherently safe reactors, but its use also carries some technological challenges in the process of reactor operation.
The system for monitoring and maintaining the quality of the heavy coolant of BREST-OD-300 reactor consists of a set of high-tech devices designed to operate and maintain the key properties of the coolant over decades of operation. The development of these systems required an innovative approach, as well as many years of work to justify them.
The report presents the advantages of the lead coolant, its technology, the impact on safety and the current state of development for the following components of the system for monitoring and maintaining the quality of the lead coolant for the BREST-OD-300 reactor: oxygen activity sensor, mass exchange devices, hydrogen purification system, coolant filter, hydrogen sensors.

Speaker: Alexander Orlov
• 146
Tensile testing of sub-sized T91 and 316L steel specimens in liquid lead

The influence of liquid lead on mechanical properties of ferritic-martensitic steel T91 and austenitic stainless steel 316L have been studied in the JRC’s LIquid Lead LAboratory (LILLA). LILLA allows testing of mechanical and corrosion properties of materials in liquid lead with controlled dissolved oxygen concentrations and for temperatures up to 650°C. The load is generated by pneumatic bellow-based devices. For the present study, prior to tensile testing, polished sub-size flat and round tensile specimens were exposed in liquid lead at 450°C for at least 100h and at 500°C for at least 1000h. During this phase, the oxygen content dissolved in liquid lead was maintained below 10^-9 wt.%. The tensile tests on T91 were performed at 400°C in both argon and lead at an initial strain rate of 1x10^-4 s^-1. For 316L, specimens from both the base metal and the 75mm thick submerged arc welded (SAW) joint were cut out and tested. Tensile tests on 316L were conducted both at 400°C and 550°C at initial strain rates of 5x10^-5 s^-1 and 5x10^-6 s^-1, respectively. During all tensile tests, the content of oxygen dissolved in lead was monitored continuously and kept below 10^-8 wt.% with the cover gas control system implemented in LILLA (through flushing of Ar-H2 mixture). The test parameters were chosen based on outcomes of tests in lead-bismuth eutectic (LBE), where it had been demonstrated that mechanical properties of T91 are strongly influenced by strain rate, temperature, and oxygen content in LBE. The first LILLA post-test analysis have shown no evident impact of liquid Pb on the tensile properties of the investigated steels and welds exposed to the above-described conditions. The paper will address the observed results with respect to the literature data. This research supports resolution of key safety and licensing aspects related to structural materials and components for heavy liquid metal cooled fast reactor technology demonstrators considered in Europe, MYRRHA (with LBE coolant) and ALFRED (with lead coolant), and is embedded in the GEMMA Euratom Horizon 2020 collaborative project.

Speaker: Kamil Tucek (European Commission, Joint Research Centre)
• 147
New ASTRID SFR - Intermediate Heat Exchanger (IHX) and internal vessel interface system: qualification tests onto a scale 1 representative mock-up

Within the framework of the new evolutive French 400MWth Advanced Sodium Technological Reactor for Industrial Demonstration project (ASTRID), qualification tests have been performed in order to demonstrate and confirm the feasibility and the performance of major components.
One of them is related to the Intermediate Heat Exchanger (IHX) and specifically about the interface between the inner vessel (redan) and the IHX. This specific device aims to seal the coolant bypass between the IHX and the stand-pipe chimney installed onto the redan.
For that, piston seals principle device is organized and supported onto a floating sleeve hung on the IHX. The sealing device aims to maintain the primary sodium coolant to pass through the IHX tube bundle. It also enables vertical and radial sliding relative displacements between IHX component and internal vessel, especially during the different reactor thermal states.
Since this innovative option has never been implemented in any French Sodium Fast Reactor, a qualification tests program is launched using a metallic 304L full scale mock-up with representative mechanical and hydraulics loadings.
The paper presents the qualification principles, the qualification process on the full scale mock-up implemented. The different experimental tests performed and the results obtained are sum-up. At last, the next steps envisioned to finalize the demonstration are mentioned following the available promising behavior of the large piston seal device to be installed on a SFR IHX.

Key Words: ASTRID, Intermediate Heat Exchanger, Sodium, Redan, stand pipe, sealing

Speakers: Mr Antony WOAYE HUNE (Framatome) , Mr Alejandro MOURGUES (Framatome) , Mr Remi MORITZ (Framatome) , Mr Alexandre Villedieu (Framatome) , Mrs Marie-Sophie Cheneau (French Commission for Atomic Energy and Alternative Energy)
• 6.1 Neutronics

Chairs: Emil Fridman and Xingkai Huo

• 148
Comparison of calculation methods for lead cooled fast reactor reactivity effects

The application of Generation IV reactors offers improved sustainability of nuclear energy production and extends the current reserves while it helps to reduce the amount of nuclear waste. Therefore multiple demonstrator reactor designs are now under development. In fast reactors, due to its characteristic spectra and the reactor design, the effects of the leakage is much higher, which suggest the need of transport approximations with higher angular expansion for various analysis, such as sensitivity and uncertainty analysis.
In this study three different reactivity feedback coefficient of the Advanced Lead-cooled Fast Reactor Demonstrator (ALFRED) reactor is analyzed: the coolant temperature coefficient, the cladding expansion coefficient, and the fuel temperature coefficient. During the evaluations, the central difference direct perturbation method (DPM) and the linear perturbation theory (LPT) was applied to determine the feedback coefficients and their uncertainties. Multiple codes with fundamentally different calculation methods were employed during our investigations. The continuous energy SERPENT Monte Carlo code, the multigroup TSUNAMI-3D sequence of the SCALE program package, and PARTISN discrete ordinate neutron transport code coupled with SEnTRi developed at BME was applied and the properties of these fundamentally different calculations are presented. Furthermore, uncertainty calculations were performed in order to estimate the uncertainty of the reactivity feedback coefficients due to cross-section data. In order to confirm the validity of the applied tools and methodologies, we have performed lead void coefficient calculations for the Comet critical assembly and compared the results with the published experimental data.
Our results confirm that good agreement can be reached between the different methods and with the measurement results with properly chosen calculation parameters. Besides the high level transport approximation the appropriate spatial resolution has outstanding importance in oreder to describe the high flux gradient. While Monte Carlo solutions may use more accurate geometry modeling deterministic solutions may be superior in respect to the uncertainty analysis of reactivity coefficients where small differences needs to be determined.

Speaker: Máté Szieberth (Budapest University of Technolgoy and Economics)
• 149
Neutronics analysis of CEFR Start-up tests at IGCAR using FARCOB and ERANOS 2.1 Code Systems

The results of Phase-1 neutronics benchmark analysis of CEFR start-up tests was presented in the 2nd Research Coordination Meeting (RCM-2), held at Beijing, China during 28th October to 1st November 2019. Participation of this benchmark exercise has provided us a wide international forum to inter-compare the method, computer codes and cross section data employed at IGCAR for the physics design calculations of sodium cooled fast reactors by keeping the experimental data as the reference. The neutronics code systems used at IGCAR are FARCOB (deterministic) and OpenMC (Monte Carlo). In general, our simulation results are agreeing well for most of the work packages. However, few discrepancies observed for some simulations which are being resolved by detailed analysis. In this paper, the methods of neutronics simulation for 4 work packages (WP-1, 2, 3 and 4) and inter-comparison of simulated results (Phase-2) with the experimental value are briefly discussed.
The FARCOB code solves finite difference neutron diffusion theory in HEX-3D geometry using 26-group self-shielded cross sections from ABBN-93 cross section library. The neutron transport equation is solved in 2-D cylindrical model by first flight collision probability method by using COHINT code for homogenizing absorber rod cross sections. Monte Carlo calculations using OpenMC code uses realistic whole core pin-wise model of all sub-assemblies. Multiple nuclear data libraries, viz. ENDF/B VIII.0, JEFF 3.3, JENDL-4, ROSFOND-2010, CENDL-3 and TENDL-2017, have been processed using NJOY-21 and used for this analysis.
For the critical core, FARCOB under predicts the core reactivity by ~375 pcm. The criticality is predicted within 100 pcm with JEFF-3.3 by OpenMC. The spread in criticality prediction by various ENDF-6 libraries is studied using OpenMC. Absorber rod worth is predicted within 1-2 % of the measured value by these codes. The maximum deviation of temperature coefficient of reactivity with measured value is 10% and 5 % by FARCOB and OpenMC codes respectively. It is to be noted that statistical error of 0.04 pcm/K is achieved in OpenMC estimation by performing simulations for 1010 particles. Sodium void measurements using a special assembly was also simulated as exactly as possible and the estimated results are found to be within 10 % of measured values. Results of optional calculations for various reactivity coefficients are also discussed.

Speaker: Mr Abhitab Bachchan (Indira Gandhi Centre for Atomic Research, Kalpakkam, India)
• 150
Verification of the SPL module of the neutron diffusion code AZNHEX through Neutronics Benchmark of CEFR Start-Up Tests

AZNHEX is a deterministic code that solves the neutron diffusion equation with Hexagonal-Z geometry. It is part of the AZTLAN Platform, an initiative gathering the leading Mexican universities and nuclear energy research centers aimed to position Mexico in the midterm as a nuclear code developer for reactor core analysis and design. To tackle the challenges of a small core, like CEFR, with high leakage and significant anisotropic scattering effects, a new module for the AZNHEX code was recently developed based on the Simplified Spherical Harmonics (SPL) scheme. In this paper, to verify the SPL implementation and assess its usefulness,the results of the AZNHEX code with this new solver are shown and compared with the ones produced with the diffusion version; furthermore, comparisons against experimental data and the stochastic code SERPENTare presented. The cases simulated consisted ofa selection of the start-up tests performed in the Chinese Experimental Fast Reactor Benchmark proposed by the IAEA-CIAE. The nodal homogenized macroscopic cross-sections (XS) were generated through full core heterogeneous models in the stochastic code SERPENT using a novel methodology that is presented in another dedicated paper. The results obtained showed a great improvement compared with the diffusion solverreducing the deviations significantly from experimental data and stochastic results.

Speaker: Dr Roberto Lopez-Solis (National Institute for Nuclear Research)
• 151
Objectives and Status of Neutronics Sub-exercises of the OECD/NEA Benchmark for Uncertainty Analysis in Modelling for Design, Operation and Safety Analysis of SFRs

The OECD/NEA Benchmark for Uncertainty Analysis in Modeling (UAM) for Design, Operation and Safety Analysis of Sodium-cooled Fast Reactors (SFRs) was launched in 2015 to study reactivity feedback coefficients and their uncertainties for a medium-sized 1,000 MWth metallic core and a large 3,600 MWth oxide core. In addition to investigations of the full core level, stand-alone multi-scale neutronics analyses of several sub-exercises on the fuel pin and assembly level were identified to be relevant for a systematic assessment of the influence of nuclear data in fast reactor simulations. Especially when studying simple models, comparisons between computational results using different methods, models, and applied nuclear data libraries can unveil the major drivers of observed differences and obtained uncertainties.
The original specifications were therefore amended by neutronics sub-exercise specifications covering a fuel pin cell, a fuel assembly, and a super cell that includes a central absorber assembly surrounded by fuel assemblies for each of the two reactor designs. The nominal values and corresponding uncertainties of the eigenvalue, reactivity effects, and collapsed cross sections are requested. For the eigenvalue and reactivity effects, the top contributing nuclide-reaction pairs to the calculated output uncertainties are requested as well.
Up to now, 15 contributions to these sub-exercises from 9 international institutions were received. The preliminary results display good agreement in the calculated nominal results for calculations that were based on the same nuclear data library. A similar conclusion was drawn for the comparison of submitted uncertainties: differences in the applied covariance data dominate the cause of differences between results that were obtained using different methods for the neutron transport calculation and uncertainty quantification.
This paper provides an overview of the objective and status of the sub-exercises of the standalone neutronics phase of the UAM SFR benchmark. Preliminary comparisons between submitted results will be presented and preliminary conclusions of this part of the benchmark will be discussed.

Speaker: Friederike Bostelmann (Oak Ridge National Laboratory)
• 152
REALISATION OF AN ADJUSTED NUCLEAR DATA LIBRARY BASED ON ENDF/B-VIII.0 NUCLEAR DATA EVALUATIONS FOR THE ALFRED CORE

During the last decades, the development of new powerful computers and high performance analytical tools, along with the reduction of the approximations due to new methods implemented in the algorithms for the solution of the transport equation, pushed nuclear cross-sections data as the main source of uncertainty in neutronic calculations. This points out the importance in quantifying nuclear uncertainties and highlights the need for additional efforts to reduce them.

To accomplish this task, more advanced experimental facilities should be designed and built, more accurate measurement devices developed, and nuclear models necessarily refined, so requiring considerable time and investments both from an economic and human perspective.

However, the heritage supplied by tens of fast spectrum systems which were built and operated in the past, provides a considerable database which can be leveraged in order to avoid performing new experiments through the nuclear data adjustment technique.

ENEA, through the participation to the FALCON Consortium, is pursuing all activities required to support the construction of ALFRED – the European demonstrator of the LFR technology – in Romania, notably to what concerns the core design, of which ENEA is responsible.

In this context, in order to refine target accuracies for complying with the ambitious safety goals and/or setting new optimized safety margins for the ALFRED reactor, an extended version of the old, proprietary AMARA code, which was developed for adjustment purposes, has been developed.

This new version, named AMARA+, uses covariances opportunely generated from the most updated nuclear data evaluations, and extends the capabilities of the old code by flexibly allowing any user-defined selection of integral experiments conducted on systems/facilities representative of the system of interest. The approach chosen for the adjustment procedure in AMARA+ is still based on the maximization of the so-called likelihood function, which correlates the original nuclear data to the selected experimental integral parameters, and the optimization problem is solved via Lagrange multipliers technique. Furthermore, a chi squared-type test has been added for better substantiating the statistical soundness of the procedure.

In the present case, the reference nuclear data are taken from the ENDF/B-VIII.0 library, while a number of integral experiments, taken from the OECD/NEA’s IRPhE database, were used for performing a preliminary adjustment for the ALFRED reactor.

A comprehensive description of AMARA+ is presented in this work, stressing the aspects related to the algorithm chosen, together with the preliminary results obtained using the restricted set of integral experiments chosen.

Speaker: Mr donato maurizio castelluccio (ENEA, Bologna, Italy)
• 153
Spatial interdependence of safety related effects in ESFR-SMART core

ESFR-SMART belongs to the family of Gen-IV sodium cooled reactors. For its safety performance demonstration the detailed knowledge of selected safety related parameters is required. It is very important to know not only their spatial distribution and amplitude but also their mutual interdependence. In this paper the sodium void effect, the Doppler effect and fuel and cladding density effects were analyzed. The ERANOS 2 code was applied for the analysis of detail mapping of these effects and of the mutual influence between the effects. For each of the effect and for each of the interdependency many direct calculation were accomplished.
It was found that the void effect alone is spatially additive in the active zone, but not in the sodium plenum above the core. Especially the top fuel zone is strongly influenced by sodium plenum and vice versa. Furthermore, there is difference between radial and axial additivity of the void effect.
The Doppler effect alone is practically additive in the core. The interdependence of void and Doppler effect is limited, but not negligible. Other studied effects were related to fuel and cladding density change and they are spatially additive in the core. Finally, the results of this study can be used to recommend an approach of ESFR-SMART core modeling during transient analysis.

Speaker: Jiri Krepel (Paul Scherrer Institut)
• 154
The solution of nuclide kinetic equation for fast reactor in the OpenBPS code with options of choosing calculation method and uncertainties analysis.

The knowledge about fuel nuclear concentration behavior, fission product etc. both in fast reactor core and after withdrawing fuel from reactor is necessary from the point of view of nuclear safety as well as in future for fuel cycle closing up. There are a lot of methods and calculation codes in practice of burnup equation solution. Three different approaches can be highlighted, among which: a solution of nuclear kinetics equation with one of the matrix exponent method, iteration method with ability of uncertainties analysis, direct analytical solution method.
The software with open source code Open BurnuPSimulation (OpenBPS) provide for users a huge amount of means both for working with input data and choosing a solution method. Codes are being developed based on other open source instruments and approaches for solution burnup equation accessible in literature.
The Chebyshev Rational Approximation Method is implemented in the program for matrix exponent solution. An ability of using uncertainties analysis of the output nuclear concentration is provided by the iteration approach based on uncertainties of input, decay and nuclear cross-section data. The authors developed an analytical calculation method using modified Bateman functions (Analytical Solution of Burning Equations) for the accelerated solution of the burning equations. The calculation code ASBE was created on the basis of this method to calculate changes in the fuel concentrations during the operation of the BN-600 and BN-800 reactors. The speed of obtaining a solution in this code is more than an order of magnitude higher than the speed of obtaining a solution in programs of the iterative method, with high accuracy of the solution.
The program based on open source software provides a flexible and user-friendly interface with a choice of any of the listed solution methods and has unlimited potential for using at all stages of the fuel cycle from fabrication and placement into a reactor core to assessing of the activity and residual heat release during processing and preservation of used material.

Speaker: Ivan Bukhtiiarov
• 155
Study on actinide conversion capabilities of Molten Salt Reactors (MSR)

To date, French nuclear power plants mainly use uranium extracted from the mines and enriched. The used fuel is reprocessed to extract useful materials, such as plutonium, which is recycled once in dedicated PWRs of the park. Once used, the quality of the plutonium decreases and it becomes more difficult to recycle this plutonium in thermal spectrum reactors.

A molten salt reactor (MSR) is a liquid fuel reactor, which flexibility allows different fuel types to be used. This concept is inherently safe thanks to its negative feedback coefficients due to the liquid fuel travelling in the entire fuel circuit, from the core to the heat exchangers. Its versatility is such that this reactor can operate in burning or breeding modes. If operated with a fast neutron spectrum, the fission over capture ratio will be improved for all heavy nuclei, allowing to fission significantly most of the actinides. This reactor concept is therefore theoretically adapted to convert actinides such as plutonium.

New studies have started in a CNRS/Orano collaboration to evaluate the efficiency and options of MSR to convert actinides. Indeed, the concept of MSR, called MSFR, the most studied up to now in France and European projects is a fast thorium-cycle breeder reactor, hence a lack of knowledge on the converter concept.

In this article, we will show some preliminary results of these conducted studies. Neutronic studies use a Monte-Carlo approach to calculate the burn-up, feedback coefficients and irradiation damages, whereas reactor optimisations are carried out with a thermohydraulic and neutronic correlation-based algorithm. The molten solvent considered in the calculations -- NaCl-MgCl$_{2}$ -- to incorporate plutonium has shown a neutronic impact according to the proportions of the two constituents. A large amount of magnesium tends to epithermalise the spectrum, which is deleterious when considering non-fissile actinides. An observed increase of the irradiation damages on large core volumes will be explained and the impact of reflectors to mitigate this effect through the mean neutron energy will be presented. Finally the paper will show that the calculated feedback coefficients are very good ($\approx$-15 pcm/K), which assure a good inherent safety. The ongoing studies thus demonstrate the promising potential of fast neutron chloride MSR for such applications.

Speaker: Laura Mesthiviers (LPSC/IN2P3/CNRS)
• Lunch Break

Chairs: Paolo Ferroni and Andrei Moiseev

• 156
The Westinghouse Lead Fast Reactor: overview and progress in development

Westinghouse continues to develop its Next Generation high-capacity nuclear power plant based on Lead‐cooled Fast Reactor (LFR) technology. By leveraging its long experience in nuclear power plant commercialization as well as strategic domestic and international partnerships established to most effectively complement capabilities, Westinghouse is progressing the plant’s design and its business and delivery model. With competitive economics and market versatility as the prime program’s missions, the Westinghouse LFR is a competitive, medium-size (~450 MWe), simple, scalable and passively safe plant harnessing a liquid lead-cooled, fast neutron spectrum core operating at high temperatures in a pool configuration reactor. The plant adopts a reference power conversion system based on Supercritical Carbon Dioxide (sCO2) technology, which being air-cooled does not require proximity of large bodies of water thus resulting in enhanced siteability. The sCO2 power conversion system is coupled to an innovative, scalable, non-reactor-based thermal energy storage system developed by Westinghouse as part of a separate but synergistic effort, and aimed at providing the plant with load-following capabilities to complement non-dispatchable grid resources. This paper will provide an overview on the plant design, including progress in reactor core, reactor internals and passive decay heat removal system design, and will discuss some of the ongoing development activities which are detailed in companion papers presented at the FR21 conference.

Speaker: Dr Paolo Ferroni
• 157
Pilot Demonstrational Fast Reactor with Lead Coolant BREST-OD-300

Severe accidents in nuclear power industry (Three Mile Island, Chernobyl) brought NIKIET to the development of a fast lead-cooled reactor (BREST) concept at the end of the 1980s, which dramatically changed approaches to safety. To overcome contradictions between safety requirements and economical efficiency when designing a new-generation fast reactor, a new approach to selecting fundamental engineering solutions was proposed. It involves successive implementation of the inherent (or intrinsic) safety principle achieved not by building up expensive engineering barriers and complex safety systems, but mainly due to natural laws, physical and chemical fuel and coolant properties and the design solutions that contribute to realization of these properties to the fullest extent possible.
The BREST-OD-300 innovative fast lead-cooled reactor is developed as a pilot demonstration prototype for base-type commercial reactor facilities of the future nuclear power industry with the closed nuclear fuel cycle. To date, experimental justification of components, elements and equipment of reactor facilities has been carried out using small- and medium-scale mock-ups and pilot models. Validated and certified software tools were used for computational design justification. The safety analysis showed that the probability of severe accidents with conservative scenarios due to internal causes was no more than 6,48∙10-9 1/year. In such accidents, there is no loss of reactor core integrity with fuel and cladding meltdown, coolant boiling, and the radioactive release does not exceed the daily reference value. In severe accidents at the unit with a probability of 3.2∙10-8 1/year there is no need for evacuation and resettlement of the population.
The developed design documentation makes it possible to proceed to the construction of the BREST-OD-300 power unit as part of a pilot and demonstration energy complex (PDEC) with concurrent completion of activities aimed at bringing safety justification documentation to conformity with current common and developed specific standards for lead-cooled reactor facilities. At the PDEC site, pilot operation of the BREST-OD-300 reactor will be carried out in the closed nuclear fuel cycle conditions, endurance characteristics of the reactor facility and its equipment will be demonstrated, and an extensive R&D program, which is also required for commercial lead-cooled reactor facilities, will be carried out.

Speaker: Mr V.V. Lemekhov (JSC NIKIET)
• 158
CHOICE OF A COOLANT FOR A MODULAR SMALL POWER REACTOR SVBR-100

In recent years, small power modular reactors (SMRs) have begun to be developed in many countries around the world. At the last IAEA meeting of Technical Working Group on Fast SMRs (September, 2019), several concepts for SMRs with different coolants were presented. Interest in such reactors is due to a number of their inherent features: reducing the construction term, a higher level of safety, a longer campaign without refueling etc.).
The main difficulty to be overcome is the increase in specific capital costs while reducing the capacity of the reactor module. This limits the market niche of such reactors to regions with a high cost of fossil fuel. The noted difficulty increases with the use of SMRs and for heat supply, which requires the placement of such NPPs at a short distance from heat consumers and necessitates an increase in NPP safety. The safety level of a reactor is largely determined by the potential energy accumulated in the reactor coolant. The release of this energy in case of violation of normal operating conditions can lead to a large release of radioactivity, the localization of which requires an increase in the number of protective barriers and their effectiveness. This causes an increase in the cost of the NPP and a decrease of competitiveness.
The report analyzes the properties of various coolants and their effect on the internal self-protection of the reactor, its safety and economy, their advantages and disadvantages. Therefore, the choice of a coolant is always a compromise solution.
Taking into account the results of the analysis performed, the choice of chemically inert heavy liquid-metal coolants with a very high boiling point the lead-bismuth eutectic (LBE) and lead - was substantiated for fast SMRs, and among them, in favor of the LBE, taking into account its mastering in the operating conditions of ship reactors, a practical solution to the problem of ensuring radiation safety associated with the formation of polonium, a significantly lower melting point, which simplifies operation, and the possibility of "freezing-defreezing" of the coolant in the reactor. It permits to realize of the SVBR-100 reactor, which is being developed within the framework of a state-private partnership, with minimal investment risk.

Speakers: Prof. Georgii Toshinskii (JSC “AKME-engineering”) , Georgii Toshinskii
• 159
Core Design of 100MWe Advanced Nitride-fueled Simplified Liquid Metal Cooled Fast Reactor

A preliminary core design of 100MWe Advanced Nitride-fueled Simplified (ANTS) Lead-bismuth cooled Fast Reactor (LFR) for civilian multi-cycle nuclear power plant has been investigated. The prime design constraint is on the core size with the active core diameter and height equal to 2.4 m and 1.0 m, respectively. The core is composed of 144 hexagonal fuel assemblies enclosed with 15-15Ti steel duct. The core adopts the onion zoning with the two zones of low-enriched uranium for a better breeding capability to maximize the core lifetime for long cycle operation. Neutronics calculation is performed by the fast reactor analysis code system Argonne Reactor Computation (ARC) and the UNIST in-house Monte Carlo code MCS. It is confirmed that the cycle length of more than 12 years with a small burnup reactivity swing of less than 1,000 pcm is feasible. Power profiles generated by MCS are employed for thermal-hydraulic (TH) analysis with three different mass flow rates depending on the core power to flatten to the fuel temperature distribution. The research designs a reactor that is capable of natural circulation at an inlet temperature of 300°C. The total mass flow rate of 13,700 kg/s produces an outlet temperature of 450°C. TH analysis also demonstrates that the fuel and cladding temperatures are within normal operating range. Furthermore, several reactivity feedback coefficients have been computed by MCS to confirm the initial safety features of the suggested LFR.

Speaker: Mr Tung Dong Cao Nguyen (Ulsan National Institute of Science and Technology)
• 160
Novel neutronics design of the MYRRHA core

MYRRHA (Multi-purpose hYbrid Research Reactor for High-tech Applications) is internationally accepted as the key project to lead the industrial development of the accelerator-driven system (ADS) technology. As a flexible system, MYRRHA can also operate in critical mode when decoupled from the proton accelerator. By using lead-bismuth eutectic (LBE) both as a coolant and as a spallation target, MYRRHA acts as a technology demonstrator and a test platform for Heavy Liquid Metal (HLM)-cooled reactor technology for Gen IV systems. Thanks to its fast neutron spectra MYRRHA targets an efficient transmutation of minor actinides, hence playing a primary role in establishing the technological feasibility of burning high-level nuclear waste in a ADS systems. Irradiation rigs are also envisaged, to exploit the high constant neutron fluxes achieved in the reactor core representative of thermal, fast and fusion conditions.

Multiple neutronics designs of the MYRRHA critical and subcritical core configurations were proposed over the last two decades. Each release aimed at complying with requirements made available by the technical advancement of the project as a whole. The optimization process driving the latest core design (version 1.6) targeted an effective core fuel management and new layouts for material irradiation. A novel neutronics core design was recently finalized where additional constraints were introduced – e.g. a lower cladding temperature (400⁰C) and a reduction of the overall reactor size. This work describes the new layout of the MYRRHA core highlighting differences and similarities between the critical and subcritical configurations. The core neutronics performances are also investigated in terms of transmutation and material irradiation and they are compared to the catalogue requirements.

Speaker: Luca Fiorito
• 161
Project of a multipurpose lead reactor with a hard neutron spectrum

The concept of a low-power 25 MW thermal reactor with a hard neutron spectrum in the core is proposed. The reactor has the following feature: a sufficiently high neutron flux density 3,4•1015 n/(sм2•s) in the center of the core, high average neutron energy 0,86 МeV in the center of the core, as well as a high proportion of hard En>0,8 МeV, neutrons 35%. Extremely high design parameters of the reactor are achieved due to the small size of the core, DxH≈0,40х0,42 м2, innovative metal fuel made from an alloy of energetic plutonium with zirconium, Pu58%-Zr42%, and a heat carrier from a natural lead melt. The mass of power plutonium loaded into the reactor is 92 kg.
A line of 10-12 proposed reactors will allow more efficiently than in fast sodium reactors to burn off energy plutonium and minor actinides and at the same time generate about 100 MW of electricity. It is proposed to use high-background power plutonium as a fuel in the reactor, obtained as a result of the regeneration of fuel discharged from power thermal and fast reactors. In addition to plutonium and fission products, spent nuclear fuel, for example, from fast reactors, contains up to 1% minor actinides - neptunium, americium and curium, which in the hard neutron spectrum are transmuted into fission products with a higher fission probability than in fast sodium and molten salt reactors.
In addition to burning off transurans and generating electricity, a reactor with a hard neutron spectrum can be in demand as a research reactor, due to the content in its spectrum of a significant fraction, 35%, of hard neutrons with energies of 0.8 - 10.5 MeV, which have a higher capacity for radiation damage to materials. nuclear technology, rather than when they are irradiated in the spectrum of BN-type reactors.

Speaker: Samokhin Dmitrii (National Research Nuclear University МЕРhI)
• 162
Proposal of a compact core design for the 1000 MWe French commercial Sodium Fast Reactor by means of the SDDS multi-objective optimization tool

This paper present the core optimization of the 1000 MW (electric) French commercial Sodium Fast Reactor, aimed at selecting few optimal configurations with respect to both core safety and reactor cost criteria. The Generation-IV reactor core design process must conciliate multiple goals (e.g. highest reactor safety levels in any situation, easy exploitability, affordable cost), which are sometimes opposed. The optimization of the core design is usually realized on the basis of expert advices and local parametric studies, which require a strong knowledge of physical phenomena. However there is no guarantee to obtain the most reliable design since the parametric space is not fully explored.
To overcome this limitation EDF has developed for more than 10 years a method for global multi-objective optimization of SFR cores: the SHADOC-based Design Development System (SDDS). The SDDS methodology relies on a multi-physics tool, including neutronics (ERANOS code), thermal-mechanics (GERMINAL code) and thermal-hydraulics (in house code) modelling. It ensures an exhaustive scan of the available design options within a given range of variation of the main reactor parameters (e.g. pin/cladding size, height and volume of the core, fertile plate height and position, etc.) by exploiting surrogate models built on a reduced number of multi-physics evaluations.
With the objective of reducing the reactor cost of the future French commercial Sodium Fast Reactor (SFR), a CEA/FRAMATOME/EDF Working Group was commissioned to work on the first specifications to be retained for a 1000 MW (electrical) commercial SFR. Reduction of the vessel and the core diameters were identified among the possible options to improve the competitiveness of SFRs. Thus, in the last two years, several studies have been carried out by EDF, leading to the definition of optimized cores by means of the SDDS multi-objective optimization tool.
Amongst all the core designs exhibiting a favorable behavior in both CRW (Control Rod Withdrawal Accident) and ULOSSP (Unprotected Loss of Station Service Power) unprotected transients, two compact low-Sodium Void Reactivity Effect cores were selected for their reduced fissile core diameters :
- The first one is based on a 12-fuel subassembly rows design, which perfectly meets the core size criteria (arbitrarily set to a fissile core diameter lower than 4 m),
- The second one, based on a 13-fuel subassembly rows design, although slightly wider, exhibits a better behavior for the two unprotected accidents considered (ULOSSP and CRW).

Their characteristics and performances will be detailed in the paper.

Speaker: Dr Sandra POUMEROULY (EDF)
• 163
Conceptual design of ultra-long life hybrid micro modular reactor cooled by potassium heat pipe

As part of achieving sustainable development, a concept of the hybrid micro modular reactor (H-MMR) has been proposed by integrating the MMR design developed by KAIST with renewable energy and energy storage systems (ESS). The reactor power is designed to be 18MWth, and it is aimed for an ultra-long core lifetime for more than 20 years without refueling. The H-MMR core consists of 18 hexagonal fuel assemblies (FAs) with a potassium heat pipe cooling system. The hexa-annulus types of UN fuel, including potassium heat pipes, are assembled into the oxide dispersion-strengthened steel (ODS) hexagonal duct. These heat pipes are connected to a sodium pool that is set up above the reactor core as an intermediate heat exchanger. The PbO and ODS reflectors are designed with a B4C shielding layer. The primary reactivity control system is placed in the radial-reflector as a drum-type, and a conventional secondary reactivity control device is located in the center of the core. The neutronic analyses were performed by Monte Carlo code, Serpent 2, with ENDF/B-VII.1 data library. The results showed that the H-MMR achieves ultra-long life of 56 years without refueling and the discharge burnup is 37.12 MWD/kgHM, while the reactivity swing over the whole lifetime is less than 0.45 dollar. The results suggest that the ultra-long life micro module type fast reactor design that guarantees the improvement of the inherent safety via the use of potassium heat pipes can be realized.

Speaker: Mr Seongdong Jang (Korea Advanced Institute of Science and Technology (KAIST))
• 2.3 Accident Analysis

Chairs: Andrei Rineiski and Koji Morita

• 164
Coupled neutronic/thermal-hydraulic simulation of Unprotected Loss of Flow Test at Fast Flux Test Facility

Fast Flux Test Facility (FFTF) was a research Sodium-cooled Fast Reactor (SFR) operated in the 1980’s with a goal to demonstrate the inherent and passive safety characteristics of the SFR design. In the frame of this study, an attempt has been made to develop a model of the FFTF suitable to predict the system’s behavior during the Unprotected Loss of Flow (ULOF) accident. In particular, stochastic Monte Carlo code Serpent 2 is employed to build the static neutronics model of the system, results of which are further utilized to construct the neutron kinetics model of the core by employing the Purdue Advanced Reactor Core Simulator (PARCS). This model is further coupled to the plant system code TRAC/RELAP Advanced Computational Engine (TRACE) in the frame of the FAST code system in order to enable comprehensive analysis of transients within the fast reactors. Special attention is devoted to the Gas Expansion Modules (GEM) and their potential to drive the reactor core into the safe shutdown state upon occurrence of initiators of the ULOF accident. On top of that, assessment of representativeness of GEM devices in modeling the coolant boiling within the fast reactors is conducted. Furthermore, a number of sensitivity studies is performed in order to assess the performance of the various simulation tools and the modeling approaches in faithfully reproducing the experimental data. Obtained results are further compared to the measurements collected during the Loss of Flow WithOut SCRAM (LOFWOS) Test #13 performed at the FFTF. According to the outcome of the aforementioned comparison, LOFWOS Test #13 can be successfully reproduced by employing the FAST code system. Spatial reactor kinetics model is proven able to reproduce the evolution of the core parameters to a high degree of accuracy. However, by employing a fairly simple point reactor kinetics model, a conclusion is drawn that the aforementioned models perform comparatively similar in modeling the neutronics of the reactor cores of similar geometry, size and fuel composition as FFTF’s. Furthermore, it is proven that the axial power profile of the core does not suffer significant degradation in the process of activation of GEM devices. Moreover, on the basis of the comparison to the performance of the sodium plenum of the Advanced Sodium Technological Reactor for Industrial Demonstration (ASTRID), GEM devices are proven to be a good representative of the coolant boiling within the fast reactors.

Speaker: Dr Mikityuk Konstantin (Paul Scherrer Institute)
• 165
MECHANISTIC MODELLING OF AEROSOL EVOLUTION IN AN SFR CONTAINMENT FOLLOWING A HYPOTHETICAL SEVERE ACCIDENT

Severe accident source term problem is the calculation of transport and release of radionuclides (RNs) from the reactor core to the environment for a hypothetical accident sequence of low probability. Generally, the source term problem is decoupled into four source terms: 1) In-vessel source term, 2) Interface source term 3) In-containment source term and 4) Environmental source term.

In present work, a medium-size oxide fuelled sodium-cooled fast reactor (1250 MWt) is chosen as the reference design for the calculations. For the analysis, a whole core melt accident initiated by the loss of flow to coolant pumps with the failure of both the shutdown systems (ULOFA) is considered. The temperature and pressure transients due to the burning of ejected sodium are studied. Additionally, the evolution of the RN and sodium aerosols generated from the sodium fire is analysed. For realistic estimates of the in-containment source term, in-vessel release fractions calculated from MINICHEM are used as an input.

The present calculations are performed for the two sets of in-vessel release fractions viz., 1) No mixture 2) Real mixture to be used as input. We found that following an accident, Cs, Rb, I, Xe, and Kr are the dominant suspended species in the containment. Additionally, it is observed that after 5 hours, less than ~15% of the total aerosol mass is suspended. After 24 hours less than ~2% of the total aerosol mass are suspended. The present calculations indicate that an effective containment, holding RN for the first 24 hours would be helpful to reduce the radioactivity available for release. The calculations performed here is valid for the short-term/instantaneous source term. Work is in progress to include aerosol generation models and multi-component aerosol modelling.

Speaker: Mr PARTHKUMAR RAJENDRABHAI PATEL (Indira Gandhi Centre for Atmoc Research, India)
• 166
Modelling and Simulation of Source Term for Sodium-Cooled Fast Reactor under Hypothetical Severe Accident: Primary System/Containment System Interface Source Term Estimation

Three Work Packages (WPs) were defined in this Coordinated Research Project (CRP) whose objective was to estimate fission-product-transportation behaviour inside the reference pool-type sodium-cooled fast reactor (SFR) volumes (i.e., in-primary vessel, cover gas system and in-containment building) at different time scales under severe accident conditions. This WP, WP-2, is defined to estimate the primary system/containment system interface source term using improved models and tools for the cover gas, sodium ejection, radionuclide chemical composition and distribution in the containment. After the discussion between the participants of this WP, it was decided to evaluate mass of primary sodium instantaneously ejected into the Reactor Containment Building (RCB) as a common benchmark problem.
The exercises were carried out for a reference pool type SFR of 1250 MWth capacity fuelled with mixed oxide fuel. The accident sequence to be considered is Unprotected Loss of Flow Accident (ULOFA) which is assumed to result in a core damage event with release of radionuclides into the primary coolant and cover gas.
Four organizations, NCEPU (China), IBRAE/RAN (Russia), IGCAR (India) and JAEA (Japan) were finally participated in this WP. NCEPU calculate the leakage of liquid sodium using CFD code, FLUENT. IBRAE/RAN did the simulations with their code EUCLID/V2. IGCAR used FUSTIN/BLVDYN and NETFLOW codes for simulations. JAEA used PLUG code for the calculations of sodium ejection.
Basic and parametric case of the calculation were carried out. The total amount of the ejected sodium onto the roof slab for basic case was in a good agreement between the participants. The results of the parametric analysis revealed that the pressure difference above and below the vessel head, which is the driving force of the sodium ejection, and the resistance coefficient of the bent portion of the plug gap make a relatively large contribution to the amount of ejected sodium.
In WP-2, the amount of sodium ejected onto the roof slab was evaluated as a common benchmark problem. The long-term release of radio nuclide would be due to the release of suspended/dissolved activity from sodium through the leak paths made available in the top shield. Quantitative evaluation of such long-term release is addressed as the future work.

Speaker: Yuichi ONODA (Japan Atomic Energy Agency)
• 167
Modeling and Simulation of Source Term for Sodium-Cooled Fast Reactor under Hypothetical Severe Accident: Sodium Fire and Radionuclide Transport in Containment

The main objective of the coordinated research project (CRP) on is to simulate the fission product transportation behavior of the reference pool-type sodium-cooled fast reactor (SFR) of 1250 MWth capacity with mixed oxide fuel under severe accident conditions. The accident considered is an unprotected loss of flow accident resulting in a core damage event with release of radionuclides. The work package 3 (WP-3) essentially models and simulates the ‘in-containment phenomena’ after the postulated severe accident, which includes sodium chemical reactions and aerosol mass evolution in the containment.

Seven organizations from six countries participated in WP-3. CIAE, CEA, and TerraPower used CONTAIN-LMR code and its derivatives. XJTU used REBAC-SFR code. IGCAR used SOFIRE based fire simulation code, PFIRE, and PANDICA for in-containment aerosol evolution. IBRAE RAN used multiphysics EUCLID/V2 code and CIEMAT used ASTEC-Na code.

To decouple this part of analysis from previous stages of calculation, the stand-alone calculation was defined for WP-3, which uses a set of pre-defined release fractions. The stand-alone case is appropriate for inter-comparison with respect to assessing the tools of WP-3 calculations. In the coupled case, the release fractions of radionuclides computed at the previous work packages were used as initial conditions. Both IBRAE RAN and IGCAR participated in this coupled simulation. The purpose of this effort is to demonstrate the conservatism built in the WP-3 inputs such as release fractions, chemical forms, and sodium ejection amounts. The obtained discrepancies in activities of airborne fission products (FP) in containment take place mainly due to differences in modeling of FP release from molten fuel and of aerosol growth and deposition.

In both integral and stand-alone calculations, the sodium fire and subsequent radionuclide release are modeled in two separate cases: (1) sodium spray fire with instantaneous reaction that results in the highest containment pressure while the containment does not leak, and (2) sodium pool fire that results in a prolonged burning of sodium in a compartment while the containment is leaking at the design leak rate.

The sodium spray fire exercise sets the baseline for the participants to compare the results based on the relatively straightforward boundary conditions. The pool fire case shows observable differences among the organizations due to the complexity of the sodium reaction phenomena, which also drive the aerosol release into the containment via the small compartment. In conclusion, there is broad consensus among the predicted results in WP-3 with respect to the stand-alone case.

Speaker: Dr Jong E. Chang (TerraPower, LLC)
• 168
The Severe Accident Management of the high-power SFR with loss of the heat removal from the core

The accident at the Fukushima NPP (Japan) showed that the design of the power unit should consider the unforeseen excess of the external influence intensity. The accidents for internal reasons can be predicted on the basis of knowledge and it depends, first of all, on the designers approach, but it is impossible to foresee catastrophic external influences.
This method is proposed the heat removing for from the vessel by air for a high-power SFR (RVAC). This report proposes to limit heat dissipation - not to exceed the shell temperature of 800 °C in a severe beyond design basis accident such as the accident at the Fukushima NPP.
The air duct composition of the Severe Accident Management (SAMG) proposed method consist of: cold air supply pipes, a hot air heat-insulated exhaust pipe and an air duct around the safety vessel, created by shell with a gap.
Heat transfer to air is carried out due to the transfer of heat by radiation to the safety vessel and shell and due to heat removal from the safety vessel and shell surface to the air in the gap between them, during air circulation through the air path. The thermal balance of the sodium and the vessel are established by intense circulation flows of sodium, which are forced by the core decay heat and the heat outflow to the air. Many-hour processes are considered, so there is no delay in the temperature of structural elements from the sodium.
As a SAMG including the gap between the main and safety vessel to the air cooling system and filling this gap with sodium is considered like option.
It is shown that taking into account heat exchange by radiation provides a safety SAMG regime for the fast reactors with a sodium coolant for removing heat from the reactor vessel through the channels of natural air circulation around the reactor. Even it’s possible to keep the reactor vessel temperature below the limit of 800 °C with heat removal only through the safety vessel - shell channel.

Speaker: Ilia Pakhomov
• 169
Over three decades of radiological protection experience at Fast Breeder Test Reactor (FBTR)

The Fast Breeder Test Reactor (FBTR) at Kalpakkam is the flagship of the second stage of the three-stage nuclear program of Department of Atomic Energy in India. Health physics services commenced at FBTR in April 1985 itself and the reactor went critical in October 1985. FBTR is a unique reactor utilizing U-Pu C as the fuel. Presently, FBTR has been operated upto a maximum power of 32 MW(t) and the fuel has attained a maximum burnup of 165 GWd/t. Health Physics experience gained from the operation of reactor for 35 years is outlined. These include area monitoring, stack monitoring, annual discharge of activity released vis-à-vis technical specification limits, personnel monitoring that include man-rem expenditure, waste disposal etc. Adequate radiation protection measures coupled with effective surveillance by the health physicists have made it possible to have low personnel exposures. Installation, calibration and usefulness of special monitors, unique to LMFBRs, such as gas flow ion chambers in the Clad Rupture Detection (CRD) argon circuit for detection of gaseous fission products, fume activity monitors in the ventilation ducts to indicate sodium leak / fire, sodium aerosol detection monitors in the primary double envelop sampling line and gas activity monitors are highlighted. The experienced gained during fuel clad failure events at FBTR has been unique on the radiological protection and failed fuel failure identification point of view. Fuel clad failure incident of short time release with wet rupture and clad failure with gas leaker of slow release due to dry raptures are experienced. Reactor cover gas activity samplings and analysis has been helpful in the identification of failed fuel assembly.
Towards controlling external exposures to occupational workers during maintenance work, initial baseline studies were conducted to assess the deposition of radioactive corrosion and activation products and dose rates in the primary sodium pipelines and various components of FBTR, which are housed in B-cells. Environmental impact is an important global issue for any energy option. The release of radioactive materials from FBTR is negligible. The unique experiences gained would serve as a guide for a safe approach and in determining the criteria from the point of view of radiation protection for future LMFBRs being planned in India.

Speaker: Sarangapani Rajagopalan (Indira Gandhi Centre for Atomic Research, Kalpakkam)
• 170
Safety Analysis of Small Modular Sodium Fast Reactors in Anticipated Transients Without Scram Scenarios

The small modular sodium-cooled fast reactor (SFR) is an important component of Generation-IV reactors. For SFR, one type beyond design basis accidents (BDBA) that has received special attention is the anticipated transients without scram (ATWS) events including unprotected loss of coolant flow (ULOF) accident, unprotected loss of heat sink (ULOHS) accident and unprotected transient overpower (UTOP) accident. The modular design for multiple purposes and remote region operation requires usually very infrequent refueling strategy. During a long-lived operation, the neutronic characteristics of SFR core, for instance, the coolant void effect and the Doppler effect, vary and hence the reactor safety performance in ATWS events. This paper focuses on the analysis of safety performance of a 300 MWth small modular MOX SFR from its beginning of life (BOL, 0 GWd/tHM) to its end of life (EOL, 75 GWd/tHM). The burnup calculation is conducted by using Monte-Carlo code OpenMC with pin-by-pin depletion mesh. The elementary reactivity feedback coefficients and core power distribution are compared at different burnup depth. The transient behavior is simulated by using a newly developed mono-channel point kinetic system code dedicated to fast reactors. The inherent reactivity feedback mechanism in ATWS is classified. The influence of the fuel burnup, the power redistribution, and the control rod positions are investigated. Some solutions, for instance, the use of burnable poisons, the design of upper sodium plenum, the addition of moderators, are discussed in the view of mitigation of the ATWS impact and the inherent safety performance for a long operating lifetime.

Speaker: Mr Xin Jin
• 171
The Versatile Test Reactor (VTR) Approach to Sodium Fire Hazards Analysis and Protection System Methodology

The Versatile Test Reactor (VTR) is a fast spectrum test reactor currently being developed in the United States under the direction of the US Department of Energy (USDOE), Office of Nuclear Energy. The mission of the VTR is to enable accelerated testing of advanced reactor fuels and materials required for advanced reactor technologies. The conceptual design of the 300 MWth sodium-cooled metallic-fueled pool-type fast reactor has been led by US National Laboratories in collaboration with General Electric-Hitachi and Bechtel National Inc.

Current VTR efforts include the development, design, and analysis of a sodium fire protection system (SFPS) strategy for the reactor facility. This paper summarizes the evolution of SFPSs in the U.S. from previous test reactors such as the Experimental Breeder Reactor II (EBR-II) and the Fast Flux Test Facility (FFTF) to larger proposed commercial reactors designs such as the Clinch River Breeder Reactor (CRBR), PRISM, the Sodium Advanced Fast Reactor (SAFR), and the Large-Scale Prototype Breeder (LSPB). Additionally, the software tools (NACOM and SOFIRE-II) utilized to analyze sodium fire scenarios for the VTR are described along with a summary of current software verification and validation (V&V) efforts. Finally, the design philosophy for the VTR SFPS strategy is presented, which is based on SPFS design evolution and insights from preliminary sodium fire analyses. The work reported in this summary is the result of studies supporting a VTR conceptual design, cost, and schedule estimate for DOE-NE to make a decision on procurement. As such, it is preliminary.

Speaker: Matthew Bucknor (Argonne National Laboratory)
• 5.1 Experimental Reactors and Facilities

Chairs: Jordi Rolans-Ribas and Hideki Kamide

• 172
Complex of experimental facilities for design and safety justification of fast reactors with liquid metal coolants

For design and safety justification of fast reactors with liquid metal coolants, a complex of more than 20 experimental facilities of various functions and purposes, equipped with modern measuring instruments, including hydrodynamic, thermohydraulic and technological facilities, has been created at IPPE JSC.
In addition IPPE JSC has a complex of fast neutron facilities, including two critical facilities - BFS-1 and the neutron facility BFS-2 which is the largest in the world.
The report presents characteristics and capabilities of test facilities designed for research in the field of hydrodynamics, heat transfer and coolant technology in support of design solutions, safety improvement and testing of equipment and components of fast reactors with sodium, lead and lead-bismuth coolants, as well as for accelerator-controlled systems and thermonuclear fusion, low-power nuclear power plants for space:
– Thermal-hydraulic liquid metal test facilities - "6B" (Na, Na-K), "AR-1" (Na, Na-K), "Pluton" (Na), "SPRUT" (Na, Na-K, Pb, Pb- Bi, water).
– Technological liquid metal test facilities - "Protva-1" (Na), "Protva-2" (Na), "Armatura" (Na), "IK-MT" (Na), "SID" ( Na), "BTS" (Na), "TT-1M" (Pb), "TT-2M" (Pb-Bi), "LIS-M" (Li).
– Hydrodynamic test facilities - "SGDI" (air), "V-2" (air), "SGI" (water), "V-200" (water), "GDK" (air).
A large-scale sodium test facility "SAZ" is under construction. It will be used for testing full-scale equipment prototypes to substantiate actual and future fast sodium reactors design.
The complex of fast neutron facilities BFS is the world's only experimental facility for full-scale modeling of the cores of nuclear reactors of various types (of any composition, geometry and configuration). Components and design of the facilities can be used for modeling cores, blankets, core reflectors etc. as well as components of fuel cycles and storage facilities for radioactive waste. Fuel materials of the facilities (metallic plutonium, oxide and metallic highly enriched uranium with enrichment of 36% and 90% in uranium-235, hundreds of tons of fertile materials, construction materials, various coolants) make it possible to assemble both complex full-scale models of fast reactors, and benchmarks, experiments for which are carried out to correct neutron-physical constants and improve computational methods.
A great number of experimental researches have already been carried out at all test facilities and at present a lot of experiments are planned.

Speaker: Mrs Iuliia Kuzina (IPPE JSC)
• 173
Physical modeling of hydrodynamics and heat exchange in fast reactors with liquid metal coolants

In order to develop numerical modeling of hydrodynamics and heat exchange for substantiation of design and safety characteristics of fast reactors planning and carrying out experiments, generalizing the experimental results according to the physical laws and to their closing relations and verification of codes are required. To solve these problems, a complex of liquid metal facilities, equipment, modeling methods, measurement techniques, sensors, etc. were created. A great number of experimental results on heat exchange for Hg, Pb, Pb-Bi, Na, Na-K, Li, Cs were obtained.
When preparing and carrying out the experiments, the most important issue is the fulfillment of the conditions for mechanical, thermal and thermodynamic similarity, which determine the dependence of the physical properties of flowing medium on the state parameters and general dependences for describing heat transfer in various liquids under various conditions of hydrodynamics and heat exchange in the objects under study. Although a lot of monographs have been written, nevertheless, the simplicity (at first view) of the foundations of this analysis and the multi-parameter nature of the modeling tasks result in misunderstandings and gross blunder.
The goal of the study was to substantiate the possibility of applying the principles of modeling of thermophysical processes in liquid-metal coolants. They are specific coolants, and the theory of similarity is used for experiments on models with other coolants to substantiate hydrodynamics and heat exchange in fast reactors and to transfer the experimental data to reactor conditions.
The results of the application of the modeling theory to the hydrodynamics and heat exchange issues in liquid metals are presented: in complex shape channels, in rod (pin) systems (core and heat exchangers of fast reactors), in a reactor vessel in various regimes, for boiling of liquid metals in the core in accidental conditions. It is shown that direct modeling can be unlimitedly applied only for such processes as hydrodynamics, for which the determined similarity numbers are the functions of the only geometric simplexes of the system and of only one determining criterion (Re).
If there are two determining criteria, for example, the Re and Pr numbers during heat exchange, the modeling is significantly complicated. And under three defining criteria (mixing chambers, collector systems), direct modeling is usually not feasible. It is necessary to carry out systematic multivariate experiments. Their objectives are the real identification of effects within the framework of a general mathematical model that cannot be reproduced either analytically or numerically.

Speaker: Mrs Julia Kuzina (SSC RF - IPPE)
• 174
Overview of a Sodium Fast Reactor Thermal Hydraulic Test Facility

The Thermal Hydraulic Experimental Test Article (THETA) is a 150-gallon reactor grade sodium facility built to study the phenomena that occur in a fast reactor during normal operations as well as accident scenarios. THETA will provide critical data to sodium fast reactor designers with higher fidelity than ever before for a scaled liquid sodium facility; THETA is instrumented with advanced sensors such as a submersible permanent flowmeter that uses cutting edge SmCo magnets with an elevated curie point to acquire coolant flow rates as well as distributed sensors that can acquire temperature readings at 10,000+ locations with the use of optical frequency domain reflectometry. Key variables such as intermediate heat exchanger outlet elevation in the cold pool, pump speed/spindown rate, core power input, etc. can be varied in real time to study their effect on the thermohydraulic behavior of the reactor. This paper will discuss the design, commissioning and current test results from the THETA facility.

Speaker: Matthew Weathered
• 175
Overview of the R&D programs led by the past at IRSN on sodium fire

In the frame of the implementation of fast breeder reactors in France initiated in the 1960’s, a lot of studies have been conducted by CEA/IPSN (now IRSN) on sodium fires and risks potentially induced by sodium in general. The experimental studies started at a fundamental level, in order to understand the basic phenomena concerning sodium fire. From 1972, R&D programs dealt with small and mid-scale experiments aiming at gathering knowledge for the development of physical models and computer codes. Dedicated experimental facilities were built, in order to (1) explore pool and jet sodium fires in confined environment, specific of the nuclear industry, atmospheric dispersion of the aerosols when released and sodium-concrete reaction, and (2) develop technological equipment (water scrubber, instrumentation for detection, active and passive mitigation systems, etc.). Gradually, the benefit from this R&D led to improve and qualify computer codes, and to improve the safety of fast breeder reactors and other facilities involving sodium. With the construction of the fast breeder reactor SUPERPHENIX in France, real scale tests appeared necessary with the objective of exploring scale effects and qualifying codes used for safety assessment. This was the aim of the ESMERALDA program launched in 1982 that consisted of performing, until 1989, 26 experiments for the purpose of the SUPERPHENIX safety assessment, apart high flowrate jet fires that were not included in the program from its start. Therefore, the additional test series IGNA 3600 and 2000 was performed in the ESMERALDA facility, owing to extend the qualification field of the computer codes used in safety assessment and to bring modifications on the reactor.
The presentation aims at giving an overview of these programs, and at sharing knowledge learnt from these important tests, that can be profitable for improving risk analysis of generation IV SFRs including SMRs.
The presentation aims at giving an overview of these programs, and at sharing knowledge learnt of from these important tests, that can profitable for improving risk analysis of generation IV reactors and SMRs.

Speaker: Jean-Marc SUCH (IRSN)
• 176
Mechanisms Engineering Test Loop (METL) Facility

The purpose of this paper is to present the Mechanisms Engineering Test Loop (METL) facility installed at U.S. DOE’s Argonne National Laboratory located outside of Chicago, Illinois. The METL facility is a multipurpose intermediate-scale 750-gallon (2839l) sodium test facility used for the testing of liquid metal systems and components in prototypic reactor-grade sodium conditions. Some examples of technologies that can be tested in METL include: components for advanced fuel handling systems, advanced sensors and instrumentation, in-service inspection and repair technologies, thermal hydraulic testing in a prototypic environment, components for reactivity control, and health monitoring technologies for liquid metal reactor applications. METL also provides development opportunities for younger scientists, engineers, and designers who will ultimately lead the advancement of U.S. liquid metal technologies. The hands-on experience with METL, both successes and perceived failures; will ultimately lead to better liquid metal technology programs that can support the commercialization of advanced reactors.
The METL facility consists of two 18-inch (45.7-cm) test vessels, two 28-inch (71.1-cm) test vessels, an expansion tank, a dump tank, a purification loop, and a plugging loop. METL has a main piping loop that includes an electromagnetic pump and flow meter. This main loop provides flowing sodium to the four test vessels. The test vessels can be connected to the main loop in parallel operation or a test vessel can be isolated from the main loop with its own sodium environment. Each test vessel can be drained independently to the dump tank without affecting the other on-going testing. The facility has a maximum operating temperature of 1000°F (537.7°C) with the larger test vessels capable of obtaining 1200°F (648.8°C). 15 55-gallon (208l) drums of sodium was transferred to the dump tank in April 2018. Sodium piping, purification, plugging loop and expansion tank was filled in September 19, 2018. The facility has been in an operational state for over two years. This paper discusses the overall design philosophy of the facility, the testing that has been performed in the facility, and the approaches used for inserting test articles (experiments) into and removing test articles from the facility.

Speaker: Mr Christopher Grandy (Argonne National Laboratory)
• 177
EXPERIMENTAL CAPABILITIES OF THE RESEARCH REACTOR FACILITY MBIR. MAIN AREAS OF THE RESEARCH PROGRAMME IN THE INTERESTS OF THE GENERATION 4 REACTORS

Multifunctional Research Fast Reactor (MBIR) is being built in the Russian Federation for justification of the new projects of nuclear power based on fast reactors operated in the closed nuclear fuel cycle. MBIR will make it possible to carry out a wide scope of reactor research including studies into new types of fuel and structural materials in combination with different coolants. It can also provide a solution to the problems of safety, reliability and economic efficiency of designed NPPs with fast reactors.
The reactor main design parameters and experimental capabilities are described in the report.
Special attention is paid to the Advanced Scientific Research Program for the 2028-2040 period, which includes potential studies on independent loop facilities for the benefit of G4 reactors.

Speaker: Dmitrii Klinov
• 178
Versatile Test Reactor (VTR) Project Mission and Status

Advanced nuclear technologies will play a significant role in meeting the growing demand for clean energy. To support the deployment of these technologies and enable long term innovation, the U.S. DOE has initiated a project to build a Versatile Test Reactor (VTR) with a compelling and urgent mission: testing innovative fuels, materials, sensors and instrumentation for various advanced reactor types under development. VTR will enable testing with very high fast neutron fluxes (≥ 4 x 1015 neutrons/cm2-s fast flux) over large volumes (≥ 7 liters per test location) and representative irradiation lengths (≥ 0.6 m). VTR is being designed for high availability to achieve high fast neutron fluences, greater than 30 displacements-per-atom per year on materials. VTR’s design will support the use of self-contained internally circulating loops for sodium, lead, lead-bismuth eutectic, helium and molten salt coolants to support commercial technology development efforts. It will also support a rabbit system to enable rapid irradiation testing. Additional experimental positions throughout the core will be available.

VTR project management follows DOE regulations on capital assets acquisition. The first critical decision, reached in February 2019, confirmed the need for a fast spectrum irradiation capability and authorized the analysis of alternatives and development of a conceptual design for the preferred strategy. The second critical decision, signed in September 2020, approved the preferred strategy and established cost and schedule ranges based on the conceptual design of a sodium-cooled fast reactor. VTR is starting the preliminary and final design phase of the project that will lead to the next critical decision that will establish a cost and schedule performance baseline and authorize the start of construction. VTR is being designed using recently approved U.S. NRC advanced reactor design criteria and risk informed performance-based safety basis development guides.

VTR will support re-establishing parts of the U.S. nuclear supply chain and will modernize the design and construction process through the use of digital engineering design tools and integrated requirements management systems. The target date for the start of operations is the end of 2026 with 5 years of contingency.

VTR fills the only major gap in the U.S. nuclear energy research and development infrastructure. Coupled with the other major infrastructure, such as Advanced Test Reactor (ATR), High-Flux Isotope Reactor (HFIR), Transient Reactor Test Facility (TREAT), fuels and materials fabrication and characterization and post-irradiation examination facilities, VTR will support nuclear energy innovations for decades to come.

Speaker: Jordi Roglans-Ribas (Argonne National Laboratory)
• 179
MULTIPURPOSE RESEARCH FACILITY MBIR AND POLY FUNCTIONAL RADIOCHEMICAL COMPLEX (R&D COMPLEX) AS A UNIQUE RESEARCH PLATFORM

Development and justification of innovative nuclear power technologies, spent nuclear fuel management and closed nuclear fuel cycle are the main issues that nuclear power industry faces nowadays.

Solving those tasks will be possible only with the help of modern research infrastructure permitting to conduct a full-scale experiments and to model various operating conditions. Russia is committed to create such an infrastructure in the form of multipurpose fast sodium research reactor MBIR and Poly Functional Radiochemical Complex (R&D Complex), which will help to resolve the scientific challenges faced in the 21 century.

Russia’s decision to combine the Poly Functional Radiochemical Complex (R&D Complex) and MBIR reactor, which are both under construction at the RIAR site in Dimitrovgrad significantly raises the capabilities of advanced research in nuclear fuel, nuclear fuel cycle and SNF treatment technologies testing.

Establishment of an International Research Center based on reactor MBIR and R&D Complex in the form of a consortium with international participation is an example of international cooperation in nuclear area.

This paper provides information on MBIR and R&D Complex.

The emphasis of the paper is on the structure, terms and conditions of the consortium as well as on the current status of the project.

Speaker: Alexander Zagornov (Rosatom)
• 180
VERSATILE TEST REACTOR: CONCEPTUAL CORE DESIGN OVERVIEW

The Versatile Test Reactor (VTR) is a reactor under development in the United States of America to provide a very high-flux fast neutron source that will support the development of advanced reactor technologies. This reactor will accelerate the irradiation testing of advanced nuclear fuels, materials, and potentially other components. The development efforts are structured in several phases to mature the design, and the conceptual design phase was recently completed. The VTR core design developed in this phase will be presented and discussed in this paper.

The conceptual design for the VTR is a 300 MWth pool-type sodium-cooled fast reactor. The core contains a total of 313 assemblies, 66 of which are fuel drivers, and 10 are representative test locations. Ternary metallic fuel is used in the driver fuel and allows achieving a peak fast flux of about 4.3x1015 n/cm2-s in the central test location, corresponding to material damage rate in excess of 50 dpa per year. This will be the highest fast flux level and material irradiation rate of any reactor, now existing or under-development. The irradiation conditions offer large irradiation volumes with very high flux levels, as well as experimental flexibility through the use of cartridge loops. Cartridge loop experiments allow using a self-contained coolant, different from the reactor coolant, to provide prototypical irradiation conditions pertaining to other types of advanced reactors. The reactor can accommodate simultaneous cartridge loop experiments in up to five locations. In addition, VTR has a rabbit facility that permits insertion and removal of irradiation samples during operation.

The VTR conceptual core has been designed relying on past US expertise and demonstrated technologies. It is able to achieve the above-mentioned flux level under nominal conditions while remaining within acceptable thermal-hydraulic conditions and retaining an extremely favorable passively safe behavior. Additional details, design characteristics and performance characteristics will be discussed in the full paper, providing a comprehensive overview of the VTR core.

Speaker: Florent Heidet (Argonne National Laboratory)
• 6.2 Thermal Hydraulics

Chairs: Nastasya Mosunova and Vladimir Kriventsev

• 181
Possibility of Simulating Natural Circulation in Fast Neutron Reactors Using a Light Water Test Facility

The paper evaluates the possibility of modeling the heat transfer phenomena in a liquid-metal coolant using a light water test facility. It considers the natural circulation of the coolant in the upper plenum of the fast-neutron reactor. A large nuclear power reactor (like the BN-1200 project) was selected as a reactor installation to be modeled. As the referent one was accepted the IPPE B-200 facility.
To validate the model, the similarity theory and the “black box” method were used. The paper uses the experience of a number of researchers in this field, in particular, the accepted assumptions which do not result in serious loss in modeling accuracy. The governing criteria of similarity were estimated based on the fundamental differential equations of convective heat transfer, so were the conditions under which it is possible to model sodium coolant by using light water with adequate accuracy. The paper presents the scales of the parameters used for the model-reactor comparison.
The introduction presents the paper purpose, considers the relevance of this topic, the utilized approaches – the similarity theory and the “black box” method, their limits to applicability. The general restrictions of the water test facility structural features are provided.
The first section provides the governing criteria derivation from the fundamental equations.
The second section includes obtaining the scales of the parameters.
The third section presents estimating the water test facility characteristics depending on its geometric scale. The conclusion about the possibility of the water-based modeling the liquid-metal coolant behavior is presented.
The paper includes 2 pictures, 2 tables, 23 references.

Speaker: Mr Dmitrii Uralov
• 182
Recent thermal hydraulic studies of Gas Fast Reactor demonstrator ALLEGRO

The helium cooled high-temperature fast-spectrum reactor (GFR) with closed fuel cycle is one of the six GEN IV reactors selected by the Generation IV International Forum (GIF) to be developed for the foreseeable future. The European reference concept of the GFR technology is a unit with an envisaged power of 2400 MWth, which is currently in the pre-conceptual design phase. Prior to the building of the full scope facility the viability of the GFR technology will be proven by means of the ALLEGRO demonstrator with an envisaged thermal power of 75 MWth. The ALLEGRO development is led by the V4G4 Centre of Excellence consortium associating research organizations, companies and laboratories from Czech Republic (UJV Rez), France (CEA) Hungary (MTA-EK), Poland (NCBJ) and Slovakia (VUJE). One of the key tasks of ALLEGRO is to test the new ceramic refractory fuel for the industrial version of GFR2400. In this paper the latest outcomes of thermal hydraulic calculations of ALLEGRO are summarized. First, the work that has been done under the EU VINCO project is reviewed. It was carried out by V4G4 consortium aiming to transfer the GFR technology know-how from the CEA to the V4G4 and to establish the platform for continuation of the ALLEGRO demonstrator development. It comprises the methods, specific calculations and outcomes of the ALLEGRO thermal hydraulic benchmark which were carried out by the V4G4 partners using the CATHARE, RELAP and MELCOR codes. Based on the benchmark future experimental program is proposed using helium-cooled experimental facilities the S-ALLEGRO build in Czech Republic and the STU helium loop operating in Slovakia. Subsequently, a short summary of a recent work is presented, in which the hot duct break scenario is studied for the two and the three-loop ALLEGRO versions. The preliminary results of this analysis showed that the three-loop ALLEGRO has better cooling performance in case of hot duct break. Finally, the gas mass flow distribution in two parallel geometrically identical pipes is investigated, when they are heated with different power and when they have the same pressure loss. The results show that the pipe (or a closed subassembly in the reactor core) heated with higher power usually has lower coolant mass flow rate, which deteriorates the cooling capabilities of the subassemblies in a real reactor.

Speaker: Mr Boris Kvizda (VUJE, a.s.)
• 183
Experimental modeling of a fuel element simulator vibration in a coolant flow

The elements of the fuel assembly including individual fuel pins are affected by the coolant flow. This can lead to mechanical vibrations. The cyclic loading of the fuel element cladding material accompanying these vibrations causes an additional effect on the fuel element material. This can cause the damage of the fuel cladding material, especially in its contact with the spacer elements. The natural frequency of vibrations of a fuel element in a liquid flow depends on various factors. One of the main ones is the method of fastening the fuel element. Usually, the lower end of the fuel element is rigidly sealed, while the upper one either rests or moves freely in a small gap. The added mass of the liquid has a noticeable effect on the frequency of rods oscillations.
In the contribution the results of experimental study of the oscillations of fuel rod simulators in fuel assembly models with a lead-bismuth eutectic (LBE) coolant flow are represented. Measurements were carried out on a 7 pins model of the fuel assembly. An annular channel with an equivalent diameter equal to the hydraulic diameter of the fuel assembly model was also used for measurements. The Reynolds number in the experiments was varied in the range of 5·103 – 4·104. In addition to the lead-bismuth coolant, a water coolant was also used, which made it possible to compare the results and determine the features of the vibrations of the fuel element simulator in liquids with different densities. In addition, for the water coolant, the distribution of the averaged and pulsating component of the liquid velocity was measured both along the axis of a single fuel element simulator and behind its head.
During the experiments, the following parameters were controlled and measured:
- temperatures of various elements;
- temperature of the coolant at the inlet and outlet of the test section;
- coolant flow rate at the test section inlet;
- profiles of displacements (pulsations) of the fuel element simulator.
As a result the data on the dynamic stability of the fuel element simulator in lead-bismuth and water coolants, data on the regularities of vibrations of the fuel element simulator, depending on the method of its fastening, the length and flow rate of the coolant, were obtained.

Speaker: Prof. Nikolaay Pribaturin (IBRAE RAN, Novosibirsk division)
• 184
DEVELOPMENT OF COOLANT VOIDING MODEL FOR FAST REACTOR CORE

Unlike thermal reactors, LMFBR core is not in the most reactive configuration. Any undesirable event may raise the reactivity in the core and can result in increase of the reactor power. Liquid metals used in LMFBRs have high boiling points and there is considerable margin between normal operating temperatures and their boiling points. By design, liquid metals used as coolant are not expected to boil under any normal operating conditions of the reactor. However, in case of loss of flow accidents due to pipe rupture, pump failure along with the failure of reactor shutdown systems, boiling of the coolant in reactor core is possible. Such accidents are termed as Unprotected Loss of Flow Accidents (ULOFA). Boiling of coolant in reactor core can also be caused by uncontrolled increase in power. Such accidents are termed as Transient Overpower Accidents (TOPA).

Improved understanding of the mechanics of sodium ejection in case of ULOFA or TOPA is of critical importance for an LMFBR safety. A reduction of sodium density can result in either a positive or a negative reactivity, depending on the location and extent of the vapour void. Therefore, an accurate description of the voiding process with respect to space and time is necessary. NaBOIL, which stands for Natrium Boiling Onset Influence in LMFBRs, is a code developed for predicting boiling behaviour of liquid metals in channels of a reactor core. It calculates the heat transfer from the fuel pin to the coolant until at some location (in coolant channel) and time the coolant reaches a specified superheat. At this point, onset of boiling occurs. The stages of bubble growth are approximated by a thin bubble assumed to occupy the entire coolant channel area, except for the liquid film remaining at the clad wall. The coupled solution of the energy and hydrodynamic equations of the coolant, and the heat transfer equations of the fuel pin are then continuously solved during the voiding process. The main purpose of this model are to predict the extent and rate of voiding that can be used for voiding reactivity calculations and to predict the heat removal from the cladding surface after the onset of boiling, for fuel and cladding temperature calculations.

Speaker: Dhrumil Ganatra
• 185
ALFRED FLOW BOCKAGE ANALYSIS

In the context of Lead-cooled Fast Reactor development and safety assessment, the flow blockage in a fuel sub-assembly is considered among the most relevant issues to be addressed. Hence, the event shall be postulated assessing its consequences, also considering that grid-spaced fuel assemblies could partially mitigate the occurrence of sudden blockages with respect to wire-spaced fuel assemblies.
The Advanced Lead-cooled Fast Reactor European Demonstrator (ALFRED) is a 300 MWth pool-type reactor aimed at demonstrating the safe and economic competitiveness of the Generation IV LFR technology. The ALFRED design, currently being developed by ANSALDO NUCLEARE and ENEA in the frame of the FALCON Consortium, is based on prototypical solutions intended to be used to boost the DEMO-LFR development.
Within the scope of FALCON consortium and in the frame of investigating the thermal-hydraulics of the average ALFRED FA, a CFD computational model is built looking for the assessment of its thermal field in nominal flow conditions and when affected by a blockage. Starting from the experience in this kind of simulations and in experimental work, the whole model of the ALFRED Fuel Assembly is first presented and calculation of flow and temperature field in nominal conditions is carried out. RANS simulations of idealized blockage scenarios adopting three different spacer grid locations (under the active length, at half active length, above the active length). Results showed that the most likely blockage in the lower grid positioned before the active region do not perturb the temperature distribution in the fuel assembly, while the ones at the central grid may have strong consequences and lead to a clad temperature peak behind the blockage with possible clad failure. In particular, the CFD analysis on the ALFRED FA suggested to install the spacer grids – or at least the first one – in the not-active region to avoid any clad failure due to an internal blockage.

Speaker: Dr Ivan Di Piazza (ENEA Brasimone R. C.)
• 186
CFD Simulations on a hexagonal 61-pin wire-wrapped fuel bundle with STARCCM+ and comparison with experimental data.

In this work, several CFD simulations of the hexagonal 61-pin fuel bundle replica of the Thermohydraulic Research Lab at Texas A&M University were performed. This fuel assembly geometry is a helically wire wrapped bundle of rod pitch-to-diameter ratio of 1.89 and helix pitch-to-diameter ratio of 29.93, as in Sodium-cooled fast breeder reactors (FBR). The experimental activity has produced a large set of data that is compared against simulations. The purpose of the present study is to optimize computational fluid dynamics and compare the code predictions of pressure drops (friction factors) against experimental data and available correlations. The friction factor is of paramount importance in determining reactor features like pump specification and safety limits.

The commercial software STARCCM+ Version 14.04.013 was chosen to perform the simulations. Four different turbulence models were utilized to compare with the experimental data: k-epsilon (standard and realizable) and k-omega (standard and SST). The code predictions were also compared with the upgraded Chen and Todreas detailed correlation (UCDT).

For the geometry under study, the laminar to transition (RebL) and transition to turbulent (RebT) Reynolds numbers are RebL=494 and RebT=13,554. The experimental data was generated within Reynolds ranging from 439 to 13,766. Therefore, the simulations were focused in the transition regime, although there were simulations run at laminar Reynolds numbers.

The Incompressible Reynolds-Averaged Navier-Stokes Equation (RANS) with an eddy viscosity turbulence model was used in the calculations. Two meshes were utilized. The first mesh was used to perform simulations with these turbulence models; however, a second mesh was created with the intention of enhancing the results of k-ω SST.

Within the laminar regime, no turbulence model was needed. The simulation results exhibited a good comparison with the experimental data with an error of 2.81 % and the UCTD, with and error of 7.60%.

At the transition regime, the k-omega standard model with the first mesh produced the best prediction of the experiment (all simulation values were within the uncertainty interval) and the UCTD, while the two k-epsilon models overpredicted the friction factor. Regarding the k-omega SST model, results obtained with the first mesh were more distant to the experiment than the standard model. A better prediction was obtained using the second mesh, because this mesh had an average Y+ less than 1, which was not the case of the first mesh. Using Y+<1 is a requirement for the SST to accurately predict pressure and velocity mean field.

Speaker: Octavio Bovati (Texas A&M)
• 187
Progress in system thermohydraulic code HYDRA-IBRAE/LM models development for fast reactor simulation

The system thermohydraulic code HYDRA-IBRAE/LM is designed for the simulation of non-stationary thermohydraulic processes in liquid metal and water circuits of fast reactors under normal operating conditions, anticipated operational occurrences and accidents. The code uses a two-fluid model in all flow regimes except for dispersed annular flow, where a three-fluid model is applied. Besides advanced mathematical models, the code has advanced pre- and postprocessor and utility for performing multivariative calculations, uses MPI and OpenMP parallelization. The code is being developed in “Codes of New Generation” subproject of “Proryv” project.
New models described in the present paper expand the code applicability and reduce the degree of conservatism in the nuclear power plants safety assessment.
One direction of the code development is the improvement of dispersed phase transport model. The new model accounts for the flow parameters time dependence and allows performing correct calculation of the bubble diameter necessary for the description of heat transfer and interfacial friction. The interfacial area density equation describes its dynamics and determines bubble diameter in one-group approximation. A more sophisticated heterogeneous multi-group model allows describing bubble dynamics and processes of separation and stratification.
The improved post dryout model was implemented for water coolant. Post dryout flow is treated as steam-water annular two-phase flow. In considered conditions deposition of droplets from the flow core on the hot tube wall takes place. The new model describes post dryout heat transfer in the pressure range 3-16 MPa with high accuracy.
The turbine model was developed and implemented in the code. The model describes the thermal-hydraulic processes taking place as the energy of superheated steam transforming into mechanical work. The model is based on the universal thermodynamical relations and self-consistently calculates enthalpy and pressure drops at the turbine stages. This approach allows determination of the thermal-hydraulic parameters of the coolant (temperature, density, pressure, enthalpy, entropy, mass fraction of water) in all elements of the turbine for a given load.
The work was performed on enabling the description of water behavior at supercritical parameters which is extremely important for correct modeling of the experiments where supercritical water was used.
Performed validation calculations made it possible to refine correlations used in the closure relations.

Speaker: Alexander Palagin (Nuclear Safety Institute of the Russian Academy of Sciences (IBRAE RAN))
• Coffee Break
• Panel: Innovative Fast Reactors: Designs, Applications, and Fuel Cycles

• Thursday, April 21
• Plenary 3. Panel Discussion on National Programmes and Visions

Chair: Ms Amparo Gonzalez

• Coffee Break
• Poster Session
• 188
ALFRED DHR system scaling verification and numerical pre-test analysis

Passive safety systems are used in generation III+ evolutionary reactors and in generation IV advanced reactor designs, especially for the decay heat removal following an accidental event. These systems allow with one or more loops the heat transfer from the primary system to the external environment through the natural circulation of fluids or through boiling and condensation phenomena. A limitation of these systems is the lack of intrinsic mechanisms that allow the passive control of the power removed, and this is particularly important for reactor designs that involve the use of liquid metals as a coolant, as the solidification temperature is always higher than the temperature of the ultimate heat sink. Under these circumstances, there is the possibility that the primary system coolant may freeze in the colder regions of the system, opposing to the natural circulation of the fluid and in eventually inducing mechanical stresses due to the volume variation between liquid and solid state. Ansaldo Nucleare has patented a passive safety system for the ALFRED reactor that allows to control the power removed to the environment by making use of non-condensable gases placed in strategic positions of the safety system. The DHR system consists of 3 loops, each connected to one steam generator through the feedwater and the steamline. These circuits have an isolation condenser immersed in a pool and connected to an expansion tank. During safety system operation, the non-condensable gases are passively transported in the system between the tank and the isolation condenser proportionally to the decay heat, degrading the heat transfer in the condenser and strongly delaying the onset of solidification in the primary coolant. This paper is based on the scientific effort dedicated through the European H2020 PIACE project and reports the scaling verification of the decay heat removal system to be carried out at the SIRIO experimental facility. Pre-test analyses performed by means of the RELAP5-3D system code are presented, assessing the applicability of an existing facility configuration to the revised design of the DHR system of ALFRED. The results show how the experimental facility is able to represent the most important phenomena underlying the operating principle of the system such as pressure behaviour, noncondensables gas transport and coolant temperature control above solidification point.

Speaker: Mr Marco Caramello (Ansaldo Nucleare)
• 189
Cognitive Information Retrieval Based on Ontological Model of Knowledge Representation

The technologies of information retrieval in a database with full-text semantic indexing are considered. The information retrieval process is considered as a cognitive-oriented process. The semantic image of the document context is presented as an ontology. An ontology is defined as a set of three interconnected systems (functional, conceptual and terminological), on which the operation of comparing elements of different systems at the level of signs is defined. A functional system (a system of tasks, objects, processes, properties of the subject area) represents objects and situational relationships between them in the context of the target activity. The objects of the conceptual system are stable concepts, and the set of relationships is limited to generic and associative relationships. The terminological system reflects the properties of a natural language at the level of signs – terms for which relationships of equivalence and inclusion, as well as linguistic relationships, are specified. The semantics of the document are represented as a multi-meta-hyper-graph of the ontology. Such graph in the nodes contains named entities (concepts, names, values, etc.), and in the edges – typed relationships, extracted from the text also taking into account the location of them. This made it possible to build up mechanisms (algorithms) for a new type of information retrieval – searching semantic dependencies and neighborhoods within the text of a document or a selection of documents. Such cognitive information retrieval is considered as a search for a path on a multi-meta-hyper-graph of an ontology dynamically formed on the basis of ontological images of found documents or their fragments. A Cognitive subject tree is used to fix and manage search directions. A Cognitive subject tree is a hierarchically ordered structure of a personified representation of a subject area (project, task). Such a structure integrally identifies tasks / knowledge, since each structural and semantic component (node) of the tree includes not only keywords, but also relevant documents texts, query expressions, indexes of classifiers, etc. The prototype of the information retrieval system has been developed and experimental databases have been created.

Speaker: Nikolay Maksimov (National research nuclear university (Moscow Engineering Physics Institute))
• 190
COMPLEX RADIATION AND HYGIENE STUDIES OF RADIATION IMPACT FACTORS ON PRODUCTION PERSONNEL, MIXED NITRIDE URANIUM-PLUTONIUM FUEL FOR FAST NEUTRON REACTORS

Тhis work is carried out in order to assess the compliance of the radiation protection of personnel working at the complex experimental installations of JSC «SСhE» with the requirements of the national radiation safety standards to limit the generalized risk of potential exposure and the IAEA recommendations for not exceeding the control level of the minimum significant radiation risk.
On the basis of monitoring the dynamics of the ambient dose rate equivalent (ADER) of photon and neutron radiation ADER at the workstations of the complex experimental installations 1 and 2, the regularities of the dose formation have been studied. Doses of external exposure of personnel were estimated. In accordance with the recommendations of the ICRP and the IAEA, the radiation risks of personnel were assessed.
The report will present the estimates obtained for the personnel working at the complex experimental facilities # 1 and # 2 on the external exposure doses for gamma radiation and neutron radiation. The annual expected effective dose of internal exposure of personnel will be estimated, and the level of the minimum significant radiation risk will be calculated.
The results obtained and the developed methods will be used to ensure the radiation safety of personnel during the transition from experimental installations to pilot industrial implementation of the technology for the production of mixed uranium-plutonium nitride (MNUP) fuel.

Speaker: Prof. Alexander Samoylov (Burnasyan Federal Medical Biophysical Center of Federal Medical Biological Agency,)
• 191
Conceptual Core configuration for increasing Power of Fast Breeder Reactor to 40 MWt

Fast Breeder Test Reactor (FBTR) in India is designed for 40 MWt Thermal 13.2 MWe. At present FBTR is operating at 32 MWt with 56 fuel sub assemblies (FSA) of 48 Mark I and 8 MOX type fuel sub assemblies. Mark I FSA are of Pu-U Carbide fuel with 70% Pu and MOX FSA are of PuO2 (44%) and UO2 (56%). Due to constraint on minimum shut down margin of 4200 pcm, the core could not be expanded and hence the power could not be increased to design power. A conceptual core configuration has been suggested and safety analysis was being carried out, by introducing four poison sub assemblies, Boron SA, with 50% B10 concentration, in the second ring, which would enable to expand the core and increase the power to the design power of 40 MWt and at the same time minimum criterion on shut down margin also will be met. The envisaged core for 40 MWt comprises 70 numbers of Mark I FSA.

Speaker: Mr RAGHUKUMAR Ganapathy (Indira Gandhi Cenre for Atomic Research)
• 192
Design, manufacturing and transportation of high capacity High Level Liquid Waste Storage tanks

High level liquid Waste (HLLW) storage tanks with large storage capacity weighing a few tens of MT are proposed to be used in Fast Reactor Fuel Reprocessing Plant (FRP) of Fast Reactor Fuel Cycle Facility (FRFCF), Kalpakkam to store the HLLW. Six years storage capacity is envisaged for allowing Ru106 to decay sufficiently before sending the HLLW for vitrification.These tanks have lot of internals such as cooling coil (2 Km piping with about 600 welded Joints) arrangement with ballast tanks positioned inside the vessel to remove decay heat. Presence of highly radioactive fields coupled with highly concentrated nitric acid and remote chance of maintenance after commissioning demands well proven design, material selection, fabrication, meticulous QA practices and transportation methodology to be employed. This presentation dwells upon the design, manufacturing aspects and transportation methodology used for the fabrication of such critical high capacity tanks.

HLW tank, weighing around 67 MT, is a Horizontal cylindrical tank with capacity of 212 m3 fabricated using torispherical formed heads with 4.7m OD and an overall length of 13 m. Material of construction for all the components of the tank is AISI 304L with tailor made chemical composition and other supplementary requirements. The welding process used was manual & semi-automatic GTAW with ER308L filler wire. Prior to fabrication, a detailed study was made to decide the simultaneous fabrication of individual components & the assembling sequence from the point of fabricability & inspectability in tandem meeting project delivery schedule. Based on the above a manufacturing sequence cum quality assurance plan. Accordingly, the individual shell courses, Baffle plates, ballast tanks, saddle supports, cooling coils & air spargers, dished ends with outlet piping, machining of nozzles, manholes, deep feed piping of all tanks were fabricated simultaneously. Production test coupon has been put in place in qualifying all special processes such as Post forming heat treatment. Assembly of individual components including assembly of closure dished end to shell were carried out as sub-assemblies in sequence and the final assembly was completed & tested. Because these were over dimensional consignments, these tanks were transported & unloaded after a detailed road survey, as per approved methodology.

A well designed, thought provoking and proven design, manufacturing sequence & QAP and its effective implementation resulted in the successful fabrication of such critical and complex tanks.

Speaker: Mr Kumst Dhananjeya
• 193
Eddy Current Flow Meter flow rate measurements in liquid Sodium at the SUPERFENNEC loop

The Eddy Current Flow Meter (ECFM) is a robust and reliable inductive sensor for measuring the flow rate of liquid metals. Since there is no direct contact between sensor and liquid metal, it can be used in chemically aggressive environments and at very high temperatures of up to 600 °C. This allows the ECFM to be deployed, for example, as part of the safety instrumentation in liquid metal-cooled fast reactors directly above the subassemblies, in order to continuously monitor the flow rate of the coolant and to detect coolant blockages. In this paper we present the measurement results that were obtained with a high temperature prototype of the ECFM at the SUPERFENNEC Sodium loop at CEA Cadarache in France. There, we were able to evaluate the performance of the ECFM between temperatures of 200 °C to 400 °C. In addition to the measurement results, we will discuss results of related numerical simulations and give a detailed description on the construction and choice of materials of the high temperature prototype.

Speaker: Dr Gunter Gerbeth (Helmholtz-Zentrum Dresden - Rossendorf)
• 194
Evaluation of EPDM and Silicone rubber compounds for application in Reprocessing Plant

Many elastomers seals are used in the nuclear industry. Among these elastomers, ethylene propylene diene monomer (EPDM) and silicone rubbers have excellent radiation stability. Both the rubbers can be used for gasket and O-ring application in Reprocessing Plants. To study the suitability of these rubbers for application in the plant, EPDM rubber compound and silicone rubber compounds were prepared and test slabs were fabricated. These rubber compounds were tested for their mechanical properties. Two test slabs of EPDM rubber compound were taken. First test slab was irradiated in gamma chamber followed by exposure to nitric acid (6 M) and the second test slab was exposed to nitric acid followed by irradiation in gamma chamber. Samples were taken out of gamma chamber at regular intervals of time and mechanical properties were tested. Similar procedure was adopted for silicone rubber compound also. The mechanical properties of both the rubber compounds were found to degrade with radiation. Elongation at break of the EPDM rubber compound decreased to 50% of its initial value at a dose of 1 MGy. For silicone rubber, an identical decrease was found even at a dose of 0.1 MGy. When silicone rubber compound was exposed only to radiation, elongation at break decreased to 50 % of its initial value at a dose of 1 MGy. Hence, for seal application in radiation atmosphere alone, silicone rubber can be used up to a dose of 1 MGy and for seal application in combined radiation and acid atmosphere, EPDM rubber can be used up to a dose of 1 MGy.

Speaker: Ananthasivan Krishnamurthy (IGCAR, DAE)
• 195
Experiment-calculated method for determination of prompt neutron lifetime in fast metal cores intended for verification of neutron transfer simulation codes

Here described a method for determination of mean prompt neutron lifetime in fast metal cores during critical experiments held in RFNC – VNIIТF using assembly machine FKBN-2. The evaluation of derivative using experimental dependence between asymptotic decrease coefficient and core parts gap α(Н) was proposed further to determination of the delayed critical state of the core. The value characterizes the transient prompt neutron process in the core and accurate within coefficient determines the mean prompt neutron lifetime in the system. By-turn the coefficient may be calculated using contemporary neutron transfer simulation.
The experimental results of and routine criticality experiments data may be used for verification of computer codes and cross section databases.
Approbation of the method was held using fast metal cores of U and Pu in different mass ratio. The experiments were performed on assembly machine FKBN-2 leaded by time-correlated measurements. Benchmark experiment modeling was carried out and Monte-Carlo simulation was done for critical and time-correlated experiments using different cross section databases.

Speaker: Sergey Andreev
• 196
Experimental investigation of the fluid-structure interaction effect between adjacent equipment supports in a fast reactor

The equipment supports are in constricted arrangement in the main vessel of the fast reactor. Under the condition of earthquake, equipment supports may sustain damage caused by the interaction between equipment supports and fluid, therefore, the evaluation of the fluid-structure interaction effect is an important aspect of the structural safety assessment of fast reactors. Using the added mass to assess the fluid-structure interaction effect is a more common method. ASME standard gives the added mass’s formula for a single cylinder immersed in an infinite fluid domain. However, this formula is not suitable for the complex arrangement of equipment supports in the annulus region of main vessel. To investigate the added mass of equipment supports in fast reactor, in this paper, a number of experiments are carried out. A simplified and scaled model of fast reactor was designed, containing a main pump support cylinder, two intermediate heat exchanger (IHX) support, and two independent heat exchanger (DHX) support. Through the modal experiment and the sine wave experiment of the shaking table, the circumferential and axial pressure distribution data under different arrangements are recorded. For the purpose of assessing the fluid-structure interaction effect between equipment supports and fluid, the additional mass matrix of the support cylinders is derived according to the experimental results. The experiments provide a reference and basis for the arrangement of the equipment supports in the annular region of the fast reactor and the correction of the additional mass formula.

Speaker: Mr Dexuan Duan (North China Electric Power University)
• 197
INVESTIGATION OF THE SOLUBILITY OF ACTINIDE FLUORIDES FOR THE CHOICE OF A SALT SOLVENT FOR A MOLTEN-SALT REACTOR-BURNER OF MINOR ACTINIDES

Characteristics of the molten-salt reactor-burner (MSR-burner) of minor actinides (MA), which are concentrated in spent nuclear fuel of power reactors, depend significantly on the physical-chemical properties of the fuel composition. In particular, the MA transmutation efficiency is mainly determined by the concentration of actinide fluorides in the molten-salt fuel composition [1]. In this regard, the theoretical and experimental research of the actinide fluorides solubility in the molten-salt solvent to justify the choice of the molten-salt fuel composition is a relevant task.
In Russia, fluoride salt systems based on LiF-BeF2 [2] and LiF-NaF-KF [3] are considered as basis of molten-salt fuel composition. The purpose of this work is an experimental determination of the solubility of actinide fluorides and their simulators in these fluoride salt systems.
This report presents the procedures for the experimental research of the actinide fluorides solubility in fluoride salt systems, such as thermal analysis method by cooling curves, differential scanning calorimetry, elementary analysis. The working out of these techniques was carried out by the determining individual solubility of NdF3 and CeF3 in salt solvents under investigation. Satisfactory agreement of experimental data with literature ones was obtained.
The results of calculating оf different MSR-burner versions and the analysis of literary data were used for determination of characteristic compositions of molten-salt fuel mixture. Temperature dependencies of individual solubility of actinide fluorides and simultaneous solubility of actinide fluorides and their simulators in investigated molten salt solvents were obtained, and the elemental composition of molten salt samples was determined.

Speaker: Полина Санникова
• 198
JUSTIFICATION OF CRITICAL EXPERIMENTS ON STAND FKBN-2 TO VERIFY NEUTRON-PHYSICAL SOFTWARE FOR CALCULATIONS OF THE MOLTEN-SALT REACTOR

In order to reduce the long-term potential hazard of waste from the reprocessing of spent nuclear fuel from thermal reactors and to increase the enviroment attractiveness of nuclear power in our country, work is underway to create a molten-salt reactor-burner of minor actinides (MSR-B). The first stage on this path is the creation of an investigative molten-salt reactor (IMSR) for testing key technological solutions for the full-scale MSR-B. A carrier fuel based on the two-component LiF-BeF2 system was chosen as the base for the IMSR.
At present, there are no computer software certified for neutron-physical calculation of a reactor with fuel based on molten salts of the LiF-BeF2 type, adopted for design work on the IMSR. There are also large uncertainties in the choice of neutron constants that exceed the permissible limits for the accuracy of calculations when justifying of the choice of specific core parameters when performing design work.
Experiments with critical multiplying systems that simulate the IMSR core in terms of the composition of materials and spectral characteristics are required to justify and subsequently certify neutron-physical software as applied to the IMSR project. One of the possible instalations for carrying out such experiments is the stand for critical assemblies FKBN-2. An important task is to justify the possibility of setting up critical experiments on this stand.

Speaker: Mr Mikhail Trapeznikov
• 199
RADIATION-HYGIENIC ASSESSMENT OF INTERNAL EXPOSURE FACTORS OF PERSONNEL WORKING AT EXPERIMENTAL FACILITIES IN THE PRODUCTION OF MIXED NITRIDE URANIUM-PLUTONIUM FUEL

The report presents results of studies of the physicochemical characteristics of radioactive workplace aerosols formed during the production of mixed nitride uranium-plutonium fuel (activity particle-size distribution, nuclide composition, lung absorption type, elemental composition, reactive properties in the air). Taking into account these characteristics, the dose coefficients were calculated and the annual committed effective doses of internal exposure of personnel were calculated.
Next impactors were used for activity particle-size distribution analysis: AIP-2, IPHRT (developed by SRC FMBC named by A.I. Burnazyan), an electric low-pressure impactor HR-ELPI, Andersen cascade impactor and others. At the study of morphological characteristics, we used a scanning electron microscope (SEM) LYRA-3 equipped with an X-ray microanalyzer (RMA) X-max 80. X-ray structural analysis was performed on an XRD-7000 X-ray diffractometer. Analysis to determine the mass fraction of nitrogen and oxygen was carried out on the LECO analyzer; measurement of the content of uranium and plutonium was conducted on the mass spectrometer "TRITON+". Lung absorption type assessment was performed by dialysis through membrane filters in a pulmonary fluid simulator.
The high reactivity of mixed nitride uranium-plutonium (MNUP) compounds causes instant oxidation of the thoracic fraction of MNUP fuel aerosols upon contact with air, however, the intake of MNUP into the body is possible orally as part of the extrathoracic fraction (particles of 100 - 500 μm in size which have oxide film emerging upon interaction with air and inhibiting further oxidation of nitride). Dissolution of these particles in gastric juice can lead to the release of the nitride core, followed by a rapid entry of radionuclides into the organs and tissues of the body through the gastrointestinal tract.

Speaker: Mr Andrey Karev (Burnasyan Federal Medical Biophysical Center of Federal Medical Biological Agency)
• 200
Revealing the dependencies of partitioning americium-241 and uranium using sorption technology based on solid-phase extractant TODGA

The objective of this work is to reveal the dependences of the separation of americium-241 and uranium using sorption technology based on the solid-phase extractant TODGA. In technologies for the purification of radioactive waste of low and medium activity levels with low contents of actinides, sorption and ion exchange methods are widely used due to their high selective. The required selectivity level, under extreme external conditions of the environment, is inherent in solid-phase extractants obtained using ligands. To extract americium-241 efficiently and separate from uranium, it is necessary to use solid-phase extractants based on N, N, N ', N' - tetraoctyldiglycolamide (TODGA), which will allow the return of fissile materials to the nuclear fuel cycle and reduce the hazard class of the removed radioactive waste. For the effective use of experimental modified TODGA sample, it is necessary to identify the dependences of the extraction of americium-241 and its separation with uranium using model solutions that simulate real radioactive waste. Therefore, the determination of the kinetic parameters of the extraction of americium-241 and its separation with uranium in the process of sorption processing using experimental modified TODGA samples is necessary for the most efficient separation process. In this work, the kinetic parameters and diffusion coefficients of americium-241 and uranium in the process of their sorption on experimental TODGA samples were determined. The kinetics of the process was investigated with the determination of the reaction rate, which was analyzed based on the dependence of the change in the concentration of the extracted element from the volume upon contact with the extractant on time, which depends on the value of the diffusion coefficient. The estimation of the reaction rate dependence on the value of the diffusion coefficients is carried out. Based on the dependence of the diffusion coefficients on the characteristics of the TODGA prototypes, the increased rate of sorption for americium-241 in comparison with uranium was determined by calculation. According to calculations of the diffusion coefficients of americium-241 and uranium in the experiments carried out with the samples under study, a prototype TODGA with the highest kinetic characteristics was determined.

Speaker: Mr Aleksandr Savelev (National Research Nuclear University MEPhI; Leading Research Institute of Chemical Technology (VNIIKhT))
• 201
VERSATILE TEST REACTOR: CORE SYSTEM DESIGN REQUIREMENTS TO SUPPORT ADVANCED REACTOR DEVELOPMENT

The Versatile Test Reactor (VTR) is a reactor under development in the United States of America to provide a very high-flux fast neutron source. This reactor will accelerate the testing of advanced nuclear fuels, materials, and other potentially irradiated components. As this reactor design effort is underway to support eventual construction and operation, a necessary step is the development of design requirements and objectives for all components and systems of the VTR. Such requirements are necessary in any engineering project to ensure the delivered product can perform its’ mission safely, while providing a means for integration of the various design teams working on interfacing systems, and providing a basis for successful project execution.

Many of the VTR nuclear core requirements are consistent with those found for typical reactor designs. For example, inherently safe feedback behavior is required as a part of the design, various fuel material performance limits shall be met during certain scenarios, and the occupational and public dose limits must be below site and regulatory limits. However, some requirements are unique to VTR due to the reactor being a test reactor that can support the needs of experimenters. For example, to support the needs of experimenters the reactor shall be designed to allow the use of any non-control/safety assembly position in the core as an un-instrumented experiment position. The reactor also must be able to accommodate multiple different materials and/or fuels under irradiations experiment campaigns at a time. This paper will present and discuss these reactor core system design requirements with the goal of disseminating these requirements to potential experimenters as early as possible and providing an example of design requirements application to a modern large engineering project.

Speaker: Adam Nelson (Argonne National Laboratory)
• 202
Assay of Waste drum based on Passive Neutron Counting Technique

Characterization of alpha emitting nuclide and other fission products in the radioactive waste generated in reprocessing plants is a regulatory requirement for their disposal. The assay of plutonium in the solid radioactive wastes could be carried out either using gamma spectrometry or neutron counting, depending mainly on the surface dose of the container. Presence of large amount of fission products renders the use of gamma spectrometry inappropriate due to the increased background radiation. In order facilitate the detection of plutonium in such instances, a passive neutron based assaying system for alpha bearing solid wastes has been designed and developed. The detector system has been fabricated in a semi-circular shape to assay the alpha bearing solid wastes in a 200 L capacity SS drum. This detector employs eight numbers of 3He neutron detectors embedded in High Density Poly Ethylene (HDPE). All the detectors were identical with an active length of 900 mm and a diameter of 50 mm filled with a gaseous mixture of 75% 3He and 25% Kr at 3 bar pressure. For deploying this system in an environment with reasonably high beta gamma background radiation, 25 mm lead shield in front of the detector was used.

Studies reveal that, the rotation of the drum during the assay improves the accuracy of the results. However, only rotation by supplementary angles was found to yield results with minimum error which was found to be in the range of ± 20%. Hence two measurements at 180º apart were found to be sufficient for satisfactory assay. This system was calibrated with plutonium source typically handled in fast reactor fuel reprocessing facilities. The minimum detection limit of typical of research reactor grade plutonium has been estimated to be 35 mg with a 99.9 % confidence level for the back ground prevailing in the site. It has been determined that a gamma radiation background of up to 5 mSv/h could be tolerated without loss of accuracy.

Speaker: Dr Sivakumar P
• 203
Comparative analysis of minor actinides transmutation in a molten-salt burner reactor based on LiF-NaF-KF and LiF-BeF2 salts

In Russia, research is actively underway to develop a specialized molten-salt burner reactor (MSR-burner) of minor actinides (MA) from spent nuclear fuel of power reactors. Two candidate fluoride salts, LiF-BeF2 [1] and LiF-NaF-KF, are considered as the solvent of the reactor fuel components.
The purpose of the present paper is to study MA transmutation in the MSR-burner based on selected salts in the equilibrium mode of reactor operation at different volumes of the core. The calculations were performed using PRIZMA+RISK software package developed at the "RFNC-VNIITF named after Academ. E.I. Zababakhin" [3,4].
The LiF-BeF2 salt has a low solubility limit of actinide fluorides, which leads to the need to feed the reactor with a significant amount of Pu and, consequently, to a low efficiency of MA transmutation. By reducing of volume of the active zone increases the consumption of Pu and reduces the efficiency of transmutation. In contrast to LiF-BeF2, LiF-NaF-KF eutectic is characterized by a relative high solubility of actinide fluorides. For a MSR-burner based on this salt, Pu is needed mainly for starting; in the equilibrium mode reactor consumes only MA. In this case, the maximum efficiency of MA transmutation can be achieved in a wide range of core volume: from 2 m3 to 30 m3 with a concentration of actinide fluorides from 17 to 10%, mol., respectively.

[1] Ignatiev V., Feynberg O., I. Gnidoi, et al. Molten salt actinide recycler and transforming system without and with Th-U support: Fuel cycle flexibility and key material properties. Ann. Nucl. Energy, 2014, v.64, p.408-420.
[2] Lizin A.A., Tomilin S.V., Gnevashov O.E., et al. PuF3, AmF3, CeF3, and NdF3 solubility in LiF-NaF-KF melt. Atomic energy, 2013, v.115, No.1, p.11-16.
[3] Zatsepin O.V., Kandiev Ya.Z., Kashaeva E.A., et al. Calculation for the VVER-1000 core by the Monte-Carlo method implemented in the PRIZMA code. Voprosy atomnoy nauki i techniki. Serija: Jadernye konstanty, 2011, No.4, p.64-73.
[4] Modestov D.G. The RISK-2014 code to solve nuclear kinetics problems, RFNC-VNIITF preprint No.243, Snazhinsk, 2014.

Speaker: Mikhail Belonogov (RFNC-VNIITF named after academ. E.I. Zababakhin)
• 204
Controlled thermonuclear fusion: potential role of a joint (Th-U-Pu) nuclear fuel cycle

This paper aims at finding solutions of so important problems of nuclear power as decreasing the scope and the number of technological operations, as well as enhancing the proliferation resistance of fissile materials in nuclear fuel cycle by means of minimal changes in the cycle. The method is including fusion neutron sources with thorium blanket into future nuclear power system. In addition to production of light uranium fraction consisting of 233U and 234U, high-energy 14-MeV neutrons emitted in the process of fusion (D,T)-reaction can generate 231Pa and 232U through (n,2n)- and (n,3n)-reactions.
It has been demonstrated that admixture of 231Pa into fresh fuel composition can stabilize its neutron-multiplying properties thanks to two well-fissile consecutive isotopes 232U and 233U, products of radiative neutron capture by 231Pa. Coupled system of two well-fissile isotopes can allow us to reach the following goals: the higher fuel burn-up and, as a consequence, the longer fuel lifetime; the shorter scope and the lower number of technological operations in nuclear fuel cycle; the better economic potential of nuclear power technologies. Such a fuel cycle presumes shifting from 235U to 233U as more attractive fuel material for thermal nuclear reactors. Uranium component will be protected from unauthorized proliferation by the presence of light uranium isotope 232U. The use of well-mastered traditional uranium-based fuels in power LWR will be preserved. The idea suggests fresh fuel fabrication for power LWR without applications of isotope separation technologies.

Speaker: Dr Evgeny Kulikov (National Research Nuclear University MEPhI)
• 205
Determination of the metallic and oxide compounds in models based on metallic uranium containing uranium dioxide, metallic neodymium, cerium as well as neodymium and cerium oxides

To determine uranium in the metallic phase in the presence of uranium oxide there is a reliable, so-called “bromine method”, which implies a metallic-oxide mixture treating in the bromine ethyl acetate solution. However, analogous manipulations with rare earth metals and their oxides do not provide such reliable data. Reduction melting of oxides in a graphite crucible with the melt composed of additional metals is another method, which allows determining the total amount of oxygen bonded in the sample. Together with the common chemical analysis of desired elements and using two aforesaid methods we obtain an algorithm of definite manipulations that provide the relation of metallic and oxide phases of different metals in the samples under study. In the present case, the “bromine” method provides reliable data on the metallic uranium and uranium dioxide, i.e. oxygen bonded to uranium. At the same time, the reduction melting method provides information on the total oxygen concentration in the sample under study, which allows calculating the amount of oxygen on every atom of the rare-earth metal considering the data on the amount of oxygen bound to uranium.

The suggested algorithm of the chemical operations was verified using model mixtures, in which various combinations of metals (U, Nd, Ce, and metallic Pd) and their oxides (U, Nd, Ce) were used. The mixtures composed of the known amounts of various substances were used as samples. Metallic uranium served as a basic component (85-95 wt.%). The experimental results were in good agreement with the theoretically obtained values on the concentrations of metallic uranium, neodymium, cerium, and their oxides.

Speaker: Prof. Yury Zaikov (Institute of High Temperature Electrochemistry of the Ural Branch of the Russian Academy of Sciences)
• 206
Development of Artificial Intelligence through PLC & SCADA to predict process related failure and abnormality in a Reprocessing Plant

Programmable Logic Controllers offer complete automation solution and flexibility to control in a plant like the nuclear fuel reprocessing plant. To ensure the safety of both the plant and personnel, continuous monitoring and diagnostics of plant parameters are implemented through various means. Audible and visual alarms are provided to alert the operator in case of process abnormality. But all the mechanisms are available only to report to the operator after an abnormal event, so that corrective action can be taken by the maintenance personnel. This increases the plant down time and involves tedious investigation of all related data in the operator’s log and historical trend of the data.

The current work aims to develop an artificial intelligence (AI) based diagnostic tool within the existing HMI application, which can continuously monitor the plant data, review all associated process conditions and then predict the possibility of occurrence of any abnormal event. This AI based tool doesn’t limit to event prediction, but also alerts the operator and suggests suitable corrective action that can be taken to avoid the event. An AI algorithm has been developed within the existing PLC/SCADA, thus requires no additional diagnostics tool and ensures smooth operation of plant.

Methods and materials: The system comprises PLC modules and SCADA servers & clients. The PLC modules continuously communicate with the field instruments to read inputs, execute logics and send control outputs to perform various liquid transfers, damper operations in ventilation/offgas systems and other utilities management. The SCADA server communicates with PLC to monitor and control various process, ventilation and radiation data. An AI based application/utility is developed which continuously monitors data in a plant scale to predict the next failure of a particular process or a system, so that predictive maintenance can be taken up by maintenance personnel. The tool predicts any unintended process events like process parameter deviation (tank level or tank pressure) due to impulse tubing leaks /failure of compressed air supply, unintended transfer of radioactive liquid, volume imbalance between source tank and destination tank, reduction in area/containment box vacuum due to exhaust system failures, temperature sensor failures and so on.

Speaker: Ananthasivan Krishnamurthy (IGCAR, DAE)
• 207
Development of density control technologies for MOX pellet using dry recycled powders

Technology to utilize a dry recycled MOX powder has been developed as a part of MOX fabrication technology development for fast reactors. The purpose of this study is to develop a technology to control the density of MOX pellets with use of dry recycled MOX powder. A roll crusher and a jet mill were employed to prepare the recycled MOX powder which had three kinds of particle sizes (coarse, medium and fine). Sintering tests of MOX pellets were carried out as parameters of particle size and addition rate of dry recycled powder. The results are summarized as follows.

• For the coarse and medium dry recycled powders, a decrease in density due to addition was confirmed, but for the fine dry recycled powders, almost no decrease in density due to addition was confirmed. From this, it is considered that the fine dry recycled powder can be used in the same manner as the raw material powder such as the raw MOX powder as long as the addition rate is up to about 40 wt %.
• When dry recycled powder (coarse or medium) and pore former were added at the same time, a synergistic effect was produced in addition to the density reduction effect of both, and the density was lower than the expected density. In addition, this synergistic effect occurred within the range of this test at 10 wt% of coarse dry recycled powder + 2 wt% of pore former, or 15 wt% of medium dry recycled powder + 2 wt% of pore former. Further, it is considered that this synergistic effect can be alleviated by adding fine dry recycled powder.
• It is considered that the addition of coarse and medium dry recycled powder can delay the progress of sintering by adding it together with the pore former, and the influence can be suppressed by adding fine dry recycled powder.
• High dry recycled powder addition caused cracks in the pellets, but addition of 2% by weight of pore formers no longer observed cracks.
Speaker: Masahiro Nishina
• 208
Electrical conductivity of multicomponent chloride melts, containing ions of mono-, di-, and trivalent metals

Melts based on the LiCl-KCl eutectic are becoming attractive in various industrial fields, including nuclear industries. However, their transport characteristics have not yet been sufficiently studied.

The purpose of this work is to study the electrical conductivity of melts similar to those formed during the dissolution of real nitride spent nuclear fuel in (LiCl-KCl)eut., and also to develop a model that would allow us to evaluate the electrical conductivity of multicomponent melts of arbitrary compositions based on the conductivity of 2-3 component mixtures.

To achieve this goal, we measured the electrical conductivity of the molten (LiCl-KCl)eut. mixtures with various mono-, di- and trivalent metal chlorides (CsCl, CdCl2, SrCl2, CeCl3, NdCl3, UCl3) in a wide temperature range. Also, the electrical conductivity of several multicomponent mixtures (LiCl-KCl)eut. - CsCl + MeCl2 + MCl3 with various combinations and concentrations of components was measured. In the present work, a capillary quartz cell with platinum electrodes and the AC-bridge method at the input frequency of 10-75 kHz were used. The density of the melts and their molar electrical conductivity was calculated.

Electrical conductivity is a non-additive property. For example, the conductivity deviations of the (LiCl-KCl)eut. + NdCl3 molten mixture from additive values reach ~ 80-90%. The stronger the interaction (complexation) between the ions in the melt, the greater the deviations from additivity. Mixtures composed of (LiCl-KCl)eut. and CeCl3, NdCl3, etc., showed such strong interactions. The results were interpreted in terms of the coexistence and mutual influence of the complexes formed by mono-, di-, trivalent cations, and counter anions in these molten mixtures. When UCl3 or LnCl3 are dissolved in the molten LiCl-KCl eutectic the nearest U3+ - Cl- or Ln3+ - Cl- distance is reduced, as well as the coordination number of the trivalent cation. In all cases, coordination number(CN) ≥ 6. This leads to a decrease in the concentration of electricity carriers Li+, K+ and, especially, Cl-, and, thus, to a decrease in the electrical conductivity of the melt, as we observed experimentally.

Speaker: Prof. Alexei Potapov (Institute of High Temperature Electrochemistry)
• 209
Electrolytic reduction of the simulated oxide spent nuclear fuel in LiCl-Li2O melt

A pyrochemical technology for reprocessing spent nuclear fuel (SNF) and fast reactors is being implemented. One of the redistributions of pyrochemical technology is the electrochemical reduction of uranium dioxide (actinide oxides) with lithium in a LiCl - Li2O melt (1-2 wt.%) uranium dioxide and rare earth oxides at 650 °C. To test the technological regimes of the reduction process, we used a model nuclear fuel (MNF). It was a mixture of uranium dioxide and rare earth oxides. Nickel oxide ceramics were used as the anode, and a stainless steel basket, into which MNT pellets were loaded, served as the cathode. The electrolysis process was carried out at a cathode potential more positive than the separation of the liquid phase of metallic lithium. The total amount of electricity consumed for the reduction of MNF in one cycle did not exceed 160% of the theoretical value required for the electrolytic production of lithium for the reduction of uranium dioxide.

The UO2 + 5-10 wt. % tablets (La2O3, CeO2, Nd2O3 in a ratio of 1:1:1) were used as samples for reduction. To determine the degree of reduction of the cathode product to metals, we proposed a combined approach for the determination of the metal phase in the reduced product.

The first, "bromine" method consists of dissolving the reduced product in a solution of bromine in ethyl acetate. The metal fraction of uranium goes into the liquid phase, and the remaining uranium oxides remain in the precipitate. This method is generally accepted for determining the conversion of uranium dioxide to metal. It is possible to accurately determine the amount of uranium metal and its dioxide and, consequently, the oxygen associated with uranium in the test sample.

The second method is the reduction melting of metals and oxides in a graphite crucible using a molten metal bath at a high temperature, carried out by us on a Metavak-AK device. This method allows determining the total oxygen content of the sample. The combination of these two methods and general chemical analysis for the elements of interest to us allows us to determine the amount of oxygen per atom of a rare earth element (lanthanide).

It has been shown experimentally that executing the reduction process, subject to the above conditions, makes it possible to obtain a product with the reduction to metallic uranium by 98-99%, while lanthanum, cerium, and neodymium remain in the form of oxides.

Speaker: Alexander Dedyukhin (Institute of High Temperature Electrochemistry of the Ural Branch of the Russian Academy of Sciences)
• 210
Optimization of Ruthenium concentration in PUREX Process during Fast reactor fuel Reprocessing

In Purex process, Ruthenium is one of the troublesome fission products due to its complex chemistry and presence of multiple oxidation states &some extractable stable complexes in nitric acid medium. The higher concentrations of both stable and radioactive ruthenium isotopes, pose many challenges. During the reprocessing of FBR spent fuel, the tri and tetra nitrato complexes of ruthenium get extracted in the first cycle in to TBP. The daughter product of Ru106 is Rh106 which is a hard gamma emitted with a half-life of only 30 s. This leads to the degradation of the solvent due to the extraction of radioactive ruthenium. The degraded solvent in turn holds high plutonium in to the organic phase rendering it unstrippable. In addition, the radioactive Ruthenium pick up will contribute significantly to the residual activity of the product stream. Also, the extractable ruthenium species pose problems in the treatment of the lean organic. Hence it is necessary to limit the co-extraction of Ruthenium in to the solvent. In order to accomplish this, a scheme to scrub the plant stream with concentrated acid has been tested in the plant.

In the CORAL plant during the reprocessing of Pu rich mixed carbide spent fuel discharged from FBTR, the first cycle extraction was carried out at an acidity of 5.5 M. This resulted in the reduction of ruthenium activity in the loaded organic by more than 50%. This reduction was due to the fact that the distribution ratio of ruthenium bears an inverse relation to the concentration of nitric acid in the aqueous phase in equilibrium with the organic in contact with the acid. Analysis of the plant data shows about 30-40% retention of ruthenium activity in the loaded organic. This could be due to the presence of tri nitrato nitrosyl complex of ruthenium RuNO(NO3)3 under the operating conditions which is extracted by TBP. In the loaded organic phase, the extracted ruthenium is initially expected to be existing as an outer sphere complex, which is slowly gets converted to an inner sphere complex which renders the complex un-scrubbable even with a high acid stream.

Speaker: Ananthasivan Krishnamurthy (IGCAR, DAE)
• 211
R&D on recovery and separation of americium and curium under "Proryv" project

New nuclear fuel cycles include reducing the long-term radiotoxicity of nuclear waste by separation and transmutation of long-lived transplutonium elements. Therefore, selective recovery of transuranic elements, especially actinides (III) – americium and curium – from high-level waste generated during spent nuclear fuel reprocessing is an important issue. Processes for extracting americium (III) from PUREX-process raffinates are under development in Europe, USA, Russia, Japan and other countries.
Russia is currently actively working on the separation of americium and curium under the "Proryv" project. A two-stage flowsheet is envisaged, including extraction group recovery of americium and curium, followed by sorption separation of americium and curium. Dynamic tests of different extraction systems (CMPO, TOGDA, Dyp7 in polar fluorinated diluent (F3), UNEX-T) were carried out after intensive laboratory studies. Based on the results of the tests, TODGA - F3 system was chosen to test on the real high-level radioactive waste. "Hot" dynamic test of actinides (III) recovery from PUREX-process raffinates using extraction system TODGA - F-3 was carried out at Production Association "Mayak". No less than 99.9% of americium was extracted from during processing of BN 600 and VVER-440 spent nuclear fuel. The total working time of the test was 70 hours.
Extractants based on asymmetric diglycolamides (DGA) are currently being studied for transplutonium elements extraction, instead of TODGA, will make it possible to use saturated hydrocarbons as diluents and to abandon fluorine containing F-3.
The chromatography separation of Cm and Am from rare earth and transplutonium elements concentrate was tested at the Production Association "Mayak". The used concentrate was obtained during processing of VVER-440 SNF. Around 14 g Cm was allocated. The Cm-Am fraction contained about 4.6 g Cm and about 40 g Am. 65 g of pure Am fraction were obtained.

Speaker: Dr Vitali Vidanov (JSC "Proryv")
• 212
Removal of Radiocesium from High-Level Liquid Waste using Inorganic Ion-exchangers

The present study demonstrates the use of inorganic ion-exchanger (IX) to condition the high-level liquid waste (HLW) by selective separation of one of the major radionuclide, cesium-137 (137Cs) from it. 137Cs possesses a broad range of potential applications in societal and agricultural area. In addition to this, the selective separation of 137Cs from HLW would drastically bring down secondary waste generation and reduce burden on the off-gas treatment in the vitrification process. Here, we have successfully demonstrated the conversion of Cs loaded IX to compact waste form.
Among the various adsorbents, Ammonium Molybdo-Phosphate (AMP) was preferred and used as IX in the present study because it shows high selectivity towards Cs+ in the presence of various metal ions (alkali, transition, lanthanides and actinides) and is stable under highly acidic & irradiation condition. Despite these advantages, the powder form of IX is not readily adaptable and does not provide ideal flow dynamics for continuous column operations. With a view to bring it to a usable form, synthesis of composite forms of IX (20-30%) in Poly-Ether-Sulfone (PES) was carried out. By adjusting the ﬂow rate of the polymer liquid, particles with an average size ranging from 150 to 710 μm were obtained using a dual nozzle device that allows the break–up of polymer solution by air blowing. The polymer particles of 355-600 μm in diameter were mainly used for Cs extraction studies.
The obtained beads were characterized for thermal stability using thermogravimetry (TG), phase purity by X-ray diffraction (XRD) and functional group identification by Fourier transform (FT)-infra-red (IR) spectroscopy. The thermograms of IX and IX-PES beads showed few steps of decomposition reactions, it may be due to the loss of moisture, ammonia, PES and MoO3 from AMP-PES. To assess the efficiency of the IX-PES beads, its cesium extraction capacity and distribution coefficient were determined using actual HLW. The extraction capacity and distribution coefficient of cesium in actual HLW (3M acidic) was 4369.3 cm3/g and 0.4 meq/g, respectively. Column studies (L/D=3) were also carried out with an HLW flow rate of 0.5mL/min. The performance of the column was evaluated by plotting a breakthrough curve. Pellet formation of the inactive cesium loaded IX beads was successfully demonstrated using a manual pelletiser with a pressure of 150 kg/cm2.

Speaker: Ms Anshul Kumari (Waste Immobilization Plant, Integrated Nuclear Recycle Plant, Nuclear Recycle Board, Bhabha Atomic Research Centre, Kalpakkam 603 102, India)
• 213
Reprocessing of nitride and metallic spent nuclear fuel using molten salts

In recent years, several countries, including Russia, have been developing a pyrochemical (anhydrous) method for spent nuclear fuel (SNF) reprocessing. Molten salts have several advantages, such as thermal and radiation stability, a wide electrochemical window, etc. They can be used practically at all technological stages of SNF processing.

The first stage of pyrochemical reprocessing of nitride spent nuclear fuel can imply its dissolution in the molten LiCl-KCl eutectic containing a chlorinating agent. It is proposed to use CdCl2 or PbCl2 as a chlorinating agent. We have studied in detail the interaction between UN and molten LiCl-KCl eutectic, containing cadmium chloride, depending on the temperature and CdCl2 concentration. It was found that at temperatures below 750 °C, the interaction proceeds through several parallel reactions and, along with UCl3, a precipitate, consisting of UNCl, nonstoichiometric nitrides UN1.59, UN1.69, U4N7, and several other compounds, is forms. At 750 °C and above, all intermediate uranium nitrides dissolve in excess CdCl2 to form UCl3. The conditions, under which the100% conversion of UN → UCl3 is possible, are provided. The use of lead chloride as a chlorinating agent has also been studied. The chlorination proceeds according to the same mechanism, but the use of PbCl2 allows the process temperature to be reduced by 100 degrees maintaining the 100% UCl3 yield.

The interaction between metallic uranium and its alloys and noble metals with the LiCl-KCl eutectic melt, containing PbCl2, was studied. The dissolution of uranium in such melt is very intense. For samples weighing ~ 15 g, the reaction is completed in 15-20 minutes. The interaction between U-Pd and U-Ru alloys proceeds much more slowly and according to a more complex mechanism. A high temperature and a large excess of PbCl2 are required to complete the reactions. Thermodynamic modeling of the interaction reactions was carried out. The kinetics was studied and the reaction products were identified.

It is shown that pyrochemistry methods may be successfully used for reprocessing of both nitride and metallic spent nuclear fuel. Virtually all processing operations can be performed using molten salts as a process medium.

Speaker: Prof. Alexei Potapov (Institute of High Temperature Electrochemistry)
• 214
Transmutation efficiency of minor actinides in fast-and thermal-spectrum molten salt reactors

Long-lived minor actinides (MA) like, Neptunium, Americium, and Curium are the major burden of nuclear power. Long-lived MAs have not yet been used as nuclear fuel. Therefore, the transmutation of long-lived MAs is introduced as an alternative to direct final disposal. In current work, we compare the performance of MA transmutation in a critical Single-fluid Double-zone Thorium-based Molten Salt Reactor (SD-TMSR) and a Small Molten Salt Fast Reactor (SMSFR). We study the dynamic of Keff and core reactivity with different MA loads, shift of the neutron spectrum, time evolution of MA and basic nuclides inventory that affect the core stability, as well as the transmutation coefficient (TC). The TC of long-lived MA is calculated using the Monte Carlo code SERPENT-2. The total neutron flux in SD-TMSR and SMSFR can reach 4.1x1014 and 1.8x1015 n/cm2s, respectively. The results show that SD-TMSR consumes about 50% of the generated Pu isotopes in the fuel salt; however, SMSFR consumes about 86.5% of the generated Pu isotopes. During burnup, we apply online reprocessing and refueling, so the core remains critical, and the total mass of fuel in the core and blanket is practically constant. The results show that both reactors efficiently transmute 237Np, 241Am, 243Am and 243Cm, while SMSFR has a higher TC than SD-TMSR. TC of total MA reaches 54.84% and 87.97% in SD-TMSR and SMSFR, respectively.

Speaker: Mr O. Ashraf (Institute of Nuclear Physics and Engineering, National Research Nuclear University MEPhI, 115409, Russian Federation, Kashirskoye Shosse, 31)
• 215
Transmutation of minor actinides in a fast reactor with uranium-curium fuel

As a result of the operation of nuclear reactors, a certain amount of Cm is produced, which is included in the minor actinides series (MA). Among the long-lived Cm isotopes, Cm243 and Cm245 should be noted. Their fission cross section is over 2.5 barn. In this regard, Cm can be used as a fuel in a fast neutron nuclear reactor.
For the scientific research, was used a model of the RBEC reactor (a fast natural circulation reactor with a lead-bismuth coolant), developed at the Kurchatov Institute (Moscow, Russia).
(U + Cm)N was used as fuel. Uranium - waste uranium with an enrichment of 0.1% in the isotope U235. The efficiency of different approaches to the placement of MA in fuel (homogeneous and heterogeneous) was considered. This was for the transmutation of Cm and other elements from the minor actinides series.

Speaker: Ms Anna Terekhova
• 3.2 Development of innovative fuels: design and properties irradiation

Chairs: Nathalie Chauvin and Min Xiao

• 216
Recent studies on fuel properties and irradiation behaviors of Am/Np-bearing MOX

There remain challenges in studies of properties and irradiation behaviors of mixed oxide (MOX) fuels, which aims at reduction in volume and toxicity of high-level radioactive wastes, because of the influential factors such that the fuel reaches very high temperature exceeding 2000 K and oxygen content in the fuel continuously varies depending on surrounding conditions. High temperature and steep temperature gradient of MOX in fast reactor bring about the unique phenomena of pore migration resulting in restructuring and redistribution of elements. In this study, we report our experimental results of the property studies on Am/Np-bearing MOX and discuss how these properties influences on the irradiation behaviors.
Oxygen potentials of Am/Np-bearing MOX have been collected by gas equilibrium technique and reported by the group of the authors. Both Am and Np inclusions in terms of substituting U increase the oxygen potential of MOX with the Am and the Np changing from quadrivalent to trivalent. The effects of Am/Np inclusion were analyzed via defect chemistry and quantitatively incorporated with the exiting models which relates oxygen-to-metal (O/M) ratio, contents of Pu, Am, and Np, temperature and oxygen partial pressure.
Inter-diffusion coefficients of U-Pu, U-Am and U-Np in MOX have been obtained by using diffusion couple technique. Although the measurement results could contain uncertainty, some important trends were obtained, i.e. the inter-diffusion coefficient of U-Am is the largest and that of U-Pu is the second. O/M is significantly influential such that the inter-diffusion coefficients were larger at the O/M=2 than those of O/M<2 by several orders.
The pore migration along temperature gradient during irradiation is considered to arise due to the vaporization and condensation of actinide species in pores. Especially, large vapor pressure of UO3 is the dominant property for the pore migration. The increase of the oxygen potential of MOX with Am/Np leads to more UO3 and the acceleration of the pore migration.
The redistributions of actinide elements were also considered with the relationship of the pore migration, i.e. diffusion in solid phase to relax the inhomogeneity caused by the vaporization and condensation of UO3. Thus, the inter-diffusion coefficients can directly influence on the magnitude of the redistribution.
The obtained properties were modelled with the parameters such as temperature and oxygen partial pressure. This enable the integrated time developing evaluation including the temperature profile of fuel irradiation by simulation code.

Speaker: Shun Hirooka (Japan Atomic Energy Agency)
• 217
Selection, testing and development of qualification procedure for ALLEGRO gas-cooled fast reactor fuel

On the basis of detailed review, the fuel types were proposed for the new design of the ALLEGRO gas-cooled fast reactor. The first core will be built with MOX or UOX fuel in 15-15Ti stainless steel cladding. These fuel types have been widely used in different sodium-cooled fast reactors. The second core of ALLEGRO will use refractory fuel. The primary candidate is carbide fuel – (UPu)C or UC – in SiC cladding.
15-15Ti and SiCf/SiC type claddings were tested in high temperature helium atmosphere with different impurities in order to investigate the effect of high temperature treatment and impurities on the mechanical load bearing capabilities of these cladding materials. Ballooning tests were performed with 15-15Ti cladding tubes and it was shown that they can keep their integrity at high temperature. The failure pressure of samples tested at 960-1000 °C was above 18 MPa.
Qualification procedures have been proposed for the start-up and refractory ALLEGRO fuel. The technology readiness level approach was applied and the basic step of qualification procedure were identified. Using the currently available information the further needs were specified, which include experimental activities, design work, development of numerical models, technology developments, establishment of fuel fabrication capabilities, irradiation in research reactors and post-irradiation examination of fuel.

Speaker: Mr Zoltán Hózer (Centre for Energy Research)
• 218
Design of metal fuel pin for test irradiation in FBTR and for future reactors.

In India, a structured R&D program on the development of metallic fuel and associated fuel cycle for Fast Breeder Reactors (FBRs) is undertaken so as to realize commercial metal fuel FBRs in the future. Towards this, initially test irradiation of sodium bonded metal fuel pins in Fast Breeder Test Reactor (FBTR) core was proposed and hence the pin design for various compositions of metal fuel was carried out and they are currently being irradiated in FBTR. The compositions include Natural U- 6%Zr, Enriched U -6%Zr, Natural U-19%Pu-6%Zr and Enriched U-23%Pu-6%Zr. Three pins of each type are being irradiated in FBTR and their current burn-up levels are 2.26, 15, 19.5 and 3.75 GWd/t respectively. For a typical test pin irradiation (Enriched U-23%Pu-6%Zr), three sodium bonded metal fuel pins of length 531.5 mm are arranged inside a capsule which is kept inside an ISZ 100 special SA. The thermal and mechanical design of the pin was carried out and the safe operation of fuel pin is ensured for a peak Linear Heat Rating (LHR) of 318 W/cm and for a target burn-up of 100 GWd/t. Also during transients, the maximum allowable flow reduction in the ISZ100 SA was found out to arrive at the blockage limits. During manufacturing of sodium bonded pin, bubbles get trapped inside the bond sodium and hence analysis was carried out to determine the allowable bubble size in a pin.
For the design basis transients, the design safety limits for the metal fuel pin (fuel and clad) have been arrived at by analysis. Also, a 2-D transient mathematical model has been developed for predicting fuel melting and movement of melt interface with respect to time. It was observed that fuel melting starts when the reactor power reaches 1.45 times the nominal power.
Based on the above inputs, for a power reactor fuel composition (Natural U-19%Pu-6%Zr), the thermal and mechanical design of sodium bonded metal fuel pin was also carried out. Thus, this paper details about the design aspects of sodium bonded metal fuel pin which includes arriving the size of fuel pin, fission gas plenum length, allowable linear power, allowable bubble size in the bond sodium and the safety limits for transient events.

Speaker: RAJKUMAR THIRUNAVUKKARASU (IGCAR)
• 219
Development of simplified fuel fabrication technologies for fast reactors

A high-density annular MOX fuel pellet fabrication technology has been developed for producing a low O/M ratio of less than 1.97 for fast reactors. The low O/M ratio sintered pellets aim to suppress the fuel-cladding chemical interaction (FCCI) at high burnup, and a simplified MOX pellet fabrication process (Short process) is a new production technology for this MOX fuel. The short process is a technology for producing pellets by tumbling granulation, die wall lubrication pressing, sintering, and O/M ratio adjustment using a raw MOX powder obtained by the microwave heating direct denitration method. Compared with the conventional process, the short process can reduce the number of processes from 23 to 8, which makes it possible to improve economic efficiency. In this report, the development situation of the short process was reviewed, and the test results of die wall lubrication pressing and O/M ratio adjustment technologies were extended for scale-up of the fabrication technology.
In the development of the die wall lubrication pressing technology, it is necessary to find the optimum operating conditions because a tumbling-granulated MOX powder is directly pressed without a mixing process with additional lubricant to fabricate annular pellets. The MOX granulated powder was fed by a feeder to a die, and was pressed with 8 cycles/min punch at about 510 MPa. The green pellets of about 55.3 %T.D. were sintered at 2023 K for 4 hours to obtain sintered pellets of 95 %T.D. or higher. The quality of the green annular pellets can be improved by optimizing the operating conditions of the die wall lubrication pressing.
Regarding the O/M ratio adjustment, as results of scaled-up tests by increasing the loading amount from 1.0 to 2.0 kgMOX/batch, and 5%H2+95%Ar mixed gas flow rate from 5.0 to 10.0 l/min/kgMOX, the average O/M ratio increased from less than 1.97 to slightly higher than 1.97. As a result of the thermo-fluid dynamics simulation, it was revealed that a large part of gas did not pass through the mesh plate and leaked through the clearance between the mesh plate and the gas inlet. Further simulations indicate that the gas flow path can be improved by lengthening the lower end of the outer frame of the tiered mesh plates and installing a rod with an ejection hole in the center of each mesh plate. It is expected that these methods can reduce the O/M ratio to less than 1.97.

Speaker: Dr Tomoomi Segawa (Japan Atomic Energy Agency)
• 220
Advanced Reactor Experiments for Sodium Fast Reactor Fuels (ARES) Project: Transient Irradiation Experiments for Metallic and MOX Fuels

Advanced Reactor Experiments for Sodium Fast Reactor Fuels (ARES) is a joint project between the U.S. Idaho National Laboratory (INL) and the Japanese Atomic Energy Agency (JAEA) to investigate the transient fuel performance of metallic and MOX fuels. The project has the specific goals of experimentally evaluating the transient failure modes of high burnup metallic and MOX fuels, guided by advanced modeling and simulation (M&S) tools, but also to support development and validation of M&S tools. The recent availability of the Transient Reactor Test (TREAT) facility provides the opportunity to renew in-pile evaluation of advanced fuel designs. Sodium Fast Reactor (SFR) experiments leverage a rich inventory of fuels irradiated in EBR-II and FFTF, currently residing at INL, and still supporting research programs at the INL. As part of the ARES project, a heat-sink capsule with liquid metal specimen bond has been designed and M&S is being used to develop the detailed experimental conditions for the planned experiments. A series of fresh fuel commissioning tests is planned in TREAT for 2021. These tests will evaluate the performance of the hardware and instrumentation to measure temperature response, fuel elongation, and fuel failure. The four irradiated fuel tests are planned for fully intact, high-burnup pins (2x metallic pins and 2x MOX pins), which achieved nearly ~13 at% burnup in EBR-II, to be irradiated in TREAT in 2022. Test instrumentation will include optical-fiber-based distributed temperature sensors, thermocouples, acoustic emission detectors, a capsule pressure sensor, self-powered neutron detectors (SPND) as well as the refurbished TREAT fuel motion monitoring system (hodoscope). Fresh fuel tests will incorporate additional advanced diagnostics including optical-fiber-coupled pyrometry and fuel elongation sensors. The experiments will benefit from advanced pre- and post-transient examination capabilities available at INL and actively used for similar examinations. Long-term test development underway includes a full circulating sodium loop and the ability to refabricate fuel specimens. This paper will present more detailed description of the planned test design, matrix, and conditions with an emphasis on modeling guiding experiment design.

Speaker: Colby Jensen (Idaho National Laboratory)
• 221
Towards design guidelines for fast reactor oxide fuel pins with high Pu content: driving post irradiation examination by benchmarking European fuel performance codes

In the framework of the European Commission call for proposal “Horizon 2020”, the project Plutonium Management for More Agility, called PuMMA, is granted. This project starts in October 2020 and will last four years. A work package is dedicated to the behaviour and safety of mixed oxide fuels with high plutonium content, which is essential for plutonium multi-recycling or plutonium burning in fast reactors. This paper describes main goals and status of this work package. Major task is the comparison of a large set of european fuel performance codes (FPC) on the basis of three passed experimental irradiations of oxide fuel pins containing around 45 % of plutonium: CAPRIX, irradiated Phenix French Reactor, TRABANT 1 and TRABANT2, irradiated in High Flux Reactor, HFR, in the Netherland.
The first phase of the work consists in the definition of irradiation conditions for fuel pins simulation, involving CEA and NRG. In a second phase, 10 various FPC will be used by 13 European nuclear research organizations in order to simulate these three irradiations: SIMMER-V, TRANSURANUS, OFFBEAT, FEMAXI, FUROM, FRED, MACROS, FINIX, TRAFIC and GERMINAL. Results will be compared in terms of global and local quantities: fission gas release, fuel pin elongation, profilometry, central hole radius, Pu redistribution, internal corrosion, etc. Moreover, this first set of simulations will be used to define the post-irradiation examinations programme, which will be carried out in the framework of the PuMMA project in JRC-ITU and CEA facilities. In a third phase, simulation tools will integrate new thermal properties measurements to be realized in PuMMA (other workpackage), the back-up of first comparisons, and these irradiations simulations will be re-launched and compared to experimental measurements. This mixed approach simulation/examination will allow to improve fuel codes reliability and to reduce uncertainties in the design process of this kind of fuel, which is outside of the validation area of all the existing codes. The experimental programme will be devoted to FPC validation as well as knowledge improvement. Last part of the work package will also tend to propose specific safety recommendations for the design of this kind of fuel.

Speaker: Victor BLANC (French Atomic Energy Commission)
• 222
Root Cause Analysis of FBTR Failed Fuel Pin

The Fast Breeder Test Reactor (FBTR) at Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam, is a loop type, sodium cooled fast reactor. Its main aim is to provide experience in fast reactor operation, large scale sodium handling and to serve as a test bed for irradiation of fast reactor fuels & materials. India has been operating FBTR with Mixed Carbide Fuel as the driver fuel since 1985. Mixed Carbide was chosen as the fuel due to its high stability with Pu rich fuel, compatibility with coolant and for its better thermal performance. Being a unique fuel of its kind without any irradiation data, it was decided to use the reactor itself as the test bed for this driver fuel. The fuel has performed extremely well, with the peak burn-up reaching 165 GWd/t. In the year 2011, MK-1 fuel SA that reached 148.3 GWd/t burnup in III ring of FBTR core had a single pin failure which was identified by both cover gas detectors as well as bulk DND detectors. Subsequently, Post Irradiation Examination (PIE) was carried out on the Failed Fuel Subassembly. Various possible causes of fuel pin failure in the SA were postulated.
One of the initial causes of failure was identified as flow reduction through the SA which was studied and ruled out by a detailed analysis. Also, deformation caused in the fuel pin geometry due to high irradiation dose, results in only 4 % reduction in flow through the SA. Subsequently, a detailed analysis of the failed fuel pin has been carried out for the estimation of Fission gas pressure & FCMI induced stress, clad strains, Clad Cumulative Damage Fraction etc. at different axial levels as a function of burnup. Studies were also carried out to find out the reasons for the ovality of the pins after irradiation. Above parameters are analyzed for the Failed fuel SA and the results are compared with first ring MK-1 fuel SA so as to assess whether failed fuel SA has experienced any abnormalities compared to first ring SA. Also, an attempt has been made to bridge the gap areas between the PIE observations and the analysis results of the failed fuel SA to ascertain the reasons for the failure.

Speaker: NAGA SIVAYYA DUDALA (IGCAR, Kalpakkam)
• 223
Uranium and mixed uranium-plutonium nitrides thermal stability

The thermogravimetric method was used to study the behavior of uranium nitride and mixed uranium-plutonium nitride (MNIT) in a helium flow and a helium with nitrogen gas mixture at temperatures up to 2173 K. When heated in helium in the low-temperature range (˂1773 K), a mass loss was found, which amounts to hundredths of a percent. In this case, mass loss occurs in 2 stages, accompanied by the release of nitrogen and it is not associated with the decomposition of uranium or plutonium mononitrides. It has been shown that sintered nitride fuel pellets may contain several percent of uranium sesquinitride U2N3, which decomposes in this range. Nitride fuel pellets were heated in a gas mixture of helium with nitrogen to study the formation of higher nitrides. In the case of uranium mononitride this led to the formation of uranium sesquinitride U2N3 in the temperature range of 673-723 K. However, upon further heating (>1173 K), U2N3 decomposes again to uranium mononitride in 2 stages. The sequential formation and decomposition of uranium sesquinitride led to the destruction of the sample. At the same time multiple heating of the MNIT fuel (U0.79Pu0.21)N in the helium-nitrogen gas mixture does not lead to the formation of U2N3. It is also shown that the partial pressure of nitrogen at its content of 5 vol.% in the helium flow significantly exceeds the equilibrium partial pressure of nitrogen over the samples of uranium nitride and MNIT fuel in the entire test temperature range, which inhibits the decomposition of uranium mononitride up to 2173 K. However, in the case of MNIT fuel at a temperature >1773 K a clearly observed mass loss on the thermogravimetric curve occurs. Therefore, even in an atmosphere containing nitrogen, it was not possible to suppress the decomposition process of the MNIT fuel.

Speaker: Mikhail Krivov (Joint Stock Company "A.A. Bochvar High-technology Research Institute of Inorganic Materials")
• 5.2 Experimental Programs I

Chairs: Daniela Gugiu and Yican Wu

• 224
Estimation of mean charge on sodium metal aerosol in the argon and nitrogen gas environment during external gamma irradiation

The cover gas region of sodium cooled fast reactors is always being subjected to intense ionization radiation field apart from radioactive aerosols and gases. The radiation produces significant ionization of the medium resulting in large amount of bi-polar ions. The acquisition of electrical charge by sodium aerosols in cover region under bipolar ionic atmosphere draws special attention as it modifies the dynamics of aerosol transport, deposition and process inside the cover gas space. Towards this, a study has been conducted to characterize the sodium aerosols present in the cover gas region using sodium loop facility (SILVERINA loop) with and without the presence of gamma radiation field using argon as a cover gas and the experiment is repeated with nitrogen gas. The experiments demonstrated that the size of sodium aerosol is found to be relatively higher and mass concentration is lower in the presence of gamma field as compared to the condition without gamma. In order to address the behavior sodium aerosols in the presence of radiation, it is customary to understand the charge acquired by the aerosols under the radiation field. The charge acquired by aerosol is defined by the modified Boltzmann theory and determined as a function of ion mobility of argon and nitrogen gases, aerosol diameter and temperature of cover gas region. The average elementary charge is found to be negative and charge number increases with increase of aerosol size which in turn depends on the sodium pool surface temperature for both the gases. The average charge on sodium aerosol is more in argon gas as compared to the nitrogen and the difference increases with sodium pool temperature. The increase in sodium aerosol charge in argon gas is due to the higher mobility ratio of positive to negative ions for nitrogen relative to the argon. The theoretically determined charge distribution is more asymmetric in argon gas compared to the nitrogen gas. Since the charged aerosols promote more coagulation and enhance surface deposition which are indirectly indicated by the changes in the measured aerosols characteristics. Finally the mean charge can be used for calculation of coagulation and deposition rates which is important for the realistic determination of aerosol characteristics in the cover gas region. The sodium aerosol characteristics and behavior in cover gas gives significant insight into the heat and mass transfer across the cover gas space, cover gas purification, roof slab and handling machines for the fuel sub-assembly.

Speaker: Dr Amit Kumar (Radiological and Environmental Safety Division, IGCAR, Kalpakkam-603102, India.)
• 225
THERMOHYDRAULIC TESTS IN JUSTIFICATION OF DESIGN CHARACTERISTICS OF THE BREST-OD-300 RP STEAM GENERATOR

In order to substantiate the design characteristics of the steam generator of the BREST-OD-300 reactor plant (RP) developed at NIKIET JSC, IPPE JSC carried out thermohydraulic tests of various models of the lead-heated steam generator. Initially, to confirm the design characteristics and thermal-hydraulic stability at the parameters of the nominal, partial and start-up modes, a model of a twisted steam generator was tested, consisting of two three-tube modules with a longitudinal lead flow around a bundle of heat transfer tubes. The influence of operating parameters on thermohydraulic characteristics and hydrodynamic stability is shown in the case of operation of one module, as well as in the joint operation of two models in the investigated range of operating parameters.
At the second stage, tests of the standard model of the BREST-OD-300 RP steam generator were carried out with lead flowing around 18 heat exchange tubes. The model consisted of two collectors, each of which included a bundle of nine heat transfer tubes. Data were obtained on the hydrodynamic stability of steam generating tubes and the entire model as a whole when operating in the entire range of change in operating parameters, necessary for creating a databank and further verification of calculation codes describing the ongoing thermohydraulic processes. The boundary of thermohydraulic stability has been experimentally confirmed.
The tests of the standard model, as well as the model of a steam generator with a longitudinal flow of a lead coolant, were carried out in a wide range of changes in operating parameters. The feed water pressure varied from 16.5 to 18.7 MPa (on the model with longitudinal media flow, tests were carried out at a supercritical pressure of 24.3 to 25.7 MPa), the water flow through the heat transfer tube varied from 10 to 120% of the nominal value. The feed water temperature varied from 340 to 350 °C, the lead temperature at the model inlet varied from 390 to 536 °C, and the lead consumption varied from 10 to 100% of the nominal value.
A series of works devoted to the study of heat transfer from the lead coolant with a transverse flow around a package of heat transfer tubes has been completed. A model with a transverse lead flow around the steam-generating tubes has been created, on which studies were carried out on the effect of the oxygen concentration in lead on heat transfer in normal heat transfer modes.

Speaker: Mrs Iuliia Kuzina (IPPE JSC)
• 226
Experimental and computational studies of heat exchange for liquid metals boiling in fuel assembly models at accidental conditions

The higher level of modeling for dynamic liquid metal boiling is important for comprehensive analysis of neutron-physical and thermohydraulic characteristics of fast reactors cores for safety justification at accidental conditions (UTOP, ULOF).
Experimental data obtained at IPPE JSC have showed that boiling of liquid metals in fuel assemblies of fast reactors has a complex structure which is characterized by both stable and pulsation regimes with significant fluctuations of parameters that can cause a boiling crisis.
Stable nucleate boiling in fuel assembly models is observed only at limited heat fluxes; and transition to unstable bubble-slug boiling regimes is determined by various factors. In an assembly with a low surface roughness of pin simulators, progressing of unstable (slug) boiling with sharp fluctuations of the coolant flow rate and overheating of the pin simulator wall causes the boiling crisis; in fact, there is no any margin before the crisis. For the pin simulators with higher roughness a transition from unstable slug boiling regimes to stable annular-dispersed boiling regimes is observed due to a liquid film on the surface of the pin simulators.
The hydrodynamic interaction of fuel assemblies can result in to considerable increase in the amplitude of fluctuations in the flow rate of the coolant ("resonance" of the flow rate pulsations) and to the "locking" or inversion of the flow rate of the coolant in the loops, increase in the temperature of the coolant and the pin claddings (the effect of interchannel instability) and, finally, to the boiling crisis.
During sodium boiling in a model fuel assembly with a "sodium cavity" located at the top of the reactor core (the cavity is designed to compensate the positive sodium void reactivity effect when boiling occurs) the possibility of prolonged sodium cooling of the pins in the fuel assemblies for these conditions is shown.
The generalization of data on heat transfer at liquid metals boiling in fuel assemblies is carried out; a cartogram of the flow regimes for two-phase flow of liquid metals in fuel assemblies is developed. The model of the two-phase flow of liquid metal used in the calculations of accidental conditions has been improved; it has a significant impact on the calculation results. The results of comparing the data of calculated and experimental studies are presented.

Speaker: Mrs Julia Kuzina (SSC RF - IPPE)
• 227
Investigation on natural circulation for decay heat removal in reactor vessel of sodium-cooled fast reactor

In Sodium-cooled Fast Reactors (SFRs), it is important to optimize the design and operate decay heat removal systems (DHRS) for safety enhancement against severe accidents which could lead to core melting. To clarify the natural circulation phenomena in a reactor vessel during operation of a decay heat removal system, water experiments have been conducted using a 1:10 scale experimental facility (PHEASANT) simulating the reactor vessel of loop-type SFRs. The dipped-type direct reactor heat exchanger (DHX), the penetrated-type DHX and reactor vessel auxiliary cooling system (RVACS) are mounted in PHEASANT. Moreover, the electric heaters are installed to simulate the core and fuel debris accumulated on the core catcher and upper plenum. Therefore, PHEASANT can simulate the natural circulation phenomena under the various conditions for decay heat sources and DHRS operation. In this paper, the natural circulation phenomena under the conditions of operating the dipped-type DHX and RVACS, respectively, were investigated by the results of PHEASANT experiments and experimental analyses. In the condition of operating the dipped-type DHX, the velocity field was quantitatively obtained by the particle image velocimetry and the characteristic of natural circulation phenomena were clarified by the velocity and temperature data. In the condition of RVACS operation, the temperature was measured using decay heat conditions as a parameter and the effect of decay heat condition on the natural circulation phenomena was investigated. In addition, from the comparison between the experimental results and simulation results, it was confirmed that the numerical simulation is applicable to the natural circulation flow field in the reactor vessel of loop-type SFRs.

Speaker: Kosuke Aizawa (Japan Atomic Energy Agency)
• 228
Coolant flow monitoring with an Eddy Current Flow Meter at a mock-up of a liquid metal cooled fast reactor

As a possible part of the safety instrumentation in liquid metal cooled fast reactors, the Eddy Current Flow Meter (ECFM) is an important and robust tool for continuously monitoring the coolant flow and detecting coolant blockages in the reactor subassemblies for coolant temperatures of up to 700 °C. This inductive sensor can be placed directly above the subassemblies, where changes of the coolant flow angle between 0° and 40° are expected in case of a coolant blockage at some subassembly. In this paper, we present the results of numerical simulations as well as experimental results for the influence of the flow angle with respect to the axis of the ECFM on the output signal of the sensor. In a second experimental setup, an array of ECFM is used to detect and localise coolant blockages by artificially blocking one or more flow channels. For both experiments, a simplified liquid metal flow model was constructed, using the eutectic alloy of Gallium, Indium and Tin to represent flow structures which may typically occur above the subassemblies. Due to the low melting point of this alloy, we were able to perform these experiments at room temperature and to validate the ECFM measurements by ultrasonic velocity measurements available at such lower temperatures. The results of our investigations give important insights into the performance of the ECFM under realistic conditions inside the reactor.

Speaker: Dr Nico Krauter (Helmholtz-Zentrum Dresden-Rossendorf)
• 229
Overview of critical experiments with fast metal cores held on assembly machine FKBN-2

A brief description given for critical experiments held at RFNC-VNIITF on assembly machine FKBN-2 and ROMB critical assembly specially built for verification of neutron transfer simulation codes. The design of ROMB allows to use it for critical experiments with nuclear materials having fast neutron spectrum (highly enriched uranium, plutonium and its’ mix). ROMB assembly contains of wide set of construction materials (such as depleted uranium, beryllium, beryllium oxide, steel, titanium, lead, tungsten, vanadium, molybdenum etc.) shape fitted to parts made of nuclear materials. This permits to hold critical experiments with heterogeneous structures which differ in content, including combinations with neutron moderators. Similar critical assemblies may appear in emergency cases concerned with nuclear fuel cell damage. Benchmark critical experiment set up and its’ features are discussed. Brief description of the experiments held earlier and planned now for verification of neutron transfer simulation codes is given as an example of possibilities of ROMB assembly and experiment using it.

Speaker: Yuri Sokolov
• 230
AN EXPERIMENTAL STUDY ON SECONDARY SODIUM SYSTEM BASED DECAY HEAT REMOVAL CIRCUIT OF A SODIUM COOLED FAST REACTOR

Decay heat removal is an important safety function of a nuclear power plant and failure of the same needs to be practically eliminated. Various concepts for decay heat removal have been adopted in different designs of sodium cooled fast reactors (SFRs) depending upon the size and type of reactor. Some of the pool type designs adopt safety grade decay heat removal system (SGDHRS) which consists of three coupled natural convection loops. SGDHRS removes decay heat through immersed decay heat exchangers in the hot sodium pool and sodium to air heat exchanger kept at a higher elevation at the bottom of a tall stack. These systems are capable of functioning under possible extreme conditions affecting the plant. In order to improve reliability of the decay heat removal function, an additional decay heat removal system capable of functioning under emergency conditions which is connected to secondary sodium system may be considered. One typical design configuration for this system adopts a forced cooling type of sodium to air heat exchanger, operating in parallel to steam generators in the secondary sodium circuit of the reactor. The secondary sodium system based decay heat removal circuit could also be considered to function as a normal shutdown cooling system if its controllability under various operational conditions is established. Accordingly, the controllability of the system under varying decay power scenario and utilization of this system for long term maintenance of cold shutdown condition in the plant are some of the important aspects to be established. Towards this, experimental studies have been carried out using 2 MWt capacity sodium to air heat exchanger in the Steam Generator Test Facility (SGTF) at Indira Gandhi Centre for Atomic Research, Kalpakkam, India under varying heat source conditions (simulating the decay power evolution in the reactor core). These studies ensure controllability and operability of the system at desired operating conditions respecting various thermal hydraulic design constraints under various simulated transient conditions.

Keywords: sodium cooled fast reactors, safety grade decay heat removal system, secondary sodium system based decay heat removal system, steam generator test facility.

Speaker: Mr S.P Pathak (IGCAR)
• 6.3 Multiscale and Multiphysics Calculations

Chairs: Armando Gomez Torres and Chirayu Batra

• 231
Codes of new generation – sustainable platform for numerical modeling of installations in the Proryv project

Since 2010, domestically produced software required for design decision making and safety assessment of nuclear power plants with fast reactors has been developed in Russian Federation under “Proryv” project, namely, under one of its subprojects – Codes of New Generation.
As a priority, a task was set and successfully accomplished covering the development of 24 software products addressing various areas: neutronics (Monte-Carlo method – MCU-FR, kinetic approximation – CORNER, diffusion approximation – DOLCE VITA); thermal hydraulics (DNS and LES approach – CONV-3D, RANS and LES approach – LOGOS, channel approximation – HYDRA-IBRAE/LM); thermomechanics and fission products release from fuel rods (BERKUT-U); probabilistic safety analysis (CRISS 5.3); radiation effects of releases (outside the industrial site boundaries – ROM, within the industrial site boundaries – ROUZ, in freshwater bodies – SIBILLA); dynamics of soil and groundwater contamination due to radioactive and chemical substances migration (GeRa); modeling the balance of material and nuclide flows within closed nuclear fuel cycle (VIZART) and others. Special attention was given to the development of integral multiphysics codes for safety assessment of power units – EUCLID and radiation safety justification – COMPLEX. The state-of-the-art mathematical models and effective numerical algorithms are used in the codes. They are developed by cooperation of leading Russian research centers and can be effectively used both on personal computers and high-performance computing systems. The modern approaches to collective software development resulting in a significant improvement of software quality are applied in the process of code development.
By the end of 2020, TRL level of the developed software averaged to 8.5 (TRL level 9 suggests that Rostechnadzor certifies the software, as well as the code is put into production). Large-scale validation on experimental data obtained on operating reactor units (BOR-60, BN-600, BN-800), as well as on unique small-scale experiments focused on investigation of separate phenomena conducted recently in the Russian Federation can be considered as an undoubted advantage of the codes developed.
The codes of new generation are actively used for the safety assessment of nuclear facilities, to teach student in universities, to model benchmarks under IAEA CRP and are considered by other industries as a promising software allowing to address production tasks.
The contribution presents the basics for new generation code development and briefly overviews the state-of-the-art in their development, verification and validation, as well as the plans for their further evolution.

Speaker: Nastasia Mosunova (Nuclear Safety Institute of the Russian Academy of Sciences (IBRAE RAN))
• 232
Dutch Thermal Hydraulic Design and Safety Support for LMFRs

Liquid metal fast reactor have a prominent role in the roadmap of the Dutch nuclear stakeholders. As nuclear service provider in the Netherlands, the Nuclear Research and consultancy Group (NRG) has established an elaborate program on liquid metal thermal hydraulics. This paper describes the thermal hydraulic design and safety support activities of NRG. The paper will start with the development of tools to allow thermal hydraulic system analyses. For this, the SPECTRA code, under development at NRG, has been adapted to facilitate to application of various liquid metals. The paper will provide a short description of the tool and show some examples of liquid metal applications. Since liquid metal reactors typically employ large liquid metal pools in which 3-D effects are unavoidable, a generic multi-scale modelling approach is being developed coupling the SPECTRA code to CFD codes. The paper will update the reader on the progress. Gradually, the paper will zoom in on more detailed analyses employing CFD codes for liquid metal pools, including important components like pumps, heat exchangers, and the core. For the core, past and ongoing activities will be shown related to validation of CFD approaches based on assemblies as they are designed on the drawing board, but also application and where possible validation of deformed and blocked fuel assemblies. Finally, the fundamental activities on understanding and pragmatic engineering model development for turbulent heat flux will be presented.

Speaker: Ferry Roelofs (NRG)
• 233
Verification of SARAX Code for the Transient Analysis of Sodium-cooled Fast Reactor

This paper describes the verification work of SARAX code for the transient analysis of a sodium-cooled fast reactor (SFR). The Advanced Burner Test Reactor (ABTR) benchmark created by Argonne National Laboratory (ANL) was modeled and calculated. The reference core is the 250 MWt sodium-cooled fast reactor, which includes neutronics calculation of the core at the beginning of equilibrium cycle, and also several transient analysis sequence such as ULOF (Unprotected Loss-of-Flow) accident. The SARAX code is a neutronics analysis package developed by the NECP team at Xi’an Jiaotong University and aiming for the advanced reactor R&D. It consists in a cross-section generation code named TULIP, a steady state neutronics calculation code named LAVENDER and a transient analysis code named DAISY. In this paper, the 33-group homogenized cross sections of all materials were generated using TULIP. LAVENDER gave the results of steady state parameters like power distribution, critical control rod position, reactivity coefficients and kinetics parameters. Then, DAISY simulated the transient progress with a space-dependent point-kinetics model and a parallel multi-channel thermal-hydraulics model and gave the results of peaking fuel temperature, cladding temperature and coolant temperature. The simulation of ULOF transients showed that SARAX gave comparable results with the design code of ANL and the SAS4A code, which verified the complete code system for transient calculations of SFR.

Speaker: Xiaoqian JIA
• 234
Phénix Control Rod Withdrawal test analysis using a multiphysics methodology

Before the definitive shutdown of the Phénix reactor, a series of end of life tests were performed in 2009 and 2010, by CEA (Commissariat à l’Energie Atomique et aux Energies Alternatives), EDF (Électricité de France) and AREVA. The main objectives were to enlarge experimental database for the research and design of Sodium cooled Fast Reactors (SFR). Due to this important opportunity, the IAEA (International Atomic Energy Agency) decided to establish a Coordinate Research Project from 2007 to 2011 to stimulate computational codes validation among different countries involved in fast reactors development. In this context, a benchmark was established on the “static Control Rod Withdrawal Test” (CRW) with the objective of the investigation on the flux and power local deformations related to different control rod insertions in the core. Such valuable experimental data are useful to improve calculation schemes used to analyze control rod withdrawal transient, which could potentially trigger a core melting accident in a SFR. The objective of the current study was to perform multi-physics simulation based on a loosely coupled approach to take into account local Doppler feedback effect on power deformations. The probabilistic particle transport code Serpent 2 (VTT Technical Research Centre of Finland, Ltd), associated with the JEFF-3.1 nuclear library, was chosen as reference neutron calculation code and was coupled to an in-house static thermal-hydraulic solver. Firstly, purely neutron transport calculations were done in order to build-up and check the overall core model. The uncoupled results were compared to benchmark results already published and show a correct accordance with experimental and calculated results providing that a heterogeneous description of control rods was used. A convergence study to estimate the required precision level of neutron calculations with respect to multiplication factors and power estimations was also performed. Secondly, coupling between neutron and thermal-hydraulics solvers were done through the Serpent 2 multiphysics interface with a regular exchange of the main coupled parameters such as fuel temperatures and neutron deposited powers for each axial node of each subassemblies of the fissile core. The coupled results on the power deviation are globally slightly nearer to the experimental ones than uncoupled results but are affected by probabilistic uncertainties and batch-to-batch inter correlation problems responsible for light power oscillations with respect to the number of simulated neutrons instead of a straight convergence.

Speaker: Ms Barbara FORNO (CEA Cadarache)
• 235
Simulation of FFTF Individual Reactivity Feedback Tests with SAS4A/SASSYS-1 Code

The Fast Flux Test Facility (FFTF) at the Hanford site in Washington was a 400 MW thermal, oxide-fueled, liquid sodium cooled test reactor, built to assist development and testing of advanced fuels and materials for fast breeder reactors. FFTF operated from 1980 until 1992, providing the U.S. Department of Energy (DOE) with the means to test fuels, materials, and other components in a fast neutron flux environment. One of the FFTF passive safety demonstration tests simulating loss-of-flow conditions without scram (LOFWOS) is currently being analyzed by the international community under an IAEA coordinated research project. In preparation for the passive safety demonstration tests in Cycle 8C, a series of individual reactivity feedback tests were carried out in FFTF. The primary goal of these tests was to check the core reactivity feedbacks in a systematic fashion by subjecting the core to various conditions of power, flow, and inlet temperature. These tests were carried out in Cycle 8A and consisted of quasi-static steps, where after each change the reactor was held at steady-state conditions for a period of about one hour to adjust to new steady-state conditions. The entire Cycle 8A test campaign consisted of about 200 steps. Each step was designed to simulate and validate specific features and reactivity feedbacks of the FFTF core. There were seven types of these individual reactivity feedback tests, targeting fuel reactivity Doppler and axial expansion feedbacks, coolant density feedback, structure reactivity feedbacks such as core radial expansion, as well as integral tests, such as the power reactivity coefficient. The data from the FFTF individual reactivity feedback tests provides a unique opportunity for validation of reactivity feedback modeling in fast reactor analysis codes. SAS4A/SASSYS-1 is one such safety analysis code that was developed at Argonne for transient simulation of liquid metal-cooled fast reactors. The structure of these tests provides data for code validation in a systematic fashion by separating reactivity feedbacks as much as was practically achievable. The quasi-static nature of these tests also simplifies code validation by eliminating the transient effects. This paper presents the results of the application of the SAS4A/SASSYS-1 code to a number of FFTF individual reactivity feedback steps, and compares the code predictions with the test data. The SAS4A/SASSYS-1 results show overall good agreement with the test, but at the same time several model improvement options were identified in this work.

Speaker: Anton Moisseytsev (Argonne National Laboratory)
• 236
Current status of development of 3D DNS CONV-3D code: one- and two-phase flow models

In IBRAE RAN in “Codes of New Generation” subproject of “Proryv” project one- and two-phase models are being developed to simulate heat and mass transfer processes in the separate elements of nuclear reactor. Those models are realized in the LES and DNS CONV-3D code.
The one-phase models are based on the algorithms with small scheme diffusion, for which the discrete approximations are constructed with use of finite-volume methods and fully staggered grids. For solving convection problem the regularized nonlinear monotonic operator-splitting scheme has been developed. The Richardson iterative method with FFT solver for Laplace’s operator as preconditioner is applied for solving pressure equation. Such approach to the elliptical equations with variable coefficients gives multiple acceleration in comparison with the usual conjugate gradients method. For modeling of 3D turbulent flows both DNS and LES approaches are used.
The one-phase module of CONV-3D code is fully parallelized and has perfect scalability, thus it is effective on high-performance computers such as “Lomonosov” (MSU, Russia). The one-phase module has been validated against data of well-known and just got experimental data for various liquids, including lead and sodium used as coolants, in a wide range of Rayleigh numbers between 10^6 and 10^16, and Reynolds numbers in the range of 10^3 – 10^5.
The two-phase models take into account interphase heat and mass transfer, stratification of the two-phase flow and separation of the gas component through the interface using equations of state such as condensed gas and the Noble-Able. The two-phase module in CONV-3D code is fully parallelized and has perfect scalability on a CPU and GPU systems. The algorithm of two-phase module is based on the use of HLL (Harten-Lax-van Leer) and HLLC (Harten-Lax-van Leer-Contac) solvers and two-step MUSCL (Monotonic Upstream-centered Scheme for Conservation Laws) predictor-corrector. The validation base includes experiments in which the heat and mass transfer and sodium boiling in the pipes were investigated.
This contribution presents several examples of code application for solving such problems as flow in fuel assemblies, tubes and ring channels, as well as natural convective flows in the elements of reactor. The results of two-phase flows modeling on the series of tests, including the problem of sodium boiling in a round pipe, are also shown.
In all cases the good agreement of numerical predictions with experimental data has been found, that specifies the applicability of the developed CONV-3D code to solve CFD problems for designing and operating NPPs.

Speaker: Dr Vladimir Chudanov (IBRAE RAN)
• 237
Development of Multi-level Simulation System for Core Thermal-hydraulics Coupled with Plant Dynamics Analysis - Prediction of Transient Temperature Distribution in a Subassembly under Inter-subassembly Heat Transfer Effect -

In the design study of sodium-cooled fast reactor, various activities from sensitivity analysis on whole plant dynamics using simple model to detailed analysis on local phenomena of interest are being performed. In conventional way, the analyses on whole plant dynamics and local phenomena are performed individually and the mutual interaction between them are considered through the settings of boundary conditions for each individual analysis. The final result through the individual analyses may contain excessive conservativeness. Therefore, JAEA has developed the multi-level simulation system in which detailed analysis codes for local phenomena of interest are coupled with a plant dynamics analysis code in order to obtain evaluation results considering the mutual interaction with reasonable conservativeness by updating the boundary conditions successively in coupling process.
In this study, focusing on core thermal-hydraulics, the coupling analysis method using a plant dynamics analysis code named Super-COPD and a subchannel analysis code named ASFRE to evaluate temperature distribution in a subassembly during the transient from forced circulation to natural circulation has been developed as a part of multi-level simulation system. During the transient, the consideration of thermal interaction between whole core and in-subassembly is important because flow re-distribution is caused and temperature distribution in a subassembly is affected by that in adjacent subassembly due to inter-subassembly heat transfer effect in the core. In the coupled analysis using the sequential two-way method, Super-COPD runs first to update flow rate, inlet temperature and heat flux on the boundary wall of the subassembly for ASFRE calculation. Subsequently, ASFRE calculates thermal-hydraulics and pressure drop in the subassembly and turns to Super-COPD calculation in the next time step.
After confirmation of basic functions of the coupling method in a simple geometry, the numerical analyses on the test in EBR-II (SHRT-45R) were performed with two models of the specific subassembly of XX09 with thermo-couples; one was the subchannel model of ASFRE in the coupling method and another one was included in the core model of Super-COPD. Through the comparison of the temperature distributions among the results using only Super-COPD and the coupling method, and the measurement in XX09, it was shown that the coupled analysis could predict transient temperature distribution in a subassembly under inter-subassembly heat transfer effect and it was indicated that the multi-level simulation by changing the level of detail of the analysis model between the method with the plant dynamics code and the coupling method could be performed.

Speaker: Dr Norihiro Doda (Japan Atomic Energy Agency)
• 238
Development of SFR core degradation simulation code SIMMER-V and its validation & verification studies

In the framework of reactor safety analysis of Sodium cooled Fast Reactors (SFR) using applicable code systems, the CEA and JAEA are involved in the achievement of SIMMER-V (owned by JAEA and co-developed by JAEA and CEA) developments dedicated to the simulation of the Severe Accident (SA) events of SFR.

The demand for new physical models into SIMMER-V rises in SIMMER previous versions related works, since new features are necessary in order to give a mechanistic evaluation of SA events. In order to enhance functions of SIMMER-V code system, a ranking of additional models has been proposed i.e. the detailed pin model or more adapted heat exchange correlations. Regarding the importance of predicting severe accident progression and consequences, verification and validation (V&V) work has been planned and conducted to get high value feedback on the applicability of the SIMMER-V to SFR severe accident simulation.

This paper outlines the CEA-JAEA collaboration on SIMMER-V development including V&V work, its achievement and perspective. The following two exercises of v&v work are highlighted.

First exercise presented in this article will consist on the validation of a natural convection correlation in fuel molten pools through the SCARABEE experimental program dedicated to the study of the Total Instantaneous Blockage (TIB) severe accident scenario in a SFR. Among the different tests performed, the program BF (“Bain Fondu” which means molten pool in French) aimed to study the behavior of molten and boiling heat-generating fuel pools using real reactor materials. Then, this study focuses on the evaluation of BF tests input data set using the latest SIMMER-V version and on the implementation and validation of the Chawla-Chan natural convection correlation in SIMMER-V sources.

The second exercise will focus on the verification of fluid dynamics scheme on SIMMER-V considering the updated test (2019) of classical benchmark problem of the Ideal Shock Tube, which was already part of SIMMER-III validation test basis. The main objectives are 1) to confirm the capability of SIMMER- V to simulate, in 1-D geometry, a single compressible fluid flow with strong pressure gradients, and 2) to verify the stability of results when varying mesh discretization, pressure solver options, time step control, domain decomposition and check basic conservation laws.

Speaker: Dr Elena Martin Lopez (CEA)
• Lunch Break
• 3.3 Reprocessing, Partitioning, and Transmutation

Chairs: Akira Yamaguchi and Amparo Gonzalez

• 239
FEASIBILITY STUDY OF HETEROGENEOUS TRANSMUTATION OF AMERICIUM IN FAST REACTORS

The most dangerous of the minor actinides is americium. Transmutation of external americium in the fuel of a fast reactor is possible when its content is over than 1% heavy atoms, however the lower content of an americium, on the contrary, it will accumulate. But curium isotopes with a high heat release are formed from it, complicating the unloading of spent assemblies. Therefore, the content of americium in the fuel should not exceed 1% (which corresponds to the equilibrium state and actually closes the possibility of transmutation of external americium), and the retaining time such fuel in the in-reactor storage should be at least 2 years.
Many researchers believe that heterogeneous transmutation in separate assemblies or blankets is preferable. However, the concentration of americium transmutation products in a small number of burnout assemblies will lead to a manifold increase in the residual heat release in them, and the discharge of such assemblies from the reactor will become very problematic.
Heterogeneous transmutation in the blankets of devices with a strong moderator (zirconium or yttrium hydride) seems to be more rational. Theoretically, this method makes it possible to convert all loaded americium into fission products in one campaign, eliminate the need for multiple handling of it and its transmutation product - curium, and also eliminate the problem of high residual heat release. In this way, all "own" americium, which is formed in the fast reactor, can be converted into fission products.
At the same time, it seems economically feasible to burn out in fast reactors not americium itself, but its predecessor, 241Pu. This is possible due to the use of "fresh" plutonium from VVER spent fuel, which will allow reducing the annual production of americium by almost 2 times without developing expensive technologies.
An unpleasant feature of neptunium transmutation is the formation of the plutonium-236 isotope, which decays into uranium-232 and then into a whole series of high-energy gamma emitters. Therefore, burning out neptunium in fuel should be recognized as inexpedient, and burning it out in the same irradiation devices with moderator as americium seems to be the most preferable.

Speaker: Prof. Andrey Gulevich (IPPE)
• 240
The role of pyrochemical processing in a NetZero economy in the UK

The programme described here is part of a UK government investment in Nuclear to be delivered by a collaboration of UK Government, The UK National Nuclear Laboratory (NNL), Industry and Academia. The programme will contribute to international understanding of development and demonstration requirements for pyro-processing as well as emerging applications for salt separations enabling a broad base of future fuel recycling. This paper will outline the programme and present early findings.

Electrorefining as an actinide separation technique offers potential to provide useful fissionable material as part of a closed fuel cycle. It is specifically designed for irradiated metallic fuel, but other fuel types can be processed with pre-treatment; it uses a molten chloride salt electrolyte such as a eutectic mixture of lithium chloride and potassium chloride. The production of a baseline flowsheet model and demonstration of key engineering aspects such as scalability and in-line process monitoring will raise the technology readiness level of pyro-processing technology in the UK.

Basic molten salts data capture is limited in the UK, data will be generated by NNL and university partners through post-doctoral research programmes and physical properties and materials data captured in a centralised database. This will not only be used longer term to develop potential future facilities but is being used directly to inform other areas of the wider pyro-processing project such as the flowsheet development and pilot-scale demonstrator work packages.

A lab-scale Pu-active pyro-processing rig being developed providing a key UK capability. This rig will provide new data on the behaviour of Pu and other components in the molten salt. It will complement the concept design of a pilot-scale rig is being developed to investigate the engineering practicalities of operating a full-scale pyro-processing plant such as scalable pumping systems, instrumentation and control (process and accountancy), materials of construction etc. this will also allow us to assess the impact of introducing a fluoride-based salt matrix.

The activities and outputs from the Salt Engineering and Salt Science activities will enable NNL to evaluate, on behalf of the UK, the case for recycle in advanced fuel cycles at a broad level and set out the potential drivers or switching points between advanced aqueous and pyro-processing technologies. The results will inform the direction of any follow-on project through the development of a future UK roadmap in pyro-processing and identify opportunities for international collaboration and leveraged investment.

Speaker: Dr Michael Edmondson (The National Nuclear Laboratory)
• 241
Fabrication and reprocessing of mixed uranium-plutonium nitride fuel for reactor BREST

Dense nuclear fuel for fast reactors (FR) is the preferred option. In the Russia, as part of the “PRORYV" project, the development of key technologies of closed nuclear fuel cycle (CNFC) for FR with dense mixed nitride uranium-plutonium fuel (MNUP) is underway. MNUP is a new complex product in the field of nuclear power technologies. CNFC with FR ensures:
− no spent nuclear fuel (SNF) accumulation;
− technological support for the non-proliferation regime;
− competitiveness with other large-scale energy technologies.
For industrial implementation FR CNFC on the basis of MNUP fuel an experimental demonstration energy complex (EDEC) is being created at the Siberian Chemical Combine site. It consists of BREST-300 reactor with lead coolant and CNFC facilities. The latter includes a MNUP fuel fabrication/refabrication module (FRM) and SNF reprocessing module (RM). In 2022 it is planned to put into operation the FRM and in 2024 to begin construction of RM of EEDC. At the FRM the technology of carbothermal synthesis of MNUP fuel will be implemented. At the RM the technology of combined (pyro+hydro) and hydrometallurgical reprocessing of FR SNF are under development.
To date, R&D have been conducted to justify the use of MNUP fuel in FR. Over 1000 fuel pins with MNUP fuel have been successfully irradiated in BN-600. A complex of post-reactor studies, including destructive radiochemical studies, has been conducted.
R&D on the reprocessing of MNUP SNF includes the fundamental possibility of using pyroelectrochemical technological operations have been shown. The technical feasibility of the following hydrometallurgy operation has been demonstrated experimentally:
• Voloxidation of the MNUP SNF (recovery > 99.9 % tritium and 98 % 14C);
• Extraction and crystallization refining of U+Pu+Np mixture (purification factor of 5*106);
• Recovery >99.9 % of actinides including Am and Cm;
• Microwave denitration for mixed U-Pu-Np, U-Am, U-Cm oxides preparation;
• Separation 1 g of Am and 0,1 g of Cm;
• Waste vitrification in borosilicate glass in a remotely removable cold crucible.
An integrated system of models and codes for all technological modifications of the non-reactor part of CNFC and for coordinated modelling of heterogeneous processes and phenomena are under development. The system under development uses both existing and newly developed models and codes designed to describe technological processes and apparatuses, nuclear and radiation safety, criteria of ignition, combustion, behaviour of structures and engineering systems under critical loads, etc.

Speaker: Dr Andrei Shadrin (JSC "Proryv")
• 242
HETEROGENEOUS BURNING OF MINOR ACTINIDES IN A FAST REACTOR

Transmutation of minor actinides (MA) into stable or short-lived ones by their irradiation in reactors will alleviate the problem of long-term activity of spent nuclear fuel (SNF), increase the efficiency of nuclear fuel due to energy produced by MA fission, and also accumulate and produce useful radionuclides. The economic efficiency of closed-cycle nuclear power cannot be achieved without MA disposal and safe final isolation of radioactive waste.
Fast reactors are the most suitable for homogeneous MA transmutation and heterogeneous MA burning.
With homogeneous transmutation, MA in a small amount (less than 5%) is introduced into the standard nuclear fuel. With this approach, MA will be both burnt and accumulated from MA introduced into nuclear fuel, as well as from uranium and plutonium of standard nuclear fuel. During repeated recycling of such nuclear fuel, its nuclide composition stabilizes and MA accumulation rate is compared with their decrease rate, and the equilibrium SNF nuclide composition is reached.
The concept of heterogeneous MA burning involves their inclusion into inert matrices (no uranium and plutonium) and placement in separate fuel assemblies (fuel rods) either in the fast reactor core or blanket.
Heterogeneous MA burning in the fast reactor blanket has a more flexible strategy for MA handling than homogeneous MA transmutation and can be used to achieve high MA burning with minimal effect on reactor characteristics. The use of inert matrices will avoid the formation of secondary MA.
In Russia, there is a unique opportunity for MA transmutation in existing fast reactors (BN-600, BN-800). Therefore, a technology for MA separation from SNF and production of fuel with MA should be developed, and scientific research and reactor experiments should be performed.

Speaker: Alexander Tuzov (JSC "SSC RIAR")
• 243
Advanced Flow-Sheet for Partitioning of Trivalent Actinides from Fast Reactor High Active Waste

The processes developed for partitioning of trivalent actinides (TA) from high-level liquid waste (HLLW) generated during reprocessing are all focused on single-cycle approaches for waste minimization. In this process the formation of third phase has to be avoided. Hence, a phase modifier is often employed in most of the processes in vouge, even though the use of the later is more desirable. To avoid these complications advanced symmetrical and unsymmetrical diglycolamides were developed in our laboratory and systematically studied for the group separation from fast reactor simulated high-level liquid waste (SHLLW), and then subjected to single-cycle separations. This method involved the separation of trivalent actinides and chemically similar lanthanides, as a group, from SHLLW followed by mutual separation of lanthanides and actinides from the loaded organic phase using aqueous soluble complexing agents.

The potential solvents identified for the group separation of trivalents from HLLW were 1) 0.2 M TODGA (N,N,N’N’-tetraoctyldiglycolamide) + 5% octanol / n-DD, 2) 0.2 M TODGA + 0.5 M TBP (tri-n-butylphosphate) / n-DD, 3) 0.1 M TODGA + 0.25 M HDEHP (di-(2-ethylhexyl) phosphoric acid) / n-DD, 4) 0.2 M TDDGA (N,N,N’N’-tetradecyldiglycolamide) / n-DD, 5) 0.2 M D3DODGA (N,N-didodecyl-N’N’-dioctyldiglycolamide) / n-DD, 6) 0.4 M DOHyA (N,N-dioctyl-2-hydroxyacetamide) / n-DD. The selective stripping of Am (III) from the loaded organic phase containing trivalent lanthanides was investigated using aqueous soluble bis-1,2,4-triazine derivatives such as SO3-Ph-BTP, SO3-Ph-BTBP and SO3-Ph-BTPhen in dilute nitric acid solution. The results revealed that the SF of Eu (III) over Am(III) decreased with increase in the concentration of nitric acid in all cases and separation factor (SF) decreased in the order SO3-Ph-BTP >SO3-Ph-BTBP >SO3-Ph-BTPhen. The co-stripping of lower lanthanides (La, Ce, Pr, Nd) was also observed during the recovery of Am(III). The distribution ratio of Am(III) and Ln(III) in all the organic phases were quite similar. However, TODGA system requires 1-octanol phase modifier for preventing the third phase formation, which is undesirable for safety concerns, whereas the other ligands were modifier-free reagents. Therefore, the other ligands (TDDGA and D3DODGA) developed in our laboratory offered a significant advantage over the TODGA. D3DODGA stripped Am(III) better. As a result the SF for Eu(III) over Am(III) was significantly high (>400) indicating the possibility of using them for MA partitioning. This presentation describes, the summary of our research and development activities carried out at IGCAR towards the development of advanced flow-sheet for TA separation from HLLW generated during fast reactor fuel reprocessing.

Speaker: desigan narasimhan (IGCAR, DAE)
• 244
Multi-criteria comparison of the efficiency of minor actinides burning in different nuclear reactors based on the INPRO/IAEA KIND approach

This paper presents a comparison of the efficiency of minor actinides (MA) burning in various type of nuclear reactors with a fast neutron spectrum. A set of criteria for comprehensive comparison of reactor technologies, based on the INPRO/IAEA KIND approach to multi-criteria assessment, has been prepared. This set of criteria includes indicators in such areas as the efficiency of MA burning, economics, safety, environment, readiness of reactor technology and infrastructure for its implementation. The evaluation and comparison procedure was carried out using the KIND-ET tool. It is shown that a comprehensive multicriteria analysis of various aspects of the technologies, as expected, led to estimates that differ from the approach in which technologies are compared only single criterion and without taking into account the influence of other equally important factors. And the cumulative assessment of technologies largely depends on the set development objectives. This means that each of the listed options can take the first place in the rating when certain priorities are selected.
Keywords: multi-criteria evaluation, sodium fast reactors, lead fast reactors, MSR, technology comparison, project KIND, IAEA, INPRO.

Speaker: Stepan Kviatkovskii (SC Rosatom)
• 245
Physical feasibility of MA transmutation in a two-component nuclear energy system in Russia

The transition to a two-component nuclear power structure using thermal (TR) and fast reactors (FR), as asserted by the «Russian nuclear power development strategy to 2050 and outlook to 2100» (Strategy-2018), is directed at finding optimal solutions and resolving relevant issues pertaining to the currently established nuclear energy system in Russia. A core issue in this regard is managing the long-lived MA inventory, which have a substantial impact on overall nuclear power radiological safety for time-frames that could be considered historically significant. The study presents findings related to analyzing the capability of a commercial fleet of FRs to successfully resolve these issues by including MA in the closed nuclear fuel cycle without any fundamental changes to their expected characteristics regarding safety and competitiveness parameters.
Three different Russian scenarios were considered in the study: 1) full transition to a large-scale FR dominated nuclear fleet reaching 92 GWe capacity by the end of the 21st century, 2) mixed composition of VVER (43% capacity) and FR (57% capacity) with comparable installed capacity 3) a moderate nuclear power development scenario reaching 72 GWe by the end of the century with a 57% VVER and 43% FR mix. At this rate it is calculated that 36-67 tonnes ofAm and 67-120 tonnes of MA (Am+Np) would be accumulated from the VVER fleet.
Two algorithms are proposed for Am and Np utilization. In the first approach, MA and Pu obtained from reprocessing VVER spent fuel are simultaneously used for FR start-up, after which the fuel will reach an equilibrium state following multiple recycling in the fast reactor. The maximum concentration of MA content in the fuel was calculated to be at 2% (1,1% Am and 0,9% Np), and 0,5% when the fuel reaches equilibrium state.
If radiation-related limiting factors for handing nuclear fuel with high MA content are taken into account, comparable MA utilization efficiency could be achieved with lower MA fuel concentration if they are introduced evenly throughout the FR operation lifecycle. Continuous addition of MA from VVER reactors to FR fuel with 2% MA concentration will increase MA utilization threefold compared to the first approach.
The results of the study conclude that the MA arising from VVER spent fuel accumulation in all scenarios considered could be successfully utilized without dedicated MA-burners, although the complexity of the issue intensifies as fewer FRs are introduced into the power mix with increased MA content in their fuel.

Speaker: Yury Khomyakov (Private institution «Innovation and technology center for the «PRORYV» project»)
• 246
Investigation of the anodic processes on the ceramic anode in the oxide-chloride melts

Alkaline halide melts and alkaline earth metals are used for the electrochemical reduction of metal oxides to their metallic forms. In practice, fluoride, chloride, and mixed chloride-fluoride melts of alkali and alkaline earth metals are used most often. Graphite is usually applied as an inert anode material in these media. However, during the electrolysis of oxide-halide melts, carbon is not an inert anode. In recent years, metals and their alloys, metal oxides, and cermets have been considered to be the candidate inert anode materials for the electrolysis of oxide-halide melts. Almost all investigated metal anodes made of individual metals such as iron, nickel, copper, chromium, and their alloys, except some noble metals are unstable in an oxygen atmosphere and oxidize at high temperatures.

At present, oxide-chloride melts based on LiCl is used for the electrolytic reduction of uranium and plutonium oxides with lithium formed at the cathode. Platinum metal is usually used as the anode material for oxygen evolution in these melts. However, platinum is highly susceptible to corrosion in these melts and therefore is not the inert anode material. Also, platinum is an expensive metal, which makes it difficult to use as an anode material in the industry. We have carried out studies of anodic processes and electrolysis tests on the ceramic anode NiO-Li2O in LiCl-KCl-Li2O melts. Studies have shown that ceramics NiO-Li2O is the inert anode material for electrolysis of LiCl-KCl-Li2O melts at temperatures of 550-650˚C. Voltammetric studies have shown that two electrode processes can occur on the NiO-Li2O anode: 1) oxidation of oxide ions with the formation of gaseous oxygen up to potentials of 2.8-2.9 V vs E (Li+/Li) and 2) chlorination of the anode material at potentials more positive than 3.0-3.1 V vs E (Li+/Li). Experiments carried out in the process of electrolytic reduction of UO2 and UO2 with additions of rare earth oxides shown that NiO-Li2O is found to be the inert anode material. The anode current efficiency of oxygen evolution at this anode is close to 100%. As a result of electrolysis experiments during 35 h, the diameter and length of the anode sample did not decrease. Thus, ceramics NiO-Li2O can be used as the inert anode material for electrolysis of oxide-chloride melts based on LiCl.

Speaker: Dr Alexander Dedyukhin (Institute of High Temperature Electrochemistry of the Ural Branch of the Russian Academy of Sciences)
• 6.4 Simulation Tools for Safety Analysis

Chairs: Evgeny Ivanov and Ian Hill

• 247
Models of the integral EUCLID/V2 code for numerical simulation of severe accidents in a sodium-cooled fast reactor with MOX and MNUP fuels

For the modeling of severe accidents in a sodium-cooled fast reactor coupled multiphysics EUCLID/V2 code is being developed in Russian Federation in Codes of New Generation subproject of “Proryv” project. Multiphysics code allow calculating all relevant processes occurring during severe accident: reactor power change including due to boiling and melting, coolant boiling and dryout, cladding and fuel melting (for MOX fuel), as well as fuel dissociation (for MNUP fuel), movement and solidification of the resulting melt, the formation of a pool of melt, the release and transport of fission products in the reactor and beyond and others.
To simulate thermohydraulic processes HYDRA-IBRAE/LM module is used in the EUCLID/V2 code. This module simulates processes in one- and two-phase coolant flow. For fuel rod behavior modeling the BERKUT module is used. The processes of core damage are represented by the severe accident module SAFR. Also EUCLID/V2 code contains the DN3D neutronics module and the AEROSOL-LM module for calculation of the FP transport in the reactor facility and in the containment compartments. MCU-FR module based on Monte-Carlo method is used to estimate the secondary criticality of the core configuration during severe accident.
In the contribution the brief description of each module is given as well as the algorithms used to make the computational grids of all modules to be consistent and for modules coupling. Some results of the EUCLID/V2 V&V calculations are also presented.

Speaker: Dr Eduard Usov (IBRAE RAN)
• 248
Development of Integrated Severe Accident Analysis Code, SPECTRA for Sodium-cooled Fast Reactor

Analytical evaluation of severe accidents (SAs) in sodium-cooled fast reactors (SFRs) becomes increasingly important. The progress of the SAs has been previously evaluated by transferring the analytical results between the multiple analysis codes with different roles. In this study, a new code named SPECTRA (Severe-accident PhEnomenological computational Code for TRansient Assessment) was developed for integrated analysis of the in- and ex-vessel phenomena. This paper provides the newly developed analytical models and the analysis of a loss of reactor level (LORL) event as one example of the SAs.
The SPECTRA code consists of the in- and ex-vessel modules which have a thermal hydraulics module as a base part. The in-vessel thermal hydraulics module computes complicated multi-dimensional behavior of liquid sodium and gas by using the multi-fluid model considering compressibility. Relocation of a molten core is computed by the dissipative particle dynamics method which has an advantage from the viewpoint of its wide applicability. A lumped mass model is employed for computation of the ex-vessel multi-component gas including aerosols. The fully implicit scheme is applied to the both thermal hydraulics modules in order to enable computation with a large time step width. The analytical models for sodium fire, sodium-concrete interaction, and debris-concrete interaction are integrated into the ex-vessel thermal hydraulics module. The in- and ex-vessel modules are coupled by exchanging the amount of leaked sodium and debris at every time step.
The LORL event is considered as one example of the SA scenarios. A sodium coolant leaks from a damaged pipe in a primary cooling loop and causes sodium fire. In case a molten core and sodium leak from a damaged lower head of a reactor vessel (RV), sodium-concrete interaction and debris-concrete interaction occur in the compartment under the RV. This event progress was computed in a simplified domain including the RV, the primary cooling loop, and the ex-vessel multi cells. The analytical result showed lowering of the liquid level due to sodium leak, boiling of the coolant around the core region, and molten core relocation in the in-vessel region. As for the ex-vessel region, the atmosphere temperature and pressure increased due to sodium fire, sodium-concrete interaction, and debris-concrete interaction. The basic capability to reproduce SA scenarios was demonstrated through this analysis.

Speaker: Dr Akihiro Uchibori (Japan Atomic Energy Agency)
• 249
Aerosol module for modeling of the fission product behavior in FR cooling circuits and NPP compartments

This contribution presents an overview of models of an aerosol module designed to simulate the behavior of fission products in the circuits and compartments of nuclear power units with fast reactors with sodium or lead coolants. Aerosol module AEROSOL-LM is included in the thermal-hydraulic HYDRA-IBRAE/LM code. Together they represent a unified code with a common interface for calculating the processes of thermal-hydraulics and the fission products transport both in gaseous and aerosol forms. The AEROSOL-LM module allows calculating the relevant processes of aerosol dynamics: nucleation, coagulation, condensation and sedimentation. A specific feature of the module is the simulation of multicomponent and polydisperse aerosols.
In particular, for sodium reactors the behavior of sodium combustion aerosols in NPP compartments is simulated. For lead cooled fast reactors the oxygen transport, the formation and behavior of corrosion particles are considered. The aerosols formation and transport between rooms of the NPP including those resulting from melt-concrete interaction are also modeled by the aerosol module.
The results of module validation are also briefly presented in the contribution.

Speaker: Andrey Sorokin (IBRAE, Moscow, Russia)
• 250
Analysis of the natural circulation capacity of decay heat removal system in pool-type sodium-cooled fast reactor

The structure of pool type sodium-cooled fast reactor (SFR) is complex, which leads to the complicated thermal-hydraulic phenomena in the process of natural circulation for decay heat removal. The determination of natural circulation flow path and the decay heat removal capacity of natural circulation of each flow path are issues to be considered in the design of SFR. Core flow distribution, flow and heat transfer of inter-wrapper flow, thermal stratification of the sodium pool, thermal hydraulic interaction between the core and the sodium pool, and arrangement of the decay heat removal system are factors that affect the decay heat removal capacity in the reactor. Therefore, this paper analyzes the influence of the arrangement scheme of decay heat removal system on the removal of decay heat.

Firstly, the system program THACS is used to establish the decay heat removal system of coupling primary circuit and external circuit of SFR, and the analysis is carried out for the condition of station black out (SBO). Secondly, sensitivity analysis is conducted for the arrangement scheme of the decay heat removal system, so as to evaluate the decay heat removal capacity of the reactor. Two decay heat removal systems selected for comparative analysis are non-penetrating direct reactor auxiliary cooling system (NPDRACS) and penetrating direct reactor auxiliary cooling system (PDRACS). The results indicate that both decay heat removal system arrangements can effectively remove decay heat from the core. And for large SFR, the decay heat removal capability of the PDRACS is better, because the cold sodium from direct heat exchanger (DHX) can cool the core assemblies directly.

Speaker: Dr Yapeng Liu (Xi'an Jiaotong University)
• 251
Regulatory Perspectives on Analytical Codes and Methods for Advanced Reactors

Analytical codes and methods are used extensively in the design and safety analysis of nuclear reactors. These are commonly used to analyse the response of a complex engineering system to postulated events with potentially severe health, financial, and environmental implications. Regulatory agencies establish requirements and/or expectations on the nuclear power plant designer or licensee for the development and use of analytical codes and methods in order to ensure the quality, credibility, and confidence in the analyses produced by the analytical codes and methods. In addition, regulatory agencies have used analytical codes and methods to perform confirmatory analyses as part of due diligence during a regulatory review. The Task Group on Analytical Codes and Methods (TGACM) of the OECD-NEA Working Group on the Safety of Advanced Reactors (WGSAR) has performed a review to (1) identify and clarify the requirements and best practices applicable to nuclear power plant designers for the development and use of analytical codes and methods used in the design and safety analysis of nuclear power plants, and (2) identify best practices for the use of confirmatory analyses by regulatory agencies.
In this paper first results of this on-going work will be presented, based on the responses to a survey from Canada, France, Germany, Italy, Russia, UK and USA. The partly different procedures and expectations on regulatory approval of codes and methods, quality assurance program and handling of possible bugs and errors will be discussed. For example, some countries require the code developer to undergo a certification process. In addition, in some countries qualification of code users and / or organisations is required.
The second part of the survey is related to confirmatory analysis. Because the objective of these confirmatory analysis is mainly linked to support the regulator, the required capabilities and expectations are partly different to codes used for the design and optimization of advanced reactors. Independent from the claimed inherent safety capabilities of reactor concept, a simulation of severe accident phenomena is expected by regulatory authority.
In the review an overview on existing codes used for advanced reactors will be provided. The capabilities of these will be discussed and compared with safety relevant phenomena of these advanced reactor concepts.
In conclusion, the regulatory expectations related to codes used for advanced reactors should be considered in the development of these codes. Comparing code capabilities with safety relevant phenomena, the review provides information on further code development needs.

Speaker: Walter Klein-Heßling
• 252
Models of the integral EUCLID/V2 code for numerical modeling of different regimes of lead-cooled fast reactor

The EUCLID/V2 integral multiphysics computer code is designed for the safety analysis and justification of the new generation NPPs with liquid metal cooled fast reactors under normal operating conditions, anticipated operational occurrences, design basis accidents and severe accidents. The EUCLID/V2 code includes the system thermohydraulics module (HYDRA-IBRAE/LM), spatial time-dependent neutronics module (DN3D/CORNER), quasi two-dimensional fuel rod module (BERKUT), the module of burnup and decay heat calculations (BPSD), the module of fission, activation and corrosion products transport in primary loop and gas system of a reactor facility (AEROSOL-LM), the tritium migration module (TRITIUM), the module of fission product source calculation, the fuel rod and core disruption module (SAFR), the modules of mass transfer and fission product transport calculation in the reactor containment compartments (HYDRA-IBRAE/LM or KUPOL-BR), the module of simulation of radiation situation beyond industrial site of a NPP (ROM).
The mentioned modules are multi-purpose, their models do not depend on a coolant or fuel type and may be used to simulate reactor facilities with sodium, lead or lead-bismuth coolant and nitride or oxide fuel. However, some special models needed for behavior simulation of reactor facilities with lead coolant have been implemented into the EUCLID/V2 integral computer code. They are the model of solid phase impurities transport in a primary loop of a reactor facility with heavy liquid metal coolant, the model for a steam generator tube rupture simulation of a reactor facility with the lead coolant, the model of fission product source calculation taking into account physicochemical interaction between the nitride fuel and the lead coolant, the nitride fuel dissociation model, the lead melt and concrete interaction model, the lead freezing model.
At present, the V&V of the listed above modules and models is being carried out on analytical and numerical tests and experimental results.
In the contribution the brief description of the above mentioned models and some V&V results are presented.

Speaker: Mr Dmitry Veprev (NSI RAS)
• 253
Target Accuracy Requirements and an evidence-based background for MSFR safety assessment

From the very beginning of the nuclear era, a mission of fast neutron reactors (FRs) has been foreseen [1], [2] in its double functionality, combining a generation of power and a conversion of fertile materials (U-238 and Th-232) into new nuclear fuels to make energy resources practically inexhaustible.
Since then, due to conjuncture changing, one used to consider fast reactors as elements of the global fuel cycle intended to play as traditional roles – to be a robust source of energy and artificial fissile materials, – as complementary ones – to burn plutonium and minor actinides accumulated in LWRs spent fuel [3].
At the same time, the nuclear engineering community considers, inter alia, some innovative FR concepts with fuel reprocessing fully immersed in a nuclear reactor like Molten Salt Fast Reactors (MSFRs) [4].
Over there, of course, enhanced flexibility of the fuel cycle in MSFRs should be supported by their extraordinary safety potential [3]. Unfortunately, in the case of MSFRs due to a fundamentally limited operational background, assessors have to rely, largely, on comprehensive simulations than on pure expert judgment.
Of course, these simulations, models, and relevant tools should be somehow validated against objective observations to ensure a sufficiency of their Predictive Capability Maturity (PCM) [5]. Yet any integral experiment (IE) taken alone cannot reproduce phenomena or processes of interest while an extrapolation basing on indirect data and Data Assimilation (DA) can [6].
These indirect IEs might be selected 1) focusing on phenomena essential as for reactor control as for accidental process management, 2) combining nuclear-driven and non-nuclear experimental data in the validation of relevant codes and models, and 3) establishing realistic Target Accuracy Requirements (TARs) for the best estimate and penalizing models.
We are discussing some factors of informativeness of IEs in terms of pre-defined “accidental states” (ASs)–subsequent phases of the core degradation.

[1] E.Fermi, The Future of Atomic Energy, United States, (1946),
http://www.osti.gov/accomplishments/documents/fullText/ACC0043pdf
[2] Nuclear Fuel Cycle Science and Engineering, 1st Edition, Ian Crossland, Woodhead Publishing, 2012
[3] J.Serp et al., The molten salt reactor (MSR) in Generation IV: Overview and perspectives, Prog. Nucl. Energy, 77, 2014
[4] I.Babuska, J.T.Oden, Verification and validation in computational engineering and science: basic concepts, Computer Methods in Applied Mechanics and Engineering, 193, 36–38, 2004
[5] G.Palmiotti et al, A global approach to the physics validation of simulation codes for future nuclear systems, Annals of Nuclear Energy, 36(3), 2009

Speaker: Dr Evgeny Ivanov (IRSN)
• 254
SOCRAT-BN INTEGRAL CODE: DEVELOPMENT, VALIDATION AND CURRENT STATUS

Integral computer codes SOCRAT-BN have been developed at the Nuclear Safety Institute of the Russian Academy of Sciences (IBRAE RAN) in the frame of the Federal Target Program «New-Generation Nuclear Power Technologies for the Period 2010–2015 and up to 2020». The first version SOCRAT-BN/V1 was developed for the period 2010-2014 to simulate design basis (DBA) and beyond design basis accident (DBDA) at a nuclear power plant with sodium fast reactor (SFR). In 2016, the first version of the code was certified by “Scientific and Engineering Centre for Nuclear and Radiation Safety” (SEC NRS). For 2014-2017, the second version of the code, SOCRAT-BN/V2, was developed. It had extended the first version to severe accident with core melting. The second version of the code was certified in 2019.
Currently, SOCRAT-BN is used for safety assessment of operating and projected SFR for Russian power plants. Also, it is planned to use SOCRAT-BN for supporting projects to be constructed abroad.
The physical models of the SOCRAT-BN are divided into two blocks: a steady state and a transient one. The steady state block is applied to simulate the accumulation of fission products and the state of the fuel elements for the period of time that the reactor operates before an emergency event. The transient block is applied to simulate the temperature state of the reactor, the transfer of fission products before release to the environment, deformation of fuel and cladding, neutronics processes during the destruction of the core, melting and movement of fuel.
The report represents description of the basic code physical models, its validation and the current state of the code.

Speaker: Mr Ruslan Chalyy (IBRAE RAS)
• 9.1 Education, Profesional Development, and Knowledge Management

Chairs: Antonella Di Trapani

• 255
TRAINING OF NEW GENERATION SPECIALISTS IN THE FIELD OF FAST NEUTRON REACTORS AND NUCLEAR FUEL CYCLE CLOSURE

The Strategy is based on the formation of a two-component nuclear energy based on a closed nuclear fuel cycle with fast neutron reactors.
The solution of long-term tasks of creating a two-component nuclear power with a closed NFC based on fast reactors is associated with the need to create a system for training, attracting and developing young professionals based on new principles.
Replenishment of the industry with young qualified personnel is a fundamental task that allows us to maintain the strategic direction of development.
Russia has a strong scientific and educational engineering and physical school. Russian nuclear education is one of the most advanced in the world. Currently, the formation of a two-component nuclear energy based on a closed nuclear fuel cycle with fast reactors poses for the profile educational system the task of training a new generation of researchers based on interdisciplinary knowledge, with fundamental training and practical skills. In addition, a new generation of specialist researcher should have a broad scientific outlook and modern management skills in science (principal investigator, etc.).
In order to solve the large-scale scientific and technological problems associated with fast reactors and closed NFC, a system of long-term planning for training qualified personnel, attracting and continuous development is also necessary, based on close cooperation of the industry with specialized universities.
The report discusses a comprehensive approach to the preparation and implementation of a program for the development of scientific and technical competencies for two-component nuclear power with a closed NFC based on fast neutron reactors. The key components of the program are considered, including the formation of a target plan for the training and retraining of personnel, improving the quality of educational programs, developing network interaction with specialized universities, and international cooperation in the field of education. The report describes the experience of NRNU MEPhI in the creation and implementation of interdisciplinary educational programs aimed at training a new generation researchers, as well as creating conditions for the development of university internationalization and the export of nuclear education.

Speaker: Mr Georgy Tikhomirov (NRNU MEPhI)
• 256
An e-learning tool on Fast Reactors and their Fuel Cycles

The United Kingdom was one of the countries pioneering fast reactor development and demonstration with the construction and operation of two fast neutron spectrum reactors at the Dounreay site; multiple support facilities including zero-power and thermal hydraulic facilities; and demonstrated several sustainable and closed fuel cycles. Since the end of the UK Fast Reactor Programme in 1994 much of the knowledge acquired has been lost or at best archived. The UK National Nuclear Laboratory (NNL), as part of the Department for Business, Energy and Industrial Strategy’s (BEIS) £505m Energy Innovation Programme – which includes the biggest nuclear fission investment in a generation, is leading on the Advanced Fuel Cycle Programme (AFCP). This work has begun activities aimed at ensuring this globally unique and highly valuable resource is not lost, through a Fast Reactor Knowledge Capture project.

A key feature of the AFCP is to support the next generation of technical experts, especially the development of these skills in early career individuals from across the nuclear sector. Using information captured from the UK’s historic fast reactor programme, a number of online e-learning training modules are to be developed in partnership with the International Atomic Energy Agency (IAEA). These resources are being developed as part of the recently announced NNL-IAEA Collaborating Centre on the Advanced Fuel Cycle. Through the Centre, the IAEA, NNL and UK partners will collaborate on a number of topics relating to the development of advanced fuels and fuel cycles required to power the reactors of the future. The collaboration will place particular emphasis on the exchange of technical expertise between the UK, IAEA and Member State representatives, and in supporting the development of the next generation of experts.

The proposed approach to develop the e-learning module is discussed, focusing on the modular nature to allow the addition to incorporate further topics as they are produced as well as the interactive nature. This will be followed by an overview of the content that provides part of the of the “Background and Introduction” section on Fast Reactors and the planned fuel manufacture and reprocessing content. In summary, the development of future fast reactor experts is a timely and costly endeavour. To support and accelerate the realisation of this aim, modern and interactive training techniques befitting the 21st century should be sought and developed that utilise the historic knowledge developed from research and reactor operations.

Speaker: Dr Nicholas Barron (National Nuclear Laboratory)
• 257
GEN IV INTERNATIONAL FORUM WEBINARS INITITATIVE

Collaboration and support among national laboratories, industry, universities, and research and development organizations are vital to not only maintain a skilled and competent nuclear workforce but also to avert the risk of human resource shortages. However, despite numerous efforts in coordinating and promoting nuclear education, there is still a lot to be done for developed and developing countries to either maintain and/or build a skilled nuclear workforce to address the increasing demand for technical skills. The Gen IV International Forum (GIF) Education and Training Working Group (ETWG) was created in November 2015 to respond to this demand, by proposing webinars that focus on advanced reactors systems and cross-cutting subjects. As of today (28 September 2020), the GIF ETWG has produced, podcasted and posted 46 Webinars covering the six Gen IV systems and various subjects addressing e.g. the economics of the nuclear fuel cycle, sustainability aspects of Gen IV systems, nuclear fuels and materials challenges, the thorium fuel cycle, energy conversion systems, and lessons learned for knowledge management and preservation. The GIF webinars are presented live by internationally recognized subject matter experts. They are recorded and archived at www.gen-4.org, and have been recently converted to the YouTube platform as video. The development of GIF webinars, with their expansion of topics, is intended to inform and stimulate not only junior scientists' interest, but also managers, key decision makers and the general public about advanced reactors introducing foreseen advantages but also key R&D to be developed. Details and examples of the GIF webinar modules from the initial concept to the full realization will be presented. Future topics for webinars that are planned beyond May 2021, will be announced.

Speaker: Dr Patricia Paviet (Pacific Northwest National Laboratory)
• 258
Preserving and transferring knowledge in the field of fast reactor technologies. Experience of the Obninsk Institute of Nuclear Power Engineering MEPhI

Obninsk Institute for Nuclear Power Engineering has been realizing education and training of specialists for nuclear industry since establishing in 1950s. Educational programs are developed for nuclear power plants as well as for research and development institutions of Russian Federation and abroad. Due to close connections with scientific Institute of Physics and Power Engineering named A.I. Leipunsky and other organizations of the Rosatom State Corporation a lot of professors and researchers in Obninsk branch of MEPhI university have experience in operation and research in fast nuclear reactors such as BN-350, BN-600, BN-800 and others.
The paper describes experience in teaching, preserving and transferring knowledge in fast reactor technologies for more than 60 years. Nowadays the specific of fast reactors can be taught in separate sections of courses (for example, fast reactor physics as a part of general Nuclear Reactor Physics course), separate courses (like Liquid metal coolants) and finally in specially tailored master’s program like Physics and technologies of fast reactors. The different ways of preserving and transferring knowledge are considered such as inviting famous specialists as professors, joint publications, providing laboratory practice and practical training (including preparing graduation thesis) for students on the base of Rosatom’s companies. The modern distant learning technologies (video courses, online lectures, using VR and AR technologies, etc.) are discussed as well.

Speaker: Aleksandr Nakhabov (Obninsk Institute for Nuclear Power Engineering NRNU MEPhI)
• 259
Investigation on Human Resources Needs and Competences Building for ALFRED Implementation in Romania

ALFRED is the demonstrator of Lead Fast Reactor (LFR) technology. According to strategic documents (at national level and of FALCON international consortium), it is planned to be built on Mioveni nuclear platform. An experimental infrastructure consisting of six experimental facilities (ATHENA, HELENA2, ELF, ChemLab, HandsON, Meltin’Pot) and a coordination Hub is planned to be built on the same site in support of the licensing process and technology development.
The development of the LFR technology faces various challenges including: (1) Research and Development (R&D) open issues (such as development and behaviour of the structural and cladding materials, the control of lead and cover gas chemistry, development of the instrumentation and control, fuel and fuel cycle, deterministic and probabilistic analyses, thermal-hydraulics for large pool configuration of molten lead, etc.) and (2) the novelty of the qualification, demonstration, validation and verification process for the ALFRED demonstrator. An appreciable number of high qualified personnel is estimated to perform the envisaged activities. In this context, the human resources including the competences building process are considered as crucial factors for the success of the implementation.
Considering the complexity of the scientific activities, the high degree of specialization and the existing offer of the workforce market, an education and training program is essential. Update/adaptation of the existing curricula in the education programme of the Romanian universities is needed, and can be achieved for example by new specializations/courses devoted to Generation IV systems (with a focus on LFR technology), a dedicated internship programme in European or worldwide experimental facilities, specialised/dedicated/thematic workshops, summer schools on simulation and experimental activities, etc.
This paper presents the outcomes of the investigation on human resources needs for ALFRED implementation in Romania, developed in the framework of Romanian PRO ALFRED project. The jobs estimation has been accounted for the ALFRED infrastructure R&D activities, operation and realization of the specific activities for each experimental facility and for the Hub, as well as for the safe operation of the ALFRED demonstrator.
Around 600 jobs have been identified for the operation of ALFRED demonstrator and its support infrastructure, as well as for the R&D related activities. For each identified job, the specializations and the minimal competencies have been established.
To see how the existing Romanian educational programs cover the minimal competences required by ALFRED infrastructure, an expert judgement evaluation has been performed in University of Pitesti and University Politehnica of Bucharest.

Speaker: Dr Minodora APOSTOL (RATEN ICN)
• 260
Overview of IAEA Fast Reactor Related Technology Development Activities

The International Atomic Energy Agency (IAEA) supplements and supports nuclear research and development with many efforts to improve and make nuclear data more accessible. Through technical community building research projects, and tool development, the Nuclear Power Technology Development section (NPTDS) provides many services and opportunities to amplify the research of member states’ experts. The Fast Reactor (FR) team is guided by the TWG-FR (Technical Working Group on Fast Reactors), the longest operating technical group at the IAEA, by identifying topics for Technical Meetings, providing data for CRPs (Coordinated Research Projects) and proposing other opportunities to further develop fast reactors technology internationally. This paper details several of the projects and work that has been recently completed and the projects currently ongoing.
One of the most visible products of NPTDS are the CRPs. Each project typically lasts for four years and aims to produce a high-quality publication detailing the work of all contributors. This method provides an avenue for information and dating sharing, as well as an opportunity to compare and contrast different simulation tools in use around the world. This paper discusses the CRP process and explain how to contribute to upcoming projects. The Fast Reactor team of NPTDS currently has two ongoing FR CRPs: Neutronics Benchmark of CEFR Start-Up Tests, and Benchmark Analysis of FFTF Loss of Flow Without Scram Test. Recently completed in 2020, Source Term Estimation for the Prototype Sodium Fast Reactor (PSFR) is currently preparing for publication.
In addition to leading and facilitating CRPs, the IAEA focuses on capacity building and reducing barriers to entry of nuclear power. Most recently, this has been accomplished through tool development and open source code community building. In 2020-2021, the IAEA conducted several studies on particular topics such as passive shutdown systems for fast reactors, benefits and challenges of small modular fast neutron systems, interaction of structural materials and liquid heavy metals. By leveraging the vast historical data preserved at the IAEA, NPTDS is engaging in building new resources for analysis. One new project is the Sodium Properties Calculator web application, which demonstrates the initiative for future module development as an additional output of CRPs. ONCORE, the open source code initiative aims to build support resources and communities around analysis tools.
This paper provides an overview of the projects and initiatives of the IAEA in the area of fast reactors technology.

Speaker: Joseph Mahanes (IAEA)
• Coffee Break
• Panel: Strengthening Fast Neutron Systems’ Community: Empowering the Next Generation’s Professionals, Towards Gender Balance, Cross-cutting Disciplines

Moderator: Mr. Joseph Mahanes

• Friday, April 22
• 1.3 System Innovations

Chairs: S. Raghupathy and Rami El-Emam

• 261
Sketch Design of Fuel Sub-Assemblies for a SFR-150 MWe

During the years 2018 and 2019 of the ASTRID program, a simulation program on SFRs has been prepared by the CEA and its industrial partners – EDF and Framatome – featuring sketch studies of a smaller-size SFR with extended experimental purposes. The power of the core has been reduced to 150 MWe to minimize investment costs while keeping the capacity to demonstrate the feasibility of Pu multi-recycling and to qualify designs and technologies expected for the future industrial SFRs.
These requirements led to an evolutive design for the core and the fuel sub-assemblies (S/A) over the reactor lifetime. At the beginning, the fuel pins will be similar to the one in former SuperPhénix SFR with UPuO2 fuel containing Pu from reprocessed PWR-UOX fuels. In the next step, Pu coming from reprocessed PWR-MOX fuels will be introduced. Then the concepts for the future industrial SFRs will be qualified: low void worth core, ASTRID-like large pin-small wire bundle, ODS cladding…
This paper presents sketch design studies of fuel S/A for a 150 MWe SFR at the end of 2019.
The hexagonal wrapper tube can host either a 169-SPX-type-pins bundle or 127-ASTRID-type-pins bundle. The thermomechanical behavior of the fuel bundle has been calculated with DOMAJEUR code. The lower gas plenum of the fuel pins has been reduced thanks simulations with GERMINAL fuel performance code, developed within the PLEIADES software environment, considering a nominal operation up to 87.5 dpa followed by an unprotected loss-of-flow transient. The upper neutron shielding is made of steel and B4C rings housed in a leaktight compartment to stay compatible with the washing process, while limiting the secondary sodium activation and the irradiation level of diversified absorber rods electromagnet. The overall S/A length of 4.20 m has been reduced by 30 cm compared to ASTRID-600 in the perspective of costs reduction.

Speaker: Mr Thierry BECK (CEA)
• 262
OPTIMIZATION OF BUILT-IN PRIMARY SODIUM PURIFICATION SYSTEM FOR ADVANCED BN REACTOR PLANT

While developing an advanced BN reactor plant the tasks were put to reduce a reactor plant cost with obligate meeting of safety requirements and reliability increase of reactor plant equipment and systems.
A purification system with cold traps (CT) placed in the reactor vessel (built-in purification system) was applied in the large BN reactor plant for primary sodium purification.
The developed CTs were equipped with unchangeable electromagnetic devices (a pump and a pump-throttle) to provide flowrate through the CT.
While developing an advanced BN, the primary sodium purification system was optimized - the cold trap design was optimized.
The electromagnetic pump and the electromagnetic pump-throttle were excluded from the cold trap design.
Sodium supply was arranged from the pressure chamber of the reactor.
To control sodium flowrate, the cold trap was equipped with changeable mechanical regulating devices with built-in flowmeters, and a changeable throttling device.
In addition, a CT cooling circuit was optimized within the framework of optimization. Eccentric collectors were used for cooling agent inlet and outlet; it permitted to increase CT impurity capacity by ~ 1.75 times and to increase purification system efficiency by ~ 1.5 times.
The performed optimization permitted to increase CT reliability and repairability, reduce quantity of CTs for replacement over the reactor plant service life by two times approximately; this made a positive effect on decrease of reactor plant construction cost and costs of CT replacement and disposal of spent CT.

Speaker: Mr Sergey Rukhlin (JSC “Afrikantov OKBM”)
• 263
DESIGN & ANALYSIS OF A NOVEL ARRANGEMENT FOR COUPLING AND DECOUPLING OF ROTATABLE PLUGS IN PFBR

Large and Small Rotatable Plugs (LRP & SRP) form part of top shield and are used to position transfer arm over any required subassembly location during fuel handling. The annular gap between the Roof Slab and LRP as well as between LRP and SRP is sealed with the help of two types of elastomeric seals - Primary inflatable seals and secondary back up seal. These seals have a design life of 10 years and need to be replaced after every 10 years. The procedure to be adopted for replacement of seals involves removal of the hexagonal socket head screws connecting top & middle ring to support ring of LRP/SRP, lowering of the LRP/SRP onto roof slab/LRP respectively and lifting of top & middle ring to gain access to the seals. Similar procedure is applicable for carrying out maintenance of large diameter bearings also. To remove top & middle ring, load on the screws connecting the top & middle ring to rotatable plugs (400t / 200 t for LRP & SRP, respectively) needs to be relieved and then the plug needs to be lowered onto the Roof Slab. To relieve the load and lower the plug, a novel concept of multiple tie rod & nut arrangement, with each one operating in a sequential way was conceived & implemented. This dedicated arrangement was designed taking into account the space constraints over the Top Shield and the eccentric nature of the loading of the plugs. In this paper, the design aspects of arrangement conceived for lowering / lifting of LRP/SRP are discussed along with the results of structural analysis carried out to confirm the design. The maximum allowable torque and nut turn during each operation in the tie rod is estimated and presented.

Speaker: Jagruti Mote (Reactor Design & Technology group, Indira Gandhi centre for Atomic Research)
• 264
Optimization of the secondary loops on the ESFR SMART project

The ESFR SMART project is a European sodium fast reactor project, which follows on from the EFR project, then the CP ESFR project, and whose goal is to present an improved project in terms of safety, taking into account the new rules to be applied in particular following the post Fukushima provisions. For the primary circuit a number of safety options have already been presented (ref 1, 2, 3), based on simplifications as well as passive and forgiving systems. In the same spirit, this paper proposes an optimization of the secondary circuits to improve their intrinsic safety and their compactness.
First, these circuits have the role of evacuating the power of the reactor, and in case of shutdown will actively participate in the evacuation of the residual power. The use of these circuits has been favored because it is the loop commonly used by the operator for this purpose, and in all operating circumstances. They were therefore pre-dimensioned (ref 4 , 5) to be able to evacuate this power alone even after the loss of the water circuits, only by natural convection of the air around the modules of steam generators. Moreover, in each secondary loop a dedicated system connected to the exchanger is capable of performing this function on its own, in passive natural convection and even in the event of draining of the secondary circuit.
Second, the REX of the SFRs shows that sodium leaks mainly take place at the level of the secondary circuits .Proposals are made for the secondary piping to increase the possibilities of rapid detection and mitigation . For this, we propose to use straight pipes where the expansions are taken up by bellows. This helps to minimize lengths and welds. This also allows the use of an offset thermal insulation allowing a quick and more reliable sodium leak detection. This point greatly improves the compactness of the circuit and therefore makes it possible to significantly reduce secondary buildings.
Then, for the sodium / water reactions, the choice of modular steam generator has been made to increase quickness of detection at the outlet of each module. It allows also minimizing consequences on the operation of the reactor able to operate with one unavailable module
Finally, some propositions are made on the use of passive thermal pump to assure passively a minimal sodium flow rate.
The conclusion summarizes the necessary R&D to allow this optimization.

Speaker: joel guidez (CEA)
• 265
Design of secondary sodium based decay heat removal system for future fast breeder reactors

Fast Breeder Reactor -1&2 (FBR-1&2) is a sodium cooled, pool type, Mixed Oxide (MOX) fuelled reactor with two sodium loops (primary and secondary). The design of this reactor is based on experience from Fast Breeder Test Reactor (FBTR) and prototype Fast Breeder Reactor (PFBR). Decay Heat Removal (DHR) system removes decay heat from the reactor after shutdown to ensure adequate cooling of core sub-assemblies. PFBR has two diverse paths for decay heat removal namely, Safety Grade Decay Heat Removal System (SGDHRS) and Operation Grade Decay Heat Removal System (OGDHRS).
OGDHR system requires at least one secondary loop, steam water circuits and off-site power supply for decay heat removal. SGDHR system is operated when OGDHR system is not available. In order to improve reliability of DHR system, it is planned to have an additional DHR system operating on secondary sodium, thus reducing the dependency on SGDHR system. The design of Secondary Sodium based Decay Heat Removal System (SSDHRS) for FBR-1&2 was carried out after, reviewing the design and operational experiences of BN 800, SUPERPHENIX and MONJU available in various forums.
SSDHRS is a part of Secondary Sodium Main Circuit (SSMC), it operates only during shutdown condition for decay heat removal. This system is designed for a heat removal capacity of 15MW. It is provided with an Air Heat Exchanger (AHX) with hot sodium flow in tube side by forced circulation using Secondary Sodium Pump (SSP) and air flow over the tubes by forced circulation using blower.
Heat removal capacity of the system with passive operational mode was also studied and found to be about 60% of the active capacity. System optimization was carried out to arrive at the sizing of various equipment of SSDHRS (Dimensions of AHX, blower capacity, height of stack and circuit design). Parametric studies have been carried out to analyse the effects of sodium temperature and flow rate on heat removal capacity of SSDHRS. SSDHR system is envisaged to cater to fuel handling and other maintenance conditions instead of relying on OGDHR system which requires external power supply, recirculation pumps, condenser cooling fans and steam generators to function.

Speaker: Mr Pritam Kumar Patel (Indira Gandhi Centre for Atomic Research)
• 266
Integration of Small Modular Lead Fast Reactor with Energy Storage for load-following operation in high V-RES penetration electricity markets

Energy decarbonisation, through the transition from fossil fuels to V-RES electricity production and the electrification of transport & heating sectors, may jeopardise the electricity supply security on the long term, because of the growing power demand and the increased production volatility. While advanced and modular reactor designs can make nuclear an attractive low-carbon solution to diversify the energy mix and address the power demand increase, a paradigmatic change is required in both NPP design and operation to increase load-following mode attractiveness. Indeed, although the current nuclear Gen-III/III+ fleet provide good load following capabilities and some operators (especially in high nuclear share markets as France) find profitable operating NPP in load-following mode, most of nuclear generating units are operated in baseload mode. This paper investigates the feasibility and the potential of integrating a cost-effective Energy Storage system into a Small Modular Lead Fast Reactor, to achieve load-following performances while maintaining the reactor at high power levels minimizing power excursions. Indeed, the integration of Energy Storages in Gen-IV reactors may significantly boost nuclear competitiveness in high V-RES penetration electricity markets, by combining the economic benefits of running nuclear reactors at high power (i.e., efficient use of capital invested in plants, simplicity and reliability of the operations) with the plant load-following capacity, compensating V-RES volatility. The paper investigates the Energy Storage option under a wide and comprehensive perspective, from the description of the reference electricity market with the identification of specific national grid requirements down to the Energy Storage technology selection, integration with the balance-of-plant and preliminary sizing, to best-fit the load-following demand and LFR specificities. Romania has been selected as reference scenario for the investigation, due to the representative energy-mix with high RES penetration (42%), a large use of hydropower (27%) to compensate for wind and solar volatility as well as the consolidated use of nuclear power (18%) as baseload.

Speaker: Mr Marco Caramello (Ansaldo Nucleare)
• 267
POWER CONTROL OF THE FAST NUCLEAR-BURNING-WAVE REACTOR

One of the most important problems in the further development of nuclear energy, from the point of view of its public acceptance, is the problem of safety. Thus, the development of new concepts for nuclear fission reactors with so-called "intrinsic safety" is a very urgent task. An equally important problem for the sustainable development of nuclear power is the need to expand the fuel base by involving uranium-238 and thorium-232.
The concept of a fast reactor (FR) operating in a self-sustaining nuclear burning wave (NBW) mode, proposed in [1], also known as the Traveling Wave Reactor and the CANDLE, if implemented, is capable to solve both of these problems, and in a very effective way. The "intrinsic safety" of the reactor is based on the specific mechanism of negative reactivity feedback inherent in the NBW mode, which ensures automatic maintenance of the critical state of the reactor even under external influences [2]. The use of depleted uranium and thorium as the main fuel with high burnup [3] makes it possible to use a “one-through” scheme instead of an expensive closed fuel cycle.
In our work, the possibility of controlling the NBW reactor power by changing the efficiency of a neutron reflector is investigated. Such a possibility is an important in the context of widespread use of weather-dependent wind and solar energy. The consideration was carried out on the basis of the approach developed in [4, 5], using the numerical solution of the multigroup nonstationary neutron diffusion equation together with the system of equations for fuel burnup and nuclear kinetics of delayed neutron precursors. A cylindrical multi-zone FR with U-Pu cycle fuel, in which NBW propagates in the axial direction, is considered. The calculations took into account the presence of a structural material Fe and a Pb-Bi coolant in the reactor core. The reflector consisted of 90% Pb-Bi and 10% Fe. Optimal algorithms are proposed for bringing the NBW reactor to a given power (both with decreasing and increasing) using a proportional-differential method for controlling the tantalum-181 content in a radial reflector.

1. Feoktistov L.P, Sov. Phys. Doklady, 34 (1989) 1071.
2. Fomin S.P., et al., Int. conf. "FR-13", Paris, 2013, paper CN-199-457.
3. Fomin S.P., et al., Prog. Nucl. Energy, 50 (2008) 163; 53 (2011) 800.
4. Fomin S.P., et al., Int. conf. "Global 2009", Paris, 2009, paper 9456.
5. Malovytsia M.S., et al., Ann. Nucl. Energy 148 (2020) 107699.
Speaker: Dr Sergii Fomin (National Science Center "Kharkov Institute of Physics and Technology")
• 268
Evaluation of an increase of the power density for the French commercial Sodium Fast Reactor and optimization study at 1100 MWe with the SDDS tool

In order to enhance the competitiveness and to reduce the construction cost of the future industrial Sodium Fast Reactors (SFR), several options are explored which need further R&D studies or design assessment. Among them, the possibility to reduce the size of the reactor vessel has been investigated through the reduction of the core diameter and the increase of the power density thanks to several optimisation studies conducted by the R&D department of EDF.

To this end, an in-house multi-physics optimization tool called SDDS has been used. The basis of the method is to predict the performance of a large number of core designs using surrogate models. The surrogate models are themselves created using the results of a parametric calculation scheme based on the following codes: ERANOS for the neutronic, MAT5DYN for the thermal-hydraulic transients and GERMINAL for the thermomechanical fuel performance.

A first optimization study has been performed in 2018 to define a compact core design for the 1000 MWe French commercial SFR. As a result, two designs were selected as they offered a good compromise between the safety criteria and a reduced core diameter: a twelve Sub-Assembly (SA) rings core with a smaller core diameter and a thirteen SA rings core with better safety margins.

This paper focuses on a second optimisation study which has been performed more recently on a 1100 MWe power reactor in order to evaluate the impact of an elevation of 10% of the nominal power on the results. The analysis of the results shows that the main trends (e.g. large pellet, large fertile plate height, etc.) are the same as the ones observed for the optimum designs in the 2018 study. However, the designs selected in 2018 do not meet some of the safety criteria anymore after the increase of their power density. Thus, a new core design with 13 SA rings has been proposed with a better compromise between safety performances and core diameter to operate at 1100 MWe.

Speaker: Dr Delphine Gérardin (EDF)
• 269
iMAGINE - a Breakthrough Technology for Closing the Fuel Cycle without Reprocessing

The energy trilemma and UN-SDG 7 are drivers for energy research to support the UK governments net-zero emissions law. Nuclear reactors are a highly attractive candidate for reliable, 24/7 available, low-carbon electricity generation. However, current technology reactors and their related fuel cycles suffer from unreasonably high cost, a lack of sustainability, and a waste problem due to the absence of recycling. Molten salt technologies will be a key step into the future of nuclear supporting a disruptive way of optimizing the whole nuclear system to enhance sustainability and affordability. Main advantage, compared to existing technologies, is elimination of complex solid fuel production and fuel cycle technologies. Molten salt systems will be a breakthrough for most efficient fuel use, by operating on existing spent fuel while drastically reducing the cost of nuclear and solving the long-term waste problem. However, developing a disruptive, highly sustainable and affordable fuel cycle – instead of just a reactor – requires a strong inter-disciplinary approach, linking physics, engineering, and chemistry.
Primary key is to deliver the essential step into any new reactor technology: a zero-power facility to research the game-changing technology in safe settings, to advance knowledge and capabilities in the technology to grow the skills base in the UK. Core activity is improving simulation and demonstration of innovative control and safety features to allow a qualified response to regulatory requests and to support the formation of the required skilled workforce to support BEIS, aiming to achieve an industrially demonstrated, market ready product in 2050.
The proposal pushes the breakthrough technology delivering significantly improved sustainability indices, characterized by:
• avoiding mining (major source of eco-toxicity, carbon emissions, and cost) & avoiding enrichment (major energy consumption, proliferation-risk, and cost)
• reducing waste production and storage demand by the reuse of existing spent fuel & eliminating highly-radiotoxic transuranium isotopes (reducing the final disposal challenge)
• eliminating reprocessing (proliferation risk, prohibitively expensive prior step for closed fuel cycle) & solid fuel production (major cost driver and radiation source in closed fuel cycle)
• replacing reprocessing with demand driven salt clean-up & applying low pressure technology
The ultimate aim is to prepare the UK for a net-zero future using highly-innovative technologies. The impact of the proposed technology and the attractiveness of the vision is evidenced by the rapid take-up through the major industrial technology developers, including Terrestrial Energy, Terrapower, Elysium Industries….
An overview on the research plan will be given.

Speaker: Prof. Bruno Merk (University of Liverpool)
• 2.4 Severe Accidents

Chairs: Frederic Serre and Shigenobu Kubo

• 270
Development of methodology to evaluate mechanical consequences of vapor expansion in SFR severe accident transients: lessons learned from previous France-Japan collaboration and future objectives and milestones

In the frame of France-Japan collaboration, one of the objectives is to define and assess the calculation methodologies, and to investigate the phenomenology and the consequences of severe accident scenarios in sodium fast reactors (SFRs). A methodology whose purpose is to assess the loadings of the structures induced by a Fuel Coolant Interaction (FCI) taking place in the sodium plenum of SFR has been defined in the frame of the collaboration between France and Japan during 2014-2019. The work progress will be spread over the period 2020-2024 and the main objectives and milestones will be introduced in this paper. The objective of studies is to comprehensively address the margin between the limit of integrity of the main vessel structures and the loadings resulting from severe accidents.
For this purpose, the SIMMER mechanistic calculation code simulates core disruptive accident sequences in SFRs. However, SIMMER cannot be used for main vessel loading assessment while it does not take into account fluid structure interactions. That is the reason why, associated with SIMMER code, a fluidstructure dynamics tool evaluates this interaction i.e. EUROPLEXUS is used in CEA studies and AUTODYN tool is used in JAEA studies. In this paper, a benchmark study is described in order to illustrate the evaluation of vapour expansion phase in the hot plenum. To do that, joint input data are used on the basis of an ASTRID 1500 MWth core degraded state after the power excursion which leads to vapour expansion. The most penalizing case was evidenced in this study by suppressing the action of transfer tube in-core mitigation devices in SIMMER input deck and thus privileging the upward molten core ejection. Since the risk of main vessel failure by cumulative stresses is an issue often discussed in the SFR concepts, the calculation methodology presented in this paper based on chaining of SIMMER code with another Fluid/structure evaluation tool is very promising. The future perspectives are highlighted.

Speaker: Andrea Bachrata (CEA)
• 271
Simulation of ULOF initiation phase in ESFR-SMART with SIMMER-III

A large 3600 MWth European Sodium Fast Reactor (ESFR) design was proposed in 2000s. It is studied now in the EURATOM ESFR-SMART project. A new core configuration with several new safety measures, including a reduced to a near-zero value coolant void reactivity effect, mainly due to introduction of a sodium plenum above the core and core flattening, has been established recently. We investigate the efficiency of these measures by simulating transients such as unprotected loss of coolant flow (ULOF) with the SIMMER-III code, starting from nominal conditions till molten fuel discharge from the core. In the initiation transient phase, before structure and fuel melting, sodium boiling happens in the described simulations. The reactivity oscillates after the boiling onset due to subsequent boiling and flooding in the upper fissile core part and the sodium plenum above, where the void effect is negative. These oscillations are associated with interaction of different flow channels. In the paper we investigate these phenomena by considering different modelling options. New SIMMER capabilities for taking in account core thermal and control rode driveline expansion reactivity effects are also presented and their influence on the transient behavior is discussed. We also compare to results of ULOF simulations performed with other codes for ESFR-SMART.

Speaker: Xue-Nong Chen (Karlsruhe Institute of Technology (KIT), Institute for Nuclear and Energy Technologies (IKET))
• 272
The SAIGA in-pile experimental program to qualify the SIMMER calculation tool in SFR Severe Accident Conditions

The CEA, together with the NNC, has carried out a feasibility study with regard to conducting an in-pile test program - the future SAIGA program (Severe Accident In-pile experiments for Gen-IV reactors and the Astrid prototype) - on the degradation of a SFR fuel bundle with molten fuel discharge device which are planned to be housed in the IGR reactor (Impulse Graphite Reactor operated by NNC-RK). The purpose of the SAIGA program is to qualify the SIMMER computer code on the SEASON platform based on one in-pile test conducted with two fuel sub-assemblies and one discharge tube representative for an in-core mitigation device dedicated to severe accident situations. This test should be representative, as much as possible, for the phenomena encountered during Unprotected Loss-Of-Flow Total (ULOF) severe accident sequences.
A sodium loop will be built and connected to the in-pile experimental device to drive the sodium flow inside the fuel pin bundle with experimental conditions close to the SFR nominal conditions before triggering of the ULOF sequence. Over accidental transient period with sodium flow reduction, the constant neutron heating from IGR reactor will lead to degrade a first sub-assembly (16 Kazakh fuel pins by sub-assembly) to produce some molten fuel material and the propagation of this degraded fuel will be followed by fine instrumentation towards both a second sub-assembly and a discharge tube.
For this scenario, the feasibility study defined the main characteristics of the experimental device and the operating conditions for the test to be conducted in the IGR reactor. The purpose of this study was to assess the capacity of the IGR reactor to provide the necessary neutron flux during the transient, to demonstrate the capacity to carry out on-line or post-test measurements of the variables of interest, and to assess the schedule of one test incorporating the safety file. Also, the sodium loop feeding the test device and its instrumentation were studied and their feasibility demonstrated.

Speaker: Frédéric Payot (CEA : Commissariat à l’Energie Atomique et aux Energies Alternatives)
• 273
A Status of Experimental Program to Achieve In-Vessel Retention during Core Disruptive Accidents of Sodium-Cooled Fast Reactors

To achieve in-vessel retention for mitigating the consequences of core disruptive accidents (CDAs) of sodium-cooled fast reactors, controlled material relocation (CMR) has been proposed as an effective safety concept. CMR is not only aiming at eliminating the potential for exceeding prompt criticality events that affect the integrity of the reactor vessel, but also enhancing the potential for the in-vessel cooling of degraded core materials during CDAs. Based on this concept several design measures have been studied, and, to evaluate their effectiveness, experimental evidences to show relocation of molten-core material were required. With this background, a series of experimental program called EAGLE (Experimental Acquisition of Generalized Logic to Eliminate re-criticalities) has been carried out collaboratively over 20 years between Japan Atomic Energy Agency and National Nuclear Center of the Republic of Kazakhstan (NNC/RK) using an out-of-pile and in-pile test facilities of NNC/RK. The EAGLE program is divided into three phases, they are called EAGLE-1, EAGLE-2 and EAGLE-3, to cover whole phase after core-melting begins. The subject for EAGLE-1 and the first half of EAGLE-2 is CMR in the early phase of CDA in which the core melting progresses rapidly driven by positive reactivity insertions. The subject for the later half of EAGLE-2 and whole EAGLE-3 is CMR in the later phase of CDA in which the gradual core melting by decay heat and relocation and cooling of degraded core materials occur. In this paper, the major achievement of the EAGLE program and future plans are presented.

Speaker: Dr KENJI KAMIYAMA (Japan Atomic Energy Agency)
• 274
Comparisons of Feedback under UTOPA with In Pin Fuel Motion Dynamics in Fast Reactors

Safety studies of fast reactors are carried out on a medium sized core and found that, under Unprotected Transient Over Power Accidents (UTOPA) there is fuel melting and there is feedback due to in pin fuel motion. From the UTOPA analyses it is found that, the in pin fuel motion feedback reduces the peak reactor power and hence it reduces the hot spot clad and coolant temperatures at the end of the transients. Under such transients, when there is change in reactor power, for the given primary & secondary systems, the balance of plant gets affected. This results in change in inlet coolant temperature, which affects the temperature profile and has a considerable contribution in shaping the power profile and its respective feedbacks. To study the effect of UTOPA with different boundary conditions, analyses has been carried out with constant inlet coolant temperature (CICT), Time Dependent Inlet Coolant Temperature (TDICT) profile, with and without In Pin Fuel Motion (IPFM) feedback, etc. Comparisons of those results give a better understanding of their respective feedbacks & calculation methodology with respect to different boundary conditions. Considered UTOPA studies with different boundary conditions are,
1. The uncontrolled withdrawal of control rod with CICT, without IPFM. It is based on the assumption that, the molten fuel is hypothetically stationary (available in the same location) even after melting.

1. The uncontrolled withdrawal of control rod with TDICT, without IPFM.

2. Uncontrolled withdrawal of control rod with CICT and IPFM after the initiation of fuel melting.

3. Uncontrolled withdrawal of control rod with TDICT and IPFM feedback after the initiation of fuel melting.

From the comparison of results, it is learnt that the IPFM is significant in bringing the reactor to a safe state under UTOPA. The involved conservatism of the result is observed in the analyses where the IPFM feedback is not considered as compared to the best estimate analyses of considering IPFM feedback. From the study, it is concluded that if a reactor is found to be safe without IPFM feedback, it can actually converged to a safe state with IPFM feedback. Thus the results without IPFM feedback are conservative.

Speaker: Mrs Sathiyasheela T. (IGCAR, India)
• 275
French-Japanese experimental collaboration on fuel-coolant interactions in sodium-cooled fast reactors

Fuel-coolant interactions (FCI) are important phenomena occurring in the event of core disruptive accidents in sodium-cooled fast reactors (SFR). A new phase of experimental research into FCIs has been launched through a collaboration between French and Japanese organisations. Bespoke high-speed, high-resolution X-ray imaging at the JAEA MELT facility has enabled successful visualisation, in the highest resolution achieved to date, of the quenching of a jet of molten steel in a pool of molten sodium. The FCI experimental program is accompanied by calibration tests undertaken using “phantom” models, which are necessary for detecting imaging artefacts, such as vignetting and optical distortion, and assisting in the development of algorithms to reconstruct melt fragments in 3D from X-ray attenuation data. The CEA have developed SPECTRA (Software for Phase Extraction and Corium TRacking) for processing the images acquired at the MELT facility, enabling the segmentation of melt and vapour phases from the raw images and the tracking of discrete melt fragments traversing the imaging window. The first experiment at the MELT facility under this collaboration revealed evidence of extensive crust formation at the interface between the melt and sodium. Rapid vaporization of entrained sodium appeared to lead to fracturing of the crust during a thermal fragmentation event, resulting in a debris population containing both fine fragments and large frozen jet shells. JAEA have now commissioned a new MELT test section, roughly 10-times greater in sodium capacity, to observe the interaction at increased scale. In parallel, the SERUA (Sodium boiling Experimental Ring for Understanding of fuel-coolant interAction) facility is currently being developed by the CEA to investigate film-boiling heat transfer at the interface between corium droplets and molten sodium, in support of computational research into FCIs in sodium.

Speaker: Christophe Journeau (CEA)
• 276
Transient 3D simulations for the ASTRID reactor: preliminary results for the ULOF initiation phase

An Unprotected Loss Of Flow transient (ULOF) in the 1500MWth Advanced Sodium Technological Reactor for Industrial Demonstration (ASTRID) reactor is investigated with SIMMER-IV, a 3D multi-phase, multi-velocity and multi-component fluid-dynamics and neutronics code. The 2D RZ code version, SIMMER-III, is a working horse for fast reactor severe accident studies at KIT and other institutions, but the 2D approach affects simulation of reactivity feedbacks and of behavior of reactor materials under accident conditions. On the other hand, 3D SIMMER ULOF calculations take a lot of time and computer memory and were not tried for a full-vessel pool-type fast reactor model at KIT and EdF before recently.

Recent developments for SIMMER-IV, including introductions of a new neutron transport solver based on the PARTISN code and of a new procedure for generation of few-group cross-sections during the transient, offer new simulation capabilities. Also more computer power is available now. Therefore, an effort was done at KIT, in collaboration with EdF, to perform calculations for a full-vessel 3D model of the primary ASTRID circuit. After a first attempt, modifications were introduced in the code and the employed reactor model for improving their performance.

In the paper we inform on our experience with 3D SIMMER calculations, present preliminary results of 3D steady state and transient simulations for the ULOF initiation phase, do preliminary comparisons with results of 2D analyses.

Speaker: Simone GIANFELICI (ENEA)
• 277
Experiment and Numerical Simulations on SFR Core-catcher Safety Analysis after Relocation of Corium

An In-vessel core catcher located in European type SFR is a safety design feature to guarantee the integrity of the SFR during a core-melting accident. The core catcher collects and distributes the relocated melt from the core region via discharging tubes to avoid firstly a significant molten pool in the core, and secondly the local thermal attack on bottom part of the vessel. The heat transfer and ablation behavior of core-catcher material under the thermal load of the core melt at high-temperature and with high decay heat will be studied experimentally with a new facility in Karlsruhe Institute of Technology and numerically in CEA within the ESFR-SMART European project.
The new facility, named ESFR-LIVE, is a 3-dimensional model of Core-catcher in a length scale of 1:6. The lower part of the tray-type core-catcher is a truncated cone and the upper part is a cylinder with 1 m diameter. The cooling of liquid Na at all boundaries is simulated by a water cooling channel enclosing the test vessel, and a cooling lid at the upper surface of the simulate. Four planes of individually controllable heater in the vessel enable the variation of the height of the core melt, and the shapes of the melt pool. The distribution of bulk temperature, boundary temperature, wall temperatures and heat flux can be measured or determined. The simulant of core-melt is the eutectic NaNO3-KNO3 mixture, which is representative for the character of the general liquid oxide melt. Similarity comparisons on geometry, material properties and heat transfer features are presented in this paper. With the similar pool height, the EFSR-LIVE facility can well capture the dimensionless heat transfer features, e.g. Ra and Nu.
Computational Fluid Dynamics (CFD) pre-calculations of the LIVE-ESFR experiments will be performed on the basis of the ESFR-LIVE test conditions. These simulations will be performed with the TrioCFD code using High Performance Computing on ten million of mesh nodes. A parametric study according to the heating power distributions will be presented.
Regarding the turbulent flows with large Rayleigh numbers, these simulations require to use Large Eddy Simulation (LES) models. This study is supplemented by a mesh resolution analysis and a validation of TrioCFD on natural convection turbulent flows using past molten-pool experiments (BALI) and academic results.

Speaker: Ms Xiaoyang Gaus-Liu (Karlsruhe Insitute of Technology)
• 6.5 Integrated Analysis and Digitalization

Chairs: Yougi Zheng and Andrey Fedorovskiy

• 278
DIGITAL TECHNOLOGIES FOR PROJECT DEVELOPMENT ODEC AND PEC AND DIGITAL TWINS

In the conditions of the modern international market, not only the safety of nuclear facilities, but also their economic performance is critically important. When developing its facilities, the project direction "Proryv" tries to solve these problems comprehensively on the basis of new reactor technologies and closing the nuclear fuel cycle. Given the high degree of novelty of projects, the following key difficulties arise:
• the need for a large number of participating organizations to participate in the project, which use in their work heterogeneous information, calculation and modeling tools that require integration;
• significant uncertainty with the way of achieving the final results, a large amount of R&D performed, the results of which constantly cause changes in the projects of objects.
In addition to the standard set of modern CAD and engineering software used in the industry for the development of NPP projects, a comprehensive digital solution has been developed and applied to ensure that the specified economic indicators are achieved in compliance with all safety requirements for the development of ODEC and PEC projects. This solution includes:
• unified information space - a set of databases, data transmission channels, hardware and software and methodologies that ensure the joint work of project participants, common information services for private projects and integration of IT systems of participating organizations;
• information models of objects-a continuously updated structured set of electronic data and documents about objects and technologies of project direction, necessary and sufficient at each stage of the life cycle;
• integrating projects and consolidated 3D models of objects that provide visual navigation of objects, link documentation with the requirements management system of projects, as well as 4D models for handling equipment of ODEC and PEC objects;
• calculation complexes based on integrated mathematical models that allow for advanced construction and simulation modeling of objects in various modes of operation – normal operation, violations of normal operation and emergency, which is necessary for the development and testing of automated control systems, search and elimination of collisions on technological parameters.
In fact, integrating projects with calculation complexes based on integral computational mathematical models are digital twins of ODEK and PEC objects, accompanying real objects at all stages of the life cycle.

Speaker: Andrei Fedorovskii
• 279
Development of an Artificial Neural Network for predicting spatial interdependencies of reactivity effects in Sodium Fast Reactors

Artificial Neural Networks (ANN) are presented as a very powerful tool for modelling complex systems. This approach is becoming increasingly widespread and it has a great potential for nuclear reactor applications. In this work, an ANN is developed for predicting sodium void effects in a large Sodium Fast Reactor core and their spatial interrelations.
The ultimate goal is to provide more realistic inputs to the thermal-hydraulics code TRACE for point-kinetics-based transient analysis of the most recent ESFR core conception. With that goal, an ANN is developed and trained to provide the global sodium density effect and Doppler effect, receiving as input the normalized sodium density and temperatures at the different regions of the core. Local reactivity effects are computed using ERANOS deterministic code for an extensive set of combined scenarios in order to train the ANN.
In this work, the main aspects regarding the optimization of the ANN are presented. A neuron trimming exercise is carried out for getting the most consistent architecture. The developed model can predict the reactivity evolution taking into account the mutual interdependencies of sodium void and Doppler effect. ANN’s performance is analyzed by comparing its output in a real transient simulated by TRACE with traditional approaches.

Speaker: Mr Antonio Jiménez-Carrascosa1 (Politécnica de Madrid (UPM))