International Conference on Fast Reactors and Related Fuel Cycles: Next Generation Nuclear Systems for Sustainable Development (FR17)

Jun 26, 2017, 8:00 AM → Jun 29, 2017, 5:00 PM Asia/Yekaterinburg

Yekaterinburg

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” and was attended by some 700 experts from 27 countries and 4 international organizations representing different fields of fast reactor and related fuel cycle technologies. The International Atomic Energy Agency (IAEA) now proposes, almost four years later, to bring the fast reactor and related fuel cycle community together again. The Russian Federation’s State Atomic Energy Corporation “Rosatom” has proposed to host the conference in Yekaterinburg, Russian Federation. One of the main reasons for this proposed venue is that the sodium cooled fast reactor BN-800 was connected to the grid in December 2015 at the Beloyarsk nuclear power plant (NPP), which is located in the vicinity of Yekaterinburg. BN-800 is a successor of the BN-600 reactor that has been in operation at the Beloyarsk NPP since 1980.

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.

Guidelines FR17-PaperReviewGuidelines.pdf

Monday, June 26

Opening Session

Conveners: Dr Vladimir Kriventsev (IAEA), Mr Vyacheslav Pershukov (ROSATOM)

1 Opening Address by Director General, ROSATOM (by video message)

Speaker: Mr Alexey Likhachev (ROSATOM)

2 Opening Address by Director General, IAEA (by video message)

Speaker: Mr Yukiya Amano (IAEA)

3 Welcome Note by Conference General Chair

Speaker: Mr Vyacheslav Pershukov (ROSATOM)

4 Welcome Note by Conference General Co-Chair

Speaker: Mr Mikhail Chudakov (IAEA)

5 Welcome Note by Deputy Presidential Envoy in the Ural Federation District

Speaker: Mr Alexander Moiseev

6 Fast Reactor Development and International Cooperation (by Honorary General Chair)

Speaker: Mr Subhash Chander CHETAL (India)

9:30 AM Coffee Break

Plenary Session 26 June

Conveners: Dr Amparo Gonzalez-Espartero (IAEA staff), Mr Mikhail Chudakov (IAEA)

7 Closed fuel cycle technologies based on fast reactors as the corner stone for sustainable development of nuclear power

This article analyzes problems and approaches to modern nuclear power development using closed nuclear fuel cycle and fast reactors. It describes specified technical requirements for nuclear power systems in large-scale nuclear power industry. Targets and scientific problems solved by Rosatom’s “PRORYV” Project which is a part of the Federal State Program “Nuclear Power Technologies of New Generation in the Period of 2010-2015 and up to 2020” are examined.

Speaker: Mr Evgeny Adamov (Institution “ITC “PRORYV” Project)

Paper IAEA-CN-245-342(Adamov).docx

8 Research, development and deployment of fast reactors and related fuel cycle in China

to be submitted

Speaker: Mr Donghui ZHANG (China)

9 Status of the French Fast Reactor Programme

to be submitted

Speaker: Mr Sylvestre Pivet (France)

11:30 AM Lunch Break

1.1 SFR DESIGN & DEVELOPMENT - 1

Conveners: Mr Aleksandr Staroverov (JSC “Afrikantov OKBM”), Mr joel guidez (CEA)

10 Advanced Design Features of MOX Fuelled Future Indian SFRs

India has been operating a Fast Breeder Test Reactor (FBTR) successfully since 1985. Currently, a 500 MWe MOX fuelled pool type Sodium cooled Fast Reactor called Prototype Fast Breeder Reactor (PFBR) is under advanced stage of commissioning. The design, R&D, safety review, construction and commissioning experience from PFBR has motivated the commercial exploitation of MOX fuelled Sodium cooled Fast Reactors (SFR) with closed fuel cycle. Accordingly, six FBRs are planned in which, the first two units (FBR 1&2) will be located at Kalpakkam. These reactors are incorporated with advanced design features towards improved economy and enhanced safety. FBR 1&2 will be of MOX fuelled to be deployed ahead of metal fuelled reactors in order to capitalize on the experience gained in all the domains of SFR technology and to sustain the program. These future reactors need to have improved economy, enhanced safety and possible higher performance parameters. Economy is achieved by design optimization, reduction of material quantities, adoption of twin unit concept with sharing of facilities, design enabling integrated manufacture and erection leading to reduced construction time. Based on detailed studies, reactor power is enhanced with a slightly larger core and by way of design optimization and exploiting the improved manufacturing technologies, the sizes of major large size components are kept close to the industrial capacity that have been built in the country. This approach has led to raising of reactor power to 600 MWe leading to economic gains. With regard to safety, the important aspects taken into consideration are the internationally evolving Gen-IV safety criteria especially after Fukushima. The enhanced safety level seek to prevent severe core damage and large radioactivity release to the public and practical elimination of severe accident scenarios involving energy release and public evacuation. The major safety enhancements envisaged are (i) improved core inherent safety characteristics with sodium void coefficient less than 1 $, (ii) passive shutdown features and additional shutdown systems employing alternative working principles to prevent events leading to accident situations and (iii) passive & augmented decay heat removal capacity. This paper presents the advanced design features envisaged, towards enhancing safety and improving economy in the future MOX fuelled Indian SFRs.

Speaker: Mr Puthiyavinayagam Pillai (Indira Gandhi Centre for Atomic Research)

Material Changes made are shown in red color font

11 Overview of U.S. Fast Reactor Technology R&D Program

This paper provides an overview of fast reactor research and development efforts in the United States. Fast reactors are envisioned for a wide variety of actinide management strategies ranging from actinide destruction in closed fuel cycles to enhanced uranium utilization. With successful technology development, fast reactors are also intended for electricity and heat production, as being pursued through the Generation-IV International Forum collaborations. Several new initiatives for industry-led R&D, advanced reactor licensing framework, and discussions on advanced test/demonstration reactors are indicative of rising national interest in advanced nuclear technologies. Because capital investment in reactors is the dominant cost of any nuclear fuel cycle, R&D efforts to improve fast reactor performance are the primary focus. A variety of innovative features that hold the promise for significant cost reduction are being pursued; the diverse R&D activities are funded by several Programs in the DOE nuclear energy portfolio. Innovative technology options that may yield significant cost reduction benefits have been identified through concept development studies: high strength structural materials, a supercritical CO2 Brayton energy conversion cycle, advanced modeling and simulation tools, and in-service inspection techniques. In addition, technology development efforts for safety and licensing, and improved transmutation fuels are ongoing. For each technical area, recent accomplishments and key facilities will be identified to provide an indication of current status.

Speaker: Ms Heather Bell (U.S. Department of Energy)

12 Development of the new generation power unit with the BN-1200 reactor

One of the most important stage of works for the BN-1200 power unit project, which are implemented since 2007 according to Rosenergoatom Concern JSC program and Target Federal program “Nuclear energy technologies of the new generation for 2010-2015 and for perspective to 2020”, became the development of technical designs of the RP, turbine plant and materials of the power unit project in 2014. Main requirements at the technical design of the RP project defined development of the technical solutions in comparison with the solutions implemented in the previous projects, and ensured not only complete integration of the primary circuit in the reactor pressure vessel, but also significantly reduced number of systems, equipment, valves and pipelines, and optimized architectural solutions of the main and auxiliary buildings and constructions of the power unit, and optimized general layout of the site. These improved the main technical and economical indicators of the BN-1200 power unit and ensured their comparability with VVER RP not only in the field of safety, but also in the field of specific capital costs and LCOE. Further development of the project was defined with the design research of systems and equipment in the second half of 2015 and 2016, which indicated the following main directions of design work: increase of the power of the unit without change of the equipment design; change of design and layout solutions of the primary circulating pumps, emergency heat removal system, cold absorption trap filter of the secondary circuit, refueling box, and the secondary circuit. Implementation of the proposed technical solutions defines further optimization of the architectural solutions for the power unit and improvement of the technical and economical indicators without reduction of the safety level.

Speaker: Mr Sergey Shepelev (JSC “Afrikantov OKBM”)

Material IAEA-CN-245-402.pdf

Slides IAEA-CN-245-402.pptx

13 Advanced sodium-cooled fast reactor development regarding GIF safety design criteria

Design studies on a next generation sodium-cooled fast reactor (SFR) considering the safety design criteria (SDC) developed in the generation IV international forum (GIF) was summarized. To meet SDC including the lessons learned from the TEPCO’s Fukushima Dai-ichi nuclear power plants accident, the heat removal function was enhanced to avoid loss of the function even if any internal events exceeding design basis or severe external event happen. Several design options have been investigated and auxiliary core cooling system using air as ultimate heat sink has been selected as an additional cooling system regarding system reliability and diversification. Even though the next generation SFR already adopts seismic isolation system, main component designs have been improved considering revised earthquake conditions. For other external events, design measures for various external events are taken into account. Reactor building design has been improved and important safety components are diversified and located separately improving independency. Those design studies and evaluations on the next generation sodium-cooled reactor have contributed to the development of safety design guidelines (GIF) which is under discussion in the GIF framework.

Speaker: Mr Hiroki Hayafune (JAEA)

Material FR17_SFR_design_R2_clean.pdf

14 Current status of GIF collaborations on sodium-cooled fast reactor system

The SFR system arrangement Phase II became effective on 16 February 2016 by signatures of CEA, JAEA, KAERI, USDOE, and Rosatom), and was extended for additional 10 years. China signed the SFR SA Phase II on 3th August 2016 and Euratom is expected to sign near future. Collaboration of GIF SFR is growing adding new reactor concepts and related R&Ds. In 2015, a project arrangement on SFR System Integration and Arrangement (SI\&A) has been signed by 7 members : China, EU, France, Japan, Korea, Russia and US. In the SI\&A project, R\&D needs from the SFR design will be shown to the R&D project, and R&D results from each R&D project will be integrated into the designs. Presently there are four SFR design concepts as shown in ATFR-2015, 1) JAEA Sodium Fast Reactor (JSFR, loop) Design Track, 2) KALIMER-600 (KAERI, pool) Design Track, 3) European Sodium Fast Reactor (ESFR, pool) Design Track, 4) AFR-100 (DOE, modular) Design Track, are proposed from each signatory. China is going to propose CFR-1200, and Russia is going to propose BN-1200 as new design tracks. After SI\&A project started, the GIF-SFR has completed the function of R\&Ds and integration of R&D result to the SFR design. These strong collaboration network is expected to provide the promising generation IV SFR concepts. This paper describes SFR design concepts in SFR project and interactions with R\&D projects under GIF framework.

Speaker: Mr Hiroki Hayafune (JAEA)

Material FR17_GIF_SFR_R3_clean.pdf

15 Feasibility of Burning Wave Fast Reactor Concept with Rotational Fuel Shuffling

Burning wave fast reactor is very attractive concept. It is possible to achieve very high burnup using natural uranium or depleted uranium as fuel. It does not need fuel reprocessing facility. This kind of reactors can be categorized into two groups. One is the reactor whose burning wave is moving for radial direction, for example, Traveling Wave Reactor. The other is that whose burning wave is moving for axial direction, for example CANDLE burning reactor. The advantage of the reactor concept with radial direction wave movement is that fuel shuffling is easy, but high burnup fuels will exist in high neutron importance region at the center of core. It can be a disadvantage from the view point of neutron economy. The advantage and disadvantage of reactor concept with axial direction movement wave is vice versa. One of the ideas to solve the problems is the concept of shuffling of fuel pins or fuel elements rotationally so that high burnup fuel can be located in low neutron importance region. In the concept, fuel shuffling, load and reload are easy. The purpose of study is to show the possibility to apply for the concept of rotational fuel shuffling in burning wave fast reactor concept. Preliminary analysis was performed using continuous energy Monte Carlo code MVP2.0, JENDL-4.0 nuclear data library, and newly developed additional programs to simulate movement of fuel elements by the shuffling. The results of preliminary analysis showed the stable neutron flux profile can be existed in the core if the shuffling procedure is proper. The detail of shuffling procedure and the burnup characteristics will be presented in the conference.

Speaker: Prof. Toru Obara (Tokyo Institute of Technology)

Paper 170415(Obara)FR17.pdf

2.1 Commissioning and Operating Experience of Fast Reactors I

Conveners: Mr Alexander FILIN, Mr Suresh Kumar Kannankara Vasudevan (IGCAR, Dept. of Atomic Energy, India)

16 USSR and Russian fast reactor operation through the example of the BN600 reactor operating experience and peculiarities of the new generation BN800 reactor power unit commissioning

The fast-neutron nuclear power industry development began with the BR-5/10 experimental reactors (1959) followed by BOR-60 (1969). The power reactor evolution started with the BN350 commissioning in 1973. In 1980 the BN600 operating up to now was put into operation. In 2015 the BN800 obtained the first criticality. BN600 fast reactor power unit No. 3 of 600 MW power has been put into operation in April 1980 and is under day-to-day operating conditions. Over the operating period the advantages of sodium-cooled fast reactor facility were highly appreciated. The complex tasks were also solved to improve safety and cost-effectiveness of the BN600 reactor facility. Since the commissioning the BN600 reactor facility core has been upgraded three times and the main equipment lifetime has been significantly increased. The work has been carried out to extend the lifetime until 2020, as part of which it has been shown that the strength conditions in all the critical reactor components are not infringed for 45 years of operation. After the events at Japanese Fukushima NPP the action plan aimed at greater resistance of the BN600 reactor facility against external impacts was put in practice. Over the operating period the following were carried out at the BN600 reactor: 1. About 500 experimental fuel sub-assemblies were tested to study structural materials and designs of different types, which in particular allowed the fuel burn-up to be dramatically increased. 2. The technologies of repairs and replacement of the large reactor and steam generator components (72 heat exchangers of the steam generators, 3 low pressure cylinders, 6 feedwater pumps, 3 emergency feedwater pumps) were mastered. 3. The experience of the production of the high-specific activity isotopes was accumulated. 4. The long life tests of the large components operating in sodium were carried out. The most important outcome of the operation is a justification of the construction of the new fast-neutron reactor power units (BN800, BN1200). For 36 years of the safe and reliable operation the main task was fulfilled, i.e. the operation of the powerful unit with the sodium cooled fast reactor and sodium steam generators was mastered. The BN800 was designed using inherent safety principles and applying an additional reactor shutdown system based on the passive operating principle.

Speaker: Mr Yuriy Nosov (Beloyarsk NPP)

Paper FR17-553.doc

17 Main R&D objectives and results for under-sodium inspection carriers – Example of the ASTRID matting exceptional inspection carrier.

In Service Inspection (ISI) of sodium cooled fast reactor prototype ASTRID implies a large R&D effort for associated tools: among others, a specific articulated carrier is being designed to allow exceptional ultrasonic controls of under-sodium core support structure (strongback) at about 200°C. This carrier has to reach deep in the main sodium vessel and yet adapt to the many different weld positions of the strongback, while being simple and robust. Its design thus includes a hollow rigid pole inside which a specific chain can deploy its ultrasonic transducers bearing head in several directions. But first the specific components needed for this carrier have to be developed and tested for these harsh “sodium” conditions : small electrical motor (reducers, sensors), dry bearings, elastomers for leaktightness… Consequently a large qualification program is starting involving tests to be performed with specific samples and prototypes, in air at 200°C, in water, then in sodium.

Speaker: Mr Francois Baque (CEA)

Material 2017-01-18_-_FR17_-_Full_Paper_IAEA-CN-245-417_-_Main_RD_objectives_and_results_for_under-sodium_robot-V2.docx

18 Development of under sodium viewer for next generation sodium-cooled fast reactor

Inspection technique in opaque liquid metal coolant is one of important issue for sodium-cooled fast reactor. To facilitate operations and maintenance activities, various under sodium viewers (USV) has been developed in several research institutes and countries. For example, a horizontal USV, which detects obstacles on the long distance and an imaging USVs, which make images from a short distance and to a middle distance were developed. In this study, the imaging USV from a middle distance, approximately 1 m, was developed. The USV of this study adopts the optical receiving system which measures the vibration of displacement diaphragm by the laser as the receiving sensor. This study mainly focused on the improvement sensitivity in the transmission sensor and the receiving sensor. In addition, the imaging experiment in the water was conducted by using the developed transmission sensor and receiving sensors. From the experimental results, it was confirmed that the developed USV sensors can make imaging with high resolution from 800 mm distance.

Speaker: Mr Kosuke Aizawa (Japan Atomic Energy Agency)

Material FR17(IAEA-CN-245-267).pdf

19 Manufacture, Installation and Adjustment of the BN-800 Reactor Plant Equipment

The construction of Power Unit No. 4 with the BN-800 reactor plant at the Beloyarsk NPP is the crucial stage in the industrial-scale development of the sodium-cooled fast neutron reactors(SFR). The activities on the development of the BN-800 reactor plant commenced in1980. During the period from 1993 to 2005 and later on until the equipment deliveries started, the activities were being in progress to check the engineering solutions of the design in test facilities. To substantiate the reactor plant safety, more than 150 unique R&D activities were accomplished mainly in full-size test facilities. Also, individual assemblies and elements of equipment were tested on mockups. The processes were being developed to fabricate individual assemblies of equipment, assemble and install articles, ensure interactions of sets of devices and mechanisms. Contracts between Rosenergoatom Concern and OKBM provided for fabrication and delivery of more than 150 items of equipment and systems: reactor vessel; heat exchange equipment; CRDMs; fuel handling equipment; purification system equipment; primary and secondary sodium pumps, electromagnetic pumps and their control systems; secondary pipelines and ECDS; sodium tanks, 10–150 m3; sodium valves; metal structures of the reactor compartment; non-standard equipment of the reactor plant; dummy fuel subassemblies (FSA);hot cell equipment; sodium technology instrumentation; ionization chambers; According to the adopted process, the equipment is installed in the Power Unit using two basic methods: 1.Installation work conjoined with the erection of the building 2.Modular installation work on large-size equipment The modular installation work is basically done on the reactor vessel. In a specially erected Reactor Vessel Assembly Building(RVAB), more than 230 supplied units were pre-assembled into the 6 mounting modules that were later transported to the construction site and installed into the reactor pit. The preoperational adjustment activities on the BN-800 reactor plant before the in-house electricity is generated-were performed according to an individual work schedule with the reactor plant equipment attributed to a startup complex. The startup complex ensured that the equipment and systems of the nuclear power station were ready for the gas heatup of the reactor, sodium filling of the reactor and FSA loading into the core with building up the minimum critical mass. The completed deliveries, installation and adjustment work made it possible to accomplish the following in 2013-2016: preoperational adjustment activities; first criticality; pilot industrial operation

Speaker: Mr Igor Petrov (JSC “Afrikantov OKBM”)

Paper IAEA-CN-245-425.pdf

20 Testing and Qualification of Trailing Cable system for Prototype Fast Breeder Reactor

Small and large rotatable plugs (SRP & LRP) are provided to facilitate in-vessel handling of core subassemblies using transfe arm. These plugs are rotated during refuelling of the reactor. The control & instrumentation signals and power to various systems / components located on the rotatable plugs are carried by cables and are connectied to their respective control panels located outside. Among large number of signals / power supply, some are needed during rotation of the plugs also. Trailing Cable System is conceived and designed to carry power/control cables whose continuity is to be ensured during rotations of SRP & LRP. The design requirement for trailing cable system is to accommodate twist of cables between the stationary roof slab and SRP by 540 deg. while maintaining their continuity, which otherwise is not possible. The system is designed with a set of posts mounted one each on SRP & LRP and set of overhanging arms through which, cables are routed in a predetermined way. The overhanging arm bring the cables to the centre of SRP / LRP as the case may be and hence avoids pulling and bending of cables, instead results in twisting. The bunch of cables are freely suspended in the form of ‘S’ shape between roof slab centre and SRP centre with a vertical separation of 3 m between clamping points. The free length of cables is designed to accommodate the twist without causing entanglement of cables. Provisions are made to swing the system away for facilitating handling of components over control plug. Free from interference of trailing cable system with other components, particularly fuel handling machine during plug rotation is also ensured in the design. To carry out functional testing and qualification of the important system, a prototype arrangement between SRP and roofslab was manufactured and erected at full scale Top Shield Layout Model. In line with the approved testing program, the rotatable plugs were rotated by designated angles and twist in cables was critically studied. From the detailed tests, it is observed that the configuration conceived for the supporting structure and the cable routing between SRP and roof slab satisfy the design intent and is capable of maintaining electrical continuity of the cables, meeting functional requirements. Subsequently, the system was qualified through systematic cyclic testing.

Speaker: Mr RAGHUPATHY S. (Indira Gandhi Centre for Atomic Research, Kalpakkam)

Material FR17_Testing_and_Qualification_of_Trailing_Cable_System_323_16-01-2017.pdf

21 Safety Upgradation of Fast Breeder Test Reactor

Fast Breeder Test Reactor (FBTR) has completed 30 years of operation and is relicensed for further operation up to 2018. FBTR has undertaken major upgradation of systems, components and structures to enhance the safety level, based on the operational feedback, maintenance difficulties and obsolescence. Further, post Fukushima, an extensive retrofitting programme is underway to protect the plant against external events such as flood, Tsunami and seismicity. As per the upgradation programme, several major components have been replaced. These include the Neutronic channels, UPS, computers of the Central Data Processing System, main boiler feed pumps, five control rod drive mechanisms, two control rods, central canal plug, deaerator lift pumps, reheaters of the steam water system, station batteries, DM plant, Nitrogen plant, starting air system of the emergency diesel generators, entire fire water system including pumps and isolation dampers of the reactor containment building. Due to obsolescence, 6.6kV MOCB were replaced with VCB and 415V electro-mechanical relays were replaced with numerical relays. Residual life assessment has been carried out for the nonreplaceable components based on the operational history, the design limits for each component by which their capability for continued operation has been ensured. As a part of seismic retrofitting programme, the adequacy of the systems to withstand SSE for safe shutdown, decay heat removal and containment integrity have been assessed. Inparticular plant buildings, anchoring of electrical & instrumentation panels and sodium tanks and other capacities were verified and wooden battery stands of UPS and control power supply were replaced with seismically qualified metallic stands. A new seismically qualified service building is under construction for housing two seismically qualified DG sets and emergency switch gears. Seismic Instrumentation to measure seismic activity in safety structures as well as free-field close to the reactor, is being procured. Supplementary control panel for monitoring the reactor during non-availability of main control room is being implemented. This paper details the various measures implemented for enhancing the safety of FBTR which includes post Fukushima retrofits also.

Speaker: Mr Suresh Kumar K V (Department of Atomic Energy, India, Indira Gandhi Centre for Atomic Reaserch, Kalpakkam, Reactor facilities Group)

Material Safety up gradation of FBTR

3.1 Safety Program

Conveners: Dr Andrei Rineiski (Karlsruhe Institute of Technology (KIT)), Mr Pavel Alekseev (NRC “Kurchatov Institute”)

22 The Status of Safety Research in the Field of Sodium-cooled Fast Reactors in Japan

This paper describes the status of safety research activity in the field of sodium-cooled fast reactors (SFRs) in Japan, mainly on severe accident related issues. Core damage sequences are analyzed by applying probabilistic risk assessment methodology and categorized into typical accident phases, i.e., initiating phase, transitions phase, and material relocation and cooling phase. In order to utilize superior characteristics of sodium as coolant, achievement of in-vessel retention is one of important objective of safety design and evaluation for SFRs. Focus is on the later phases of accidents for which experimental data acquisition and code development are going on. A series of out-of-pile and in-pile experiments in EAGLE-3 and related tests in the MELT facility are being conducted for molten fuel discharge and cooling. Study on debris bed formation and self-leveling effect is also conducted. A fast-reactor safety analysis code, SIMMER, is developed to enhance its capability to be applicable such phenomena in the later phase of accidents.

Speaker: Prof. Koji Morita (Kyushu University)

Material FR17_paper_IAEA-CN-245-28_mod2.pdf

23 RECENT ACTIVITIES OF THE SAFETY AND OPERATION PROJECT OF THE SODIUM-COOLED FAST REACTOR IN THE GENERATION IV INTERNATIONAL FORUM

The Generation IV (GEN-IV) international forum is a framework for international cooperation in research and development (R&D) for the next generation of nuclear energy systems. Concerning the sodium-cooled fast reactor (SFR) system, there are five cooperation projects for R&D. The SFR Safety and Operation (SO) project addresses the area of the safety technology and the reactor operation technology developments. The aim of the SO project includes (1) analyses and experiments that support establishing safety approaches and validating performance of specific safety features, (2) development and verification of computational tools and validation of models employed in safety assessment and facility licensing, and (3) acquisition of reactor operation technology, as determined largely from experience and testing in operating SFR plants. The tasks in the SO topics are categorized into the following three work packages (WP): WP-SO-1 “Methods, Models and codes” is devoted to the development of tools for the evaluation of safety, WP-SO-2 “Experimental Programs and Operational Experiences” includes the operation, maintenance and testing experiences in experimental facilities and SFRs (e.g., Monju, Phenix, BN-600 and CEFR), and WP-SO-3 “Studies of Innovative Design and Safety Systems” relates to safety technologies for GEN-IV reactors such as active and passive safety systems and other specific design features. In this paper, recent activities in the SO project are described.

Speaker: Dr Alfredo Vasile (CEA)

Paper FR17_SO_Paper_v8.pdf

24 “ASTRID safety design: Radiological confinement improvements compared to previous SFRs”

ASTRID is the Advanced Sodium Technological Reactor for Industrial Demonstration which is intended to prepare the Generation IV reactor, with improvements in safety and operability. In order to meet the objectives of the 4th generation reactors and comply with the related specifications, the ASTRID project integrates innovative options. In the earlier phase of ASTRID project, a specific safety approach was set and its main guidelines were agreed by the French Nuclear Safety Authority. This basic safety design guide is currently applied as reference for the choices of the design options. The paper presents the safety approach, called “top-down” approach, relating to the “confinement” safety function. The confinement design of ASTRID has several safety objectives from both radiological point of view and sodium chemical risk, and its design is based on “plant state” approach. As concerns potential radiological risk, main objectives are to postpone a hypothetical off-site release of radiological material coming from core degradation and also to decrease its health and environmental possible consequences. As concerns the sodium chemical risk, main objectives are to prevent by design an overpressure of the containment, introducing drawbacks in terms of confinement, and also to cope with the risk of off-site release of soda aerosols with possible health effect. In order to meet all these objectives, design provisions are taken, considering the different release ways inside the confinement. The paper presents the lessons learned from the previous SFR confinement and the method applied to choose for ASTRID consistent design options. Major part of these design provisions has, in particular, an important function of severe accident mitigation. The design of these mitigation provisions takes into account the lessons learned from Fukushima event, in order to prevent any cliff edge effect in terms of radiological consequences.

Speaker: Mr Stéphane BEILS (AREVA NP)

Paper Paper_IAEA-CN-245-476_ASTRID_Radiological_confinement_improvement_upload....docx Paper_IAEA-CN-245-476_ASTRID_Radiological_confinement_improvement_upload....pdf

25 Safety Assurance for BN-1200 Power Unit During Accidents

Safety analysis of BN-1200 RP Unit is performed with account of requirements specified for innovative nuclear technologies including IAEA recommendations. The following BDBAs are considered within BN-1200 project:  Heat removal accidents with loss of energy supply, failures of active systems for reactor shutdown, failures of normal cooldown system and active components of emergency cooldown system;  Accidents with insertion of positive reactivity and failure of active reactor shutdown systems;  Accidents with plugging up of FSA flow area. Computational validation of safety in accidents is performed using verified computational codes of new generation. To validate safety, an integrated approach is applied that analyses main processes and phenomena occurring in the RP and Unit’s rooms during accident. Computational results showed that there is no need to evacuate or move out population under projected population radiation doses for population beyond the NPP site.

Speaker: Mr Artem Anfimov (Joint Stock Company "Afrikantov Experimental Design Bureau for Mechanical Engineering")

Paper IAEA-CN-245-385.pdf

Slides IAEA-CN-245-385.pptx

26 The SAIGA experimental program to support the ASTRID Core Assessment in 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 an ASTRID-like fuel in the IGR reactor (Impulse Graphite Reactor operated by NNC). The purpose of the SAIGA program is to qualify the SIMMER computer code on the SEASON platform based on tests conducted with axially heterogeneous CFV type ASTRID inner core pins or pin bundles in hypothetical severe accident situations. These tests should be representative, as much as possible, for the phenomena encountered during severe accident sequences considered for ASTRID. The feasibility study aimed to study the generic accident families of loss of coolant and power excursion situations. It is important to point out that the fuel used for these tests can only be a non-irradiated fuel. The feasibility study focused on tests based on the degradation of one or more fuel pins during Total Instantaneous Blockage (TIB) sequences in a sub-assembly and power excursion (Transient OverPower: TOP) sequences as in SCARABEE and CABRI with homogeneous pins. For both scenarios, the feasibility study defined the main characteristics of the experimental devices and the operating conditions for the tests to be conducted in the IGR reactor. The purpose of the studies was to assess the capacity of the IGR reactor to provide the necessary neutron flux during all the transients, to demonstrate the capacity to carry out on-line or post-test measurements of the variables of interest, to study the feasibility of the sodium loop feeding the test device and to assess the cost and timetable for a program of 3 tests. Preliminary calculations carried out using the SAS-SFR and SIMMER codes were used to simulate the degradation of the fuel during TOP and TIB type tests, respectively. Based on the information obtained during the feasibility study, specification requirements were given to perform three useful and potentially feasible tests inside the SAIGA program i.e.: 1) Ejection and relocation of fuel in a narrow hydraulic channel (CFV type) with a heterogeneous fuel during a power excursion (TOP type scenario) 2) Loss of flow test on a CFV-type fuel sub-assembly 3) Propagation of a corium pool outside the sub-assembly in the presence of a corium discharge area filled with sodium

Speaker: Mr Frédéric Payot (CEA : Commissariat à l’Energie Atomique et aux Energies Alternatives)

Material apn-Paper_FR17_PAYOT_vf.docx

27 Study on Safety Design Concept for future Sodium-cooled Fast Reactors in Japan

This paper describes safety design concept for future sodium-cooled fast reactors (SFRs) in Japan, which is based on the safety design criteria and safety design guidelines under development in the international forum of generation IV nuclear energy systems. The future safety design of SFRs should be advanced taking the feedback of experiences, achievement of existing technology, and innovative technology into account. Inherent and/or passive design features are utilized based on SFRs characteristics such as low pressure, high thermal inertia of the system. Lesson learned from the Fukushima Dai-ichi accident is one of important issue to be incorporated into the safety design concept. In order to realize commercial SFRs in the future, robust and rational safety design should be pursued by integration of various factors in the design, limiting additional specific systems, structures and components. Existing engineering bases for design and manufacturing of SFRs components, and innovative technologies introduced in the FaCT project are keys to realize the safety concept.

Speaker: Mr Shigenobu Kubo (Japan Atomic Energy Agency)

Paper kubo.shigenobu

5.1 Advanced Fast Reactor Fuel Development I

Conveners: Mr CLEMENT RAVI CHANDAR SOWRINATHAN (INDIRA GANDHI CENTRE FOR ATOMIC RESEARCH), Mr Vladimir Troyanov (JSC "Rosenergoatom Concern")

28 Fabrication Characteristics of Injection-cast Metallic Fuels

The fabrication process of metallic fuels for sodium-cooled fast reactor (SFR) was developed using the injection casting. U-Zr-RE(Nd-Ce-Pr-La) fuel slugs were fabricated and characterized to optimize the injection casting process. The microstructure examined by SEM showed that precipitates were uniformly distributed over the fuel slug. The fuel weight loss after the injection casting was measured to be about 1.5%. The reaction between the melt and the crucible was found to be significant in the fabrication of RE-containing fuel slugs compared to U-Zr fuel slugs. The pressurized injection casting method was also developed to fabricate the fuel slugs containing volatile elements. U-Zr-Mn fuel slugs were fabricated as a surrogate for Am-bearing metallic fuels under three different melting pressure conditions. From the chemical composition analysis by the ICP-AES method, no evaporation of Mn was detected in the fuel slugs fabricated under Ar atmosphere higher than 400 torr.

Speaker: Dr Jeong-Yong Park (Korea Atomic Energy Research Institute)

Paper FR17_JYPark_revised_final.pdf

29 Metal fuel for fast reactors, a new concept

Choice of the fuel composition is important question to improve the competitiveness of fast reactors. It should have a high density and thermal conductivity, a high concentration of fissile nuclide, and high manufacturability. The best fuel composition of fast reactors remains a metallic nuclear fuel, based on uranium and plutonium alloys. The undeniable advantages of a metal fuel composition is high density of 15-18 g/cm3; high thermal conductivity λ = 30-40 W/m∙K; the ability to achieve ultra-deep burnup; simplicity of recycling spent nuclear fuel, based on conventional metallurgical methods. Significant disadvantages of metallic fuels are a large swelling by the gas and the possibility of irradiation growth in the case of injection-molded parts with a pronounced texture, as well as the opportunity to interact with the fuel cladding above 700 C. As a solution to these problems, connected channel creation in the fuel core to the output of the gas fission products for the entire fuel campaign is proposed. It can be realized by creation of open porosity 15-25% of the entire volume of the fuel pellet by applying the methods of powder metallurgy. Due to the deformation of the porous structure of the fuel core reduces the risk of cladding damage, and close contact "fuel-clad" provides minimal contact resistance, which improves heat dissipation. Usage of metallic fuel tablet simplifies the fuel element technology and equipment, it makes possible to use fuel elements with gas fuel-cladding gap (as evidenced by the thermophysical calculations). Compatibility metal fuel with cladding can be increased by usage of vanadium alloys or ferritic steels as cladding materials. Preparation of porous structure in fuel is impossible without usage the powder metallurgy techniques. Wherein a feature of uranium alloys is the difficulty of obtaining compacts by conventional pressing and sintering, it is associated with high oxidative capacity of uranium powder. We use advance methods of compaction. High-voltage electrodischarge consolidation is based on passing an electric current through the powder compact, with the simultaneous application of pressure. The advantages of this method is the short time of compaction (milliseconds), high density products. Due to the short sintering time consolidation comes with minimal changes in the microstructure (grain growth, recrystallization). Combining technological stages of sintering and pressing has a positive impact on performance. The final density of the product is achieved by selecting parameters such as pressure, voltage, current density.

Speaker: Dr Boris Tarasov (NRNU MEPHI)

Material IAEA-CN-245-347.docx

30 Development of innovative fast reactor nitride fuel in Russian Federation: state-of-art.

The nitride fuel is selected as an advanced fuel for fast reactors in Russia. Within the framework of "PRORYV" project a comprehensive program of calculation-experimental study of mixed uranium-plutonium nitride performance for BN-1200 and BREST-OD-300 reactors has been developed. The program provides for works to improve the fabrication technique, composition and structure of nitride fuel, to measure out-of-pile properties, to carry out reactor tests in the MIR, BOR-60 research reactors and in the BN-600 commercial reactor, as well as post-irradiation examination (PIE) of all experimental fuel assemblies (FA). Reactor tests are accompanied by pretest calculations by DRAKON and KORAT fuel codes. For the nitride fuel fabrication the carbothermal synthesis technology of nitride oxide powders, which are the product of the current radiochemical industry, is used. The laboratory technique of carbothermal synthesis of starting powders developed at JSC "VNIINM" is implemented on a larger scale at JSC "SCC" in Seversk, where the possibility of full-scale production of experimental FAs of BN-600 reactor is created. Nitride fuel pellets have been fabricated for more than 500 fuel pins for all BN-600 experimental FAs. Today 9 FAs are under irradiation in the BN-600 reactor. PIEs of one FA have been completed. 7 dismountable FAs with 7 nitride pins in each are under irradiation in the BOR-60 reactor. Fuel and fuel pins have been fabricated at JSC "VNIINM". All BOR-60 and BN-600 experimental nitride fuel pins are intact.

Speaker: Dr Liudmila Zabudko (Innovative &Technology Center by "PRORYV" Project)

Paper FR17-062-Development_of_innovative_fast_reactor_nitride_fuel_in_Russian_Federation.docx

31 Conceptual design of fuel and radial shielding sub-assemblies for ASTRID

The French 600 MWe Advanced Sodium Technological Reactor for Industrial Demonstration (ASTRID) project reached in 2015 the end of its Conceptual Design phase. The core design studies are being conducted by the CEA with support from AREVA and EDF. Innovative design choices for the core have been made to comply with the GEN IV reactor objectives, marking a break with the former Phénix and SuperPhénix Sodium Fast Reactors. The CFV core of ASTRID demonstrates an intrinsically safe behavior with a negative sodium void worth achieved thanks to a new fuel sub-assembly design. This one comprises (U,Pu)O2 and UO2 axially heterogeneous fuel pins, a large cladding versus small spacer wire bundle, a sodium plenum above the fuel pins, and an upper neutron shielding with both enriched and natural boron carbide. The upper shielding also maintains a low secondary sodium activity level and is made removable on-line through the sub-assembly head for washing compatibility. Calculations have been performed to increase the stiffness of the stamped spacer pads in order to analyse its effect on the core mechanical behaviour during hypothetical radial core compaction events. Concerning the radial shielding sub-assemblies surrounding the fuel core, heavy iterative studies were performed in order to fulfill ASTRID requirements of minimising the secondary sodium activity level and maximising the in-core life-time. Evaluated options were reflectors sub-assemblies made of steel or MgO rods, and radial neutron shielding sub-assemblies made of B4C or borated steel, with different configurations in the design and in the core layout. This paper describes the design of the fuel and radial shielding sub-assemblies for the ASTRID CFV v4 core at the end of the Conceptual Design phase. Focus is placed on innovations and specificities in the design compared with former French SFRs.

Speaker: Dr Victor BLANC (CEA)

Paper FR17_-_IAEA-CN-245-128_-_Design_of_Astrid_fuel_&_radial_shielding_subassemblies.pdf

32 Development of Electromagnetic Devices for Sodium Cooled Fast Reactor Application

Liquid sodium is used as coolant due to its suitable neutronic and thermal properties in fast reactors. Good electrical conductivity of sodium is used for development of electromagnetic devices such as electromagnetic pumps & flowmeters and level probes for use in sodium cooled fast reactors, where conventional devices used in chemical plant cannot be used due to high chemical activity of sodium and high temperature. Design, development and testing of a Sodium Submersible Annular Linear Induction Pump (ALIP) was carried out recently. The developed pump can be used for sodium draining from main vessel of pool type of Sodium Cooled Fast Reactor (SFR) and any other application where pump has to be submerged in sodium. The developed pump does not require any external cooling when submerged in radioactive sodium of 200oC. The winding of submersible ALIP can withstand 550oC. The submersible ALIP was tested in sodium loop for obtaining pump characteristics. AC Conduction pump for low flow application in sodium loop has been developed. Design, analysis and manufacturing aspects are brought out in the paper. Development of three different types of compact electromagnetic flowmeters based on Samarium Cobalt permanent magnet, electromagnet formed from soft iron in combination with mineral insulated cable and small probe type permanent magnet flowmeters were successfully demonstrated. Samarium Cobalt magnet helps in reducing the size and weight of flowmeter due to its high energy product. Flowmeter having electromagnet coil made from mineral insulated cable has high temperature withstand capability of around 500oC. The electromagnet coil in combination with soft iron replaces permanent magnet, hence it provides diversity in flow measurement in critical applications. The probe type flowmeter uses small permanent magnet encapsulated in a slender probe which can be inserted inside the pipe where sodium flow measurement is required. Eddy current based ex-vessel level probe was developed for measurement of sodium level in the vessel without insertion of probe inside the vessel. It works on the principle of eddy current and using this probe, sodium level inside the stainless steel vessel can be obtained by keeping the probe outside the vessel. This technique of discrete sodium level measurement is first of its kind. This paper enumerates development of sodium submersible ALIP, newly developed flowmeters and development of ex-vessel sodium level probe. Test results obtained from sodium testing are also brought out in the paper and FEM analysis carried out for different devices are also depicted.

Speaker: Dr B.K. Nashine (Indira Gandhi Centre for Atomic Research, Kalpakkam (T.N.) - India)

Material FR_17_IGCAR_IAEA-CN-245-174_EM_Devices.docx

33 Fuel Cladding Chemical Interaction Tests of Irradiated Metallic Fuel

To investigate the fuel cladding chemical interaction for the irradiated metallic fuel, high temperature heating tests were performed. The fuel rod consisting of U-10Zr-5Ce fuel with T92 cladding were irradiated in HANARO reactor. After the irradiation, the fuels was cut into cylindrical specimens, and then the top and the bottom plates of the specimens were put into contact with FC92 and HT9 plates, respectively. The specimens were exposed at high temperature in the range of 650–800 oC for one hour. Microstructural examinations were conducted by utilizing optical microscope, scanning electron microscope, and electron prove micro-analysis. Migration phenomena of U, Zr, Fe, and Cr as well as Nd lanthanide fission product were observed at a melting region. Elements distribution at the melting region demonstrates that eutectic melting occurs during high temperature experiment. The penetration depth of the eutectic melting in FC92 and HT9 were compared with that for T92 cladding.

Speaker: Dr Ju-Seong Kim (Korea atomic energy research institute)

Material FR17_JuSKIM.docx

6.1 CFD and 3D Modeling

Conveners: Prof. Leonid Bolshov (Nuclear Safety Institute of the Russian Academy of Sciences), Dr Velusamy Karuppanna Gounder (Indira Gandhi Centre for Atomic Research)

34 Numerical simulation of hydraulics and heat transfer in the BREST-OD-300 LFR fuel assembly

This paper presents an analysis of hydraulic and heat transfer phenomena in a lead coolant flow in a fuel assembly (FA) of the BREST-OD-300 reactor core central subzone based on CFD simulations (RANS) using the STAR-CCM+ code. Based on the simulation results, coolant pressure, velocity and temperature fields, and the fuel cladding temperature distribution have been obtained. The FA design comprises a pin bundle (p/d = 1.38) with spacer grids, and the outer shroud has been eliminated. The influence of the FA design features on the hydraulic characteristics and the cladding temperature distribution has been shown. The CFD model was validated on experimental data on the hydraulic characteristics of a full-scale FA model. A good agreement has been shown between calculated and experimental data on the pressure drop both for the FA parts (head, pin bundle and tail) and for the FA as the whole.

Speaker: Dr Dmitry Fomichev (JSC NIKIET)

Material paper548_fomichev_eng_final.docx

35 Steady State Modelling and Validation of Once Through Steam Generator

One dimensional steady state model is developed for counter current shell and tube once through steam generator for PFBR with water flowing through tubes and sodium through shell. All the heat transfer processes i.e. preheating, evaporation of water and superheating of the steam happen along the length of the tube. The length of the steam generator is divided into a number of control volumes. For modelling water side, continuity, momentum and energy equations are solved for single phase water, two phase steam and super heated vapour over the length of the tube using homogeneous model. The momentum and continuity equation on the sodium side are not solved as the density variation of sodium with pressure is negligible. The differential equations are discretised to linear algebraic equations using finite difference scheme. All three terms (advection, frictional and gravitational) are considered in momentum equation of water, while only advection and source term are considered in energy equation of water and sodium. The discretisation of the advection term is carried out using upwind scheme and of frictional term by backward differencing. The continuity equation for water is discretised using backward differencing. The linear algebraic equations are then solved simultaneously using numerical methods to get the temperature and pressure profiles in the tube and shell side. Water side heat transfer coefficient is calculated using Steiner and Taborek asymptotic model and Subbotin correlation is used for sodium side heat transfer coefficient. The code developed can be applied to simulate similar steam generators of any length with any number of tubes. Towards validation of the mathematical model and the solution method, 19 tube steam generator tested in in-house steam generator test facility (SGTF) has been simulated and the predicted results are compared with the measured data. Results from the code match well with experimentally observed data from the facility. In addition, grid sensitivity studies have been carried out to establish consistency in the solution.

Speaker: Dr Velusamy Karuppanna Gounder (Indira Gandhi Centre for Atomic Research)

Paper Paper_Number_332(VivekSingh).pdf

36 Applications of the DNS CONV-3D Сode for Simulations of Liquid Metal Flows

For the simulation of the thermalhydraulic processes in fast reactors with liquid metal coolant DNS CFD code CONV-3D has been developed. This code has ideal scalability and is very effective for calculations on high performance cluster computers. The code has been validated on the set of analytical tests and experiments in a wide range of Rayleigh and Reynolds numbers, in particular, at extremely small Prandtl numbers. The paper presents the results of the application of CONV-3D code for simulation of sodium natural convection in the upper plenum of the MONJU (Japan) and BN-600 (Russia) reactor vessel. A satisfactory agreement of the numerical predictions with experiments is demonstrated. The calculation results of the experiment conducted on the Phenix facility (France) with sodium coolant are demonstrated. The experiment focuses on the mixing of two fluxes at different temperatures in the secondary circuit of reactor facility with liquid metal coolant in the presence of a bending tube. A small pipe is connected via T-connection to the main pipe and unloads of sodium in the main pipe at a temperature which is higher than in the main pipe. A satisfactory agreement of the numerical predictions with experiments and commercial codes is demonstrated, in particular for the temperature distribution vs the coordinates. The results of simulation of heavy-liquid metal (LBE) flow and heat transfer along a hexagonal 19-rod bundle with wire spacers (KALLA, Germany) are presented. A convergence on a sequence of grids and convergence with the experiment is demonstrated. The results obtained allow to conclude that using of CONV-3D code with high predictive power can be recommended for reactor applications.

Speaker: Mrs Vladimir Chudanov (Nuclear Safety Insitute RAS (IBRAE))

Material paper_CN245_eng_corrected.docx

37 3D Modeling of Fuel Handling System for PFBR Operator Training Simulator

Recent advances in computer science have paved way to increased ease in design of nuclear power plants and imparting efficient operator training thereby improving the safety of nuclear plants. As nuclear power plant is considered safety critical, need to improve the skill set of the control room operators by getting the operator trained in simulator is becoming a mandatory requirement before commissioning the plant. Theoretical models, response and actions in control room panels/consoles, human reasoning etc. are simply not enough to impart full understanding of complex systems in reactor to the operators. Over the last few years, addition of 3D models along with the simulated control systems has gained wide attention in training simulators. Additional 3D models aid the operator in understanding the system and helps in quicker and more efficient training. One of the most important systems in Prototype Fast Breeder Reactor (PFBR) is its fuel handling system. It requires complex predefined sequential blind operations to be performed by the operator. Many of the operations are needed to be carried out in handling control room through panels and consoles. The skill set of the operator needs to be enhanced as the operations are critical and some of them are directly linked with the safety of the nuclear reactor. This paper covers the details of PFBR Full Scope Replica Simulator, various 3D models developed for Fuel handling System along with their associated process and logic models. The sequential procedure based on interlocks and the responses of the 3D fuel handling component models to the operator actions in simulator are also elaborated.

Speaker: Mr Venkatesan Arasappan (Indira Gandhi Centre for Atomic Research)

Material This paper covers the details of PFBR Full Scope Replica Simulator, various 3D models developed for Fuel handling System along with their associated process and logic models.

38 Numerical Simulation Method of Thermal Hydraulics in Wire-wrapped Fuel Pin Bundle of Sodium-cooled Fast Reactor

A numerical simulation system, which consists of a deformation analysis program and three kinds of thermal-hydraulic analysis programs, is being developed in Japan Atomic Energy Agency in order to offer methodologies to clarify thermal-hydraulic phenomena in fuel assemblies of sodium-cooled fast reactors under various operating conditions including fuel deformation. In this paper, we focus on SPIRAL which is one component program of the numerical simulation system and plays the role to simulate detailed local flow and temperature fields in a wire-wrapped fuel pin bundle. SPIRAL adopts finite element method in order to treat complicated geometries and a hybrid turbulence model which has computation efficiency similar to high Re number models and high accuracy similar to low Re number models. As a validation study, SPIRAL was applied to several kinds of analyses of water/sodium experiments using wire-wrapped fuel pin bundles. Applicability of SPIRAL to the prediction of flow and temperature fields as well as pressure loss coefficients will be discussed.

Speaker: Hiroyuki Ohshima (Japan Atomic Energy Agency)

Paper FR17-453.docx

39 Modelling and Simulation of Heat Transport System and Steam Power Transition System of CEFR

In this paper, the graphics interface and parameter modeling real-time simulation software, Jtopmeret, was used to model and simulate the intermediate heat transport system and the steam power conversion system of China Experimental Fast Reactor (CEFR). The two-phase, multi-component models were taken into consideration to simulate the flow and heat transfer of working medium sodium and steam-water Ranking circle. The matrix solving method was used in this paper to solve the mass, momentum, and energy conservation equations accurately, quickly and steady. Operating characteristics under steady, transient and malfunction operations of CEFR were researched. The simulation results showed that the errors of main parameters under different steady operations were less than 1%, the trend curves under transient operations and malfunction operations were reasonable, and the response of the secondary and third loop could show the operating and safety characteristics of CEFR. The models had been applied to full scale simulator of CEFR. Keywords: China Experimental Fast Reactor (CEFR), modeling and simulation, two-phase, multi-component models

Speaker: Prof. Zhao-Fei TIAN (Fundamental Science on Nuclear Safety and Simulation Technology Laboratory, Harbin Engineering University)

Material Simulation_Study_on_Secondary_and_Third_Loop_of_CEFR.docx

7.1 Sustainability of Fast Reactors

Conveners: Mr Andrey Gulevich (SSC RF-IPPE), Mr Seiichiro Maeda (Japan Atomic Energy Agency)

40 Current Status of Next Generation Fast Reactor Core & Fuel Design and Related R&D in Japan

The next generation fast reactor is being investigated in Japan, aiming at several targets such as “safety”, “reduction of environmental burden” and “economic competitiveness”. As for the safety aspect, FAIDUS (fuel assembly with inner duct structure) concept is adopted to avoid re-criticality in core destructive accidents. The uranium-plutonium mixed oxide (MOX) fuel, in which minor actinide (MA) elements are included, will be applied to reduce the amount and potential radio-toxicity of radioactive wastes. The high burn-up fuel is pursued to reduce fuel cycle cost. The candidate concept of the core and fuel design, which could satisfy various design criteria by design devisals, has been established. In addition, JAEA is investigating material properties and irradiation behavior of MA-MOX fuel. JAEA is developing the fuel design code especially for the fuel pin with annular pellets. Furthermore, JAEA is developing oxide dispersion strengthened (ODS) ferritic steel cladding for the high burnup fuel.

Speaker: Mr Seiichiro Maeda (Japan Atomic Energy Agency)

Material ID269_JAEA_Maeda_Proceeding_of_Core_fuel_design_and_related_RDs.docx

41 Assessment of a nuclear energy system based on the integral indicator of sustainable development

Ability of nuclear power to meet the requirements of sustainable development is a critical point for public acceptance of the energy technology and its further development. The method for assessment of the nuclear energy system (NES) compatibility with the requirements of the UN concept of sustainable development was developed in the IAEA within the activities of the INPRO international project. The approach discussed in the paper is aimed at decreasing the uncertainty of the method in assessing NES sustainability in order to duly arrange the NESs from those at the low limit of sustainability up to advanced systems with much better sustainability features. It is shown that the integral index determined with the use of common approaches of the multi-attribute theory may provide a measure of the NES sustainability to be used for quantitative comparison of sustainability of NESs based on different technical and institutional solutions. The value of the index depends on the reaching by an assessed NES of certain key developments in the seven subject areas: economy, nuclear safety, resource supply, waste management, non-proliferation, public acceptance and international cooperation. It is shown that assessment of the NES based on OTFC with thermal reactors and UOX fuel on a large time horizon will be characterized by a low value of the sustainability index in case of lack of natural uranium resource and continued accumulation of spent nuclear fuel and plutonium therein. The index radically increases due to synergistic use of thermal and fast reactors with closed nuclear fuel cycle. This effect can be gained within the country that has mastered both technologies or in the system in which user country receives services on closed NFC from supplying country. It is noted that the model needs further development in order to take into account more factors related to the sustainability notion.

Speaker: Mr Vladimir Usanov (JSC «SSC RF-IPPE», State Corporation “ROSATOM”)

Paper NES SUSTAINABILITY INDEX

42 Performance and sustainability assessment of nuclear energy deployment scenarios with fast reactors: advanced tools and application

A performance and sustainability assessment of nuclear energy deployment scenarios with fast reactors is a multi-criteria problem which is determined by a wide spectrum of criteria characterizing resource consumption, economic performance, risks of unauthorized proliferation, safety and waste management etc. In determining the most promising scenario, it is necessary to consider the conflicting nature of the criteria because an improvement in one of the criteria causes, as a rule, a deterioration of values in the others. To increase the validity of judgments formulated on the basis of calculations, an uncertainty analysis is also required. There is a need to use multiple criteria decision making methods. Multiple criteria decision making methods are a support tool aimed to help experts and decision makers who are faced with numerous conflicting assessments to highlight conflicts and perform proper trading off during the decision making process. A multi-criteria decision analysis and multi-objective decision making constitute the main components of multiple criteria decision making. The major distinction between these groups of methods is based on whether the solutions are defined explicitly or implicitly. Although tasks that can be solved using these methods are different, their combined use seems to be appropriate to identify the most suitable and balanced nuclear energy deployment scenarios with fast reactors despite various contradiction criteria. The both groups of methods may be applied to assess the performance and sustainability of nuclear energy deployment scenarios with fast reactors by searching for compromises between the conflicting system factors that determine the nuclear energy deployment scenarios and selecting the trade-off option, to carry out multi-objective optimization of nuclear energy structures, taking into account the nuclear energy system evolution, the structure and the organization of nuclear fuel cycle and the most important system constraints and restrictions. The paper presents the toolkit developed in the National Research Nuclear University MEPhI for a performance and sustainability assessment of nuclear energy deployment scenarios with fast reactors providing a solution to the problem of optimizing and comparing nuclear energy deployment scenarios with fast reactors in multiple criteria formulation. Some results of implementation of this toolkit for the performance and sustainability assessment of nuclear energy deployment scenarios with fast reactors are presented which demonstrate that technologically diversified nuclear energy structures in which several different fast reactor technologies are present may produce a synergetic effect in terms of nuclear energy system sustainability and performance improvement.

Speaker: Mr Andrei Andrianov (National Research Nuclear University, MEPhI)

Material _Andrianov_IAEA_FR-17.docx

43 Comparison of Innovative Nuclear Energy Systems Based on Selected Key Indicators and Their Weighing Factors

The paper presents a methodological study on comparison of nuclear energy systems whose commissioning and commercial-scale operations are in the planning stage. These studies are carried out within the framework of the joint project INPRO. Systems with numerous technical and economic uncertainties are poorly amenable to technology assessment using the INPRO methodology (for all areas, basic principles and criteria). Furthermore, it seems unreasonable to assess reactor systems in isolation from the system they were designed for. Some indicators from among those that should be referred to the key ones are system-related indicators. They directly affect not only the assessment of reactor facility, but also the characteristics of nuclear energy system as a whole. At the same time, there are indicators, which are slightly related to the system or are generally unrelated to it. Therefore, the study compares particularly the systems with their inherent key indicators, rather than the technologies taken separately. For the comparison of innovations, a set of key indicators was selected and aggregated into a single function by assigning weights to each indicator. At this stage, there is also an uncertainty both in key indicator assessments, and in determining the importance of each indicator among those selected (weighing factors of indicators). The paper discusses several key indicators of sustainability for innovative nuclear energy systems from different areas of assessment (economics, technology readiness, waste management). As an example, consideration is given to several types of countries with different nuclear capacity scales. Key indicator weights were selected based on the intrinsic features of the countries; the comparison of innovations is presented. The paper presents the sensitivity of the weights of selected key indicators to the result of innovations comparison. In addition, the paper identifies the ways of key indicators development. The study of assessment uncertainties using the key indicators could be one of the methodological improvements. Together with the sensitivity analysis of key indicators’ weighing factors, it can significantly promote the study of the comparison of innovations in the context of the methodology described in the paper.

Speaker: Mr Alexander Yegorov (researcher of Laboratory of Nuclear Fuel Cycle system analysis taking into account National and International tendencies)

Material Contribution_ID434_EgorovRus_v1.docx

44 Evaluation results of BN-1200 compliance with the requirements of GENERATION IV and INPRO

The project of a power unit with BN-1200 reactor is designed using advanced technical solutions, which define evolution of the fast breeder technology in the field of safety parameters and in the field of technical and economical indicators. At present there was completed estimation of the BN-1200 project from the point of view of its compliance with the requirements of nuclear energy systems of Generation IV in the frames of International forum Generation IV, and comparison of the BN-1200 project with other fast breeder projects using NESA INPRO procedure, developed and verified for comparison of nuclear energy systems with PWR. The paper presents the results of preliminary estimation by the INPRO procedure, which showed that BN-1200 has good margin of safety and economical characteristics in comparison with the previous projects; and BN-1200 meets all the basic principles in the fields of ‘safety’ and ‘economics’; and BN-1200 can ensure sustainable development of the nuclear energy system. Estimation results of the BN-1200 concept for compliance with the requirements to Generation IV plants testify that BN-1200 project, as a whole, has good potential from the point of view of compliance with the stated requirements.

Speaker: Ms Elena Marova (JSC “Afrikantov OKBM”)

Paper IAEA-CN-245-399.pdf

Slides IAEA-CN-245-399.pptx

3:00 PM Coffee Break

1.2 SFR DESIGN & DEVELOPMENT - 2

Conveners: Mr Puthiyavinayagam Pillai (Indira Gandhi Centre for Atomic Research), Mr Sergey Shepelev (JSC «Afrikantov OKBM»)

45 Progress of Design and related Researches of Sodium-cooled Fast Reactor in Japan

In Japan, we have implemented the development of a sodium-cooled fast reactor from the viewpoint of severe accident measures in order to strengthen safety of a fast reactor since the Great East Japan Earthquake. This paper describes the progress of design study and research and development (R&D) related to safety enhancement and severe accident measures. For the purpose of strengthening of decay heat removal function, we are performing R&D and development of test facilities on the decay heat removal after core disruptive accident (CDA), the application of a variety of heat removal system, and the evaluation methods for thermal hydraulics. In order to elucidate the behavior of molten fuel during CDA, we are conducting the in-pile and out-of-pile tests by international collaboration, the basic experiments, and the development of evaluation methods for CDA. Also, we have promoted the improvement of core design from the viewpoint of preventing the occurrence of severe accident.

Speaker: Dr Hideki Kamide (Japan Atomic Enery Agency)

Paper FR17_IAEA-CN-245-298_R6.pdf

46 Lessons and strategies from PFBR to Future Fast Breeder Reactors

BHAVINI, a public sector unit under Department of Atomic Energy, is responsible for construction, commissioning and operation of fast reactors in India. Prototype Fast Breeder Reactor (PFBR) which is in advanced stage of commissioning is the forerunner for the second stage of India’s three stage nuclear programme. PFBR is a 1250 MWt (500 MWe), MOX fuelled, sodium cooled, pool type fast reactor. It is a first of its kind reactor with total indigenous technology. Starting from civil construction, manufacturing of over-dimensional & precision machined components, installation, integration, till commissioning and operation of all the mechanical, electrical and control & instrumentation systems, there were many challenges and surprises which have been addressed one by one in a systematic manner. The experiences gained during various phases of PFBR project have enriched the scientists and technologists to fine tune the specific aspects in design, sizing of layout, manufacturing & transportation methodologies, sequence of installation and commissioning of the plant and equipment. It is clear that special attention is needed for achieving leak-tightness, making provisions for pre-service and in-service inspections, appropriate routing of power & instrumentation cables and protecting nuclear & process instrumentations. The project management for the future fast breeder reactors, twin units of 2 x 600 MWe will be well established based on the feedback from PFBR. Concept of twin units will be beneficial for both economy and time schedule. The site assembly shop can cater the need for fabrication of individual components of reactor assembly meeting the stringent tolerance limits and appropriate integration, so that handling and erection of the assembly will be cost effective and time beneficial. Advance planning is required for achieving leak-tightness of integrated assemblies. The well planned sequence of layout of sodium and associated piping, their interfaces with the equipment, provision of redundant heaters, thermocouples, and leak detectors will play key role in project schedule. This paper details out the experiences gained and lessons learnt from PFBR and the strategies to be adopted for future fast reactors towards safety, economy and time schedule.

Speaker: Mr Rajan Babu Vinayagamoorthy (Director (Technical))

Paper FR-17_Paper_522_Lessons_and_strategies_from_PFBR_to_Future_Fast_Reactors_in_India_28.04.2017.pdf

47 Status of ASTRID Nuclear Island Design and Future Trends

In the frame of the French ASTRID project led by CEA, AREVA NP is in charge of the design studies of the whole Nuclear Island. The conceptual design was completed end of December 2015 with the issuance of a large amount of engineering files (few thousands. The design of ASTRID intends to cope with GEN IV objectives regarding the new reactor concepts and includes ambitious performances to deal with concerning safety and availability for instance. Thus, numerous technical challenges were faced and successfully managed by AREVA NP during the conceptual design phase dealing with: - Deployment of design process based on System Engineering standards, - Selection of adequate architectures and design justification (at the conceptual level stage) for the various main systems and components: primary circuit, secondary loops, decay heat removal systems, fuel handling and component handling systems, I&C platforms, electrical systems, general layout of nuclear island building etc. Throughout the design progress, AREVA NP experimented new approaches in terms of management of innovations, advanced numerical simulations, management of large CAD models and the related interfaces with the other industrial partners, introduction of Virtual Reality tools to enhance the complexity mastering of the layout. In addition, this paper describes the main technical achievements regarding the NI and main systems or component definition at the end of the conceptual design phase. Future trends for the design of the NI are presented in terms of evolution of the technical configuration and enhancement of the engineering tools.

Speaker: Mr JEAN-MARIE HAMY (AREVA NP)

Material IAEA-CN-245-528_JMH_2.docx

48 Analysis of the Characteristics of the Fast Breeder Reactor with Metallic Fuel

A lot of approaches are considered to increase a marketability of fast breeder reactors producing two products – electricity and exceeding nuclear fuel. To increase a production of exceeding nuclear fuel it is proposed to switch from widely used oxide fuel to carbide, nitride and the densest metallic uranium fuel. In a fabrication chain of the exceeding nuclear fuel a cost of spent nuclear fuel refabrication is also important. From all kinds of nuclear fuel, considered worldwide at the fast breeder reactors’ area, the metallic fuel provide the highest values of the exceeding nuclear fuel production i.e. the highest value of the breeding rate (BR) and the lowest refabrication cost for spent nuclear fuel due to melting technology. But the reactors with metallic fuel have issues which lead to the absence of completed projects and their realizations. The main problem of the safety assurance of such reactors is related to a weak reactivity feedback by fuel temperature. To solve this problem an approach with heterogeneous placement of the fuel at the axial direction is suggested. Layout of the depleted metallic fuel is proposed at the bottom blanket region and at the top blanket region above the sodium cavity to receive high breeding rate. In addition, placement of the oxide fuel with central thin layer made of metallic fuel in the core is proposed to provide sufficient level of temperature feedback. An improvement of this approach with replacement of the oxide fuel from the bottom part of the core by a metallic plutonium fuel is considered at the paper. It is shown by calculations that the suggested approach together with the replacement of the oxide fuel by the metallic depleted uranium fuel at the assemblies of a radial blanket region ensures the high reactor BR with sufficient level of the temperature feedback. The high BR value is provided by using of the metallic fuel in the majority of reactor’s volume. Substantial feedbacks are provided by the utilizing of the oxide fuel at the area of high coolant, fuel and cladding temperatures. At the same time the metallic plutonium fuel is placed at the area of high power density and low temperature of the core components.

Speaker: Mr IURII DROBYSHEV (VNIIAES JSC)

Material Drobyshev_Analysis_of_the_Characteristics_of_the_Fast_Breeder_Reactor_with_Metallic_Fuel_v2.pdf

49 FASTER Test Reactor Preconceptual Design

The FASTER test reactor was designed as part of the U.S. Advanced Demonstration and Test Reactor Options (ADTR) Study in 2015/2016. The ADTR study provided an assessment of advanced reactor technology options and is intended to provide a sound comparative technical context for future decisions concerning these technologies. Point designs for a select number of concepts were commissioned. One of the two test reactor point designs was a sodium-cooled fast test reactor called FASTER. FASTER is a sodium-cooled, metal alloy fueled fast reactor with a core thermal power rating of 300MW. The FASTER plant was designed with extended testing capabilities in mind while trying to keep the reactor plant as simple as possible. The main function of the FASTER plant is to provide high neutron flux irradiation capability for both fast neutron spectrum and thermal neutron spectrum applications. The FASTER reactor plant incorporates an innovative core arrangement that also provides for irradiation testing in closed loops with different working fluids. This paper will describe the design characteristics of the FASTER plant and provide background information on the ADTR study and its objectives.

Speaker: Mr Christopher Grandy (Argonne National Laboratory)

Paper FR2017_FASTER_PAPER_-_Draft_rev4.docx

3.2 Core Disruptive Accident

Conveners: Mr Andrey VOLKOV, Prof. Koji Morita (Kyushu University)

50 Status of severe accident studies at the end of the conceptual design: feedback on mitigation features

The ASTRID reactor developed by the CEA with its industrial partners, will be used for demonstration of the safety and operability, at the industrial scale, of sodium fast reactors of the 4th generation. Among the goals assigned to ASTRID, one is to improve the safety and the reliability of such reactor (compared to previous sodium-cooled fast reactors). Among the innovations promoted in the ASTRID design, a low sodium worth core concept (CFV core) has been developed. By means of various design provisions enhancing the neutron leak in case of sodium draining, the overall sodium void effect of the ASTRID core is near zero and could even be negative. Additionally, mitigation devices should be implemented into the core in order to limit the calorific energy released in the fuel during the secondary phase of the accident. This paper deals with a synthesis of severe accidents studies performed during the second period of the pre-conceptual design stage of the ASTRID project (2013-2015). The main insights of the studies in term of mitigation strategy and of mitigation device design are highlighted in the paper. The core transient behavior has been investigated in case of generalized core melting situations initiated by postulated reactivity insertion ramps (UTOP) and unprotected loss of flow (ULOF). In case of postulated reactivity insertion ramps, the mechanical energy release assumed to be released by molten fuel vapor expansion does not exceed several tenths of megajoule ULOF transient does not lead to energetic power excursions neither thanks to the mitigation provisions and to the core design. Moreover, the ULOF early boiling phase leads to core power decrease. Thus, the primary phase of the accident is not governed by a power excursion. The paper deals with the approach and the presentation of preliminary findings regarding mitigation provisions. Those provisions are investigated by considering a core degraded state representing the end of the transition phase. The scenario possible evolutions from this degraded state provide the following parameters: mass and temperature of molten materials, mass and flow rate of materials relocated on the core catcher and possible ejected material mass above the core. Those parameters are used for the determination of approximate loadings for the primary vessel and for the design of the core catcher. Finally, a methodology and first results dedicated to assess the efficiency of mitigation device design is presented as well as first preliminary results of checking process.

Speaker: Frédéric BERTRAND (CEA Cadarache)

Paper Paper_CEA_410_final.docx

51 Source Term Estimation for Radioactivity Release under Severe Accident Scenarios in Sodium cooled Fast Reactors

Due to the inherent characteristics and robust design of Sodium cooled Fast Reactors (SFR), the core disruptive accident (CDA) is considered a very unlikely event. Nevertheless, to confirm the safety of the reactor, one of the hypothetical scenarios arising from the loss of coolant flow coupled with the failure of the shutdown system, referred to as the Unprotected Loss of Flow (ULOF) accident is postulated to serve as a basis for containment design and severe accident management measures. Determination of the corresponding radioactive source term released into the containment is an important initial condition for the assessment of the adequacy of the containment and subsequently the radiological impact at the site. Estimation of the source term for the sodium cooled fast reactors requires computational tools similar to those developed for the assessment of thermal reactor source term. Towards improving the current state of the art for modeling the in-vessel and in-containment source terms IAEA launched the CRP in which participants from nine countries are doing benchmark simulations for the source term estimation with different models and tools. The technical aspects to be addressed are divided into three main parts. First is the in-vessel source term estimation, consisting of risk important fission product distribution in the fuel pins, their release mechanisms into the coolant and subsequent reaction and transport in the coolant and release to the cover gas. Second is the primary system/containment interface source term estimation consisting of models for the cover gas, sodium ejection and radionuclide chemical composition and distribution in the containment. The third part is the estimation of the fission product evolution within the containment considering sodium burning scenarios, aerosol behavior and physical boundary conditions. Towards this a benchmark SFR model has been defined and developed. The input data required for the simulations have been calculated and boundary conditions identified and specified. The paper presents the problem definition, approach and results obtained from the preliminary modeling. The resulting models are expected to provide a more realistic than existing conservative estimates and would further help to identify areas for experimental investigations through sensitivity and uncertainty analysis of the improved integrated models.

Speaker: Dr Vladimir Kriventsev (IAEA)

Material fr17-source-term_revised_30_04_2017.docx

52 Advances in the Development of the SAS4A Code Metallic Fuel Models for the Analysis of PGSFR Postulated Severe Accidents

The SAS4A safety analysis code, originally developed for the analysis of postulated Severe Accidents in Oxide Fuel Sodium Fast Reactors (SFR), has been significantly extended to allow the mechanistic analysis of severe accidents in Metallic Fuel SFRs. The new SAS4A models track the evolution and relocation of multiple fuel and cladding components during the pre-transient irradiation and during the postulated accident, allowing a significantly more accurate description of the local fuel and cladding composition. The local fuel composition determines the fuel thermo-physical properties, such as freezing and melting temperatures, which in turn affect the fuel relocation behavior and ultimately the core reactivity and power history during the postulated accident. The models describing the fission gas behavior, fuel-cladding interaction, clad wastage formation and cladding failure models have been also significantly enhanced. The paper provides on overview of the SAS4A key metal fuel models emphasizing their new capabilities, and presents results of SAS4A whole core analyses for selected PGSFR postulated severe accidents.

Speaker: Mr Seok-Hun Kang (KAERI)

Paper FR17_Paper_#56_v15_040517_revised_sent.pdf

53 Computational modeling of flow blockage in fuel subassemblies and molten material relocation in sodium cooled fast reactors

The heat generating fuel pins in Sodium cooled Fast Reactors (SFR) are arranged in a tightly packed triangular pitch within a hexagonal sheath forming a fuel subassembly (SA). Due to the compact design, formation of local flow blockage inside the SA is possible. Such blockages are expected to grow gradually and the core monitoring thermocouples which are located at the top of the SA are capable of detecting these blockages at their infancy. But, large size blockages may not be detected by the thermocouples due to low velocity of sodium issuing from the blocked subassembly eventually leading to core damage. The extent of damage propagation before reactor shuts down depends on the size of the blockage and its rate of growth. The thermal hydraulics phenomena involved during damage progression are very complex, involving phase change heat transfer with moving solid-liquid interfaces. To investigate (i) the sequence of damage progression, (ii) possibility of Total Instantaneous Blockage (TIB) detection by online monitoring of the sodium outlet temperature from the neighboring SA and iii) determination of number of SA that are likely to get damaged severely before reactor shutdown etc. a transient a enthalpy based thermal hydraulic model has been developed. The transient model considering multi-material and multi-phase flow features adopts an explicit finite difference method employing Voller’s algorithm for interface tracking. The model has been validated against published benchmark data. The SA that are likely to get damaged during a TIB event is determined to be seven. During a CDA, a significant fraction of the hot molten fuel moves downwards and gets relocated to the core catcher. The core catcher design requires prior knowledge of core-melt relocation time which is the time taken for the molten fuel to reach the lower plenum from the active core region. The initial thermal load on the core catcher is primarily dictated by the core melt relocation time. By mathematical models, upper and lower bounds for core-melt relocation time for postulated accident conditions of Protected Loss of Heat Sink (PLOHS) accident and Unprotected Loss of Flow Accident (ULOFA) have been determined. The potential of a multi layer core catcher in handling the debris generated from a whole core melt down accident has been assessed by CFD studies.

Speaker: Dr Velusamy Karuppanna Gounder (IGCAR)

Material Velu_Paper_-_Ver_-_2.doc

54 Quantitative Evaluation of the Post Disassembly Energetics of a Hypothetical Core Disruptive Accident in a Sodium Cooled Fast Reactor

The analyses of Hypothetical Core Disruptive Accidents (HCDAs) play a fundamental role in the safety assessment of Sodium Fast Reactors (SFRs). The accident sequence is subdivided into different phases suggested by dominant key phenomena. The Initiation Phase (IP) describes the fatal deviation from nominal operation until the failure of single sub-assemblies (SAs), while the subsequent Transition Phase (TP) considers possible damage propagation up to the formation of a large fuel/steel pool. During the TP, a coherent movement of the liquid pool may result in a more compact fuel arrangement leading to recriticality events with consequent upward discharge of the pressurized hot fuel/steel mixture. The power peaks are considered as starting points for the Post-Disassembly Expansion Phase (PDE). The sodium vapor rapidly produced by Fuel-Coolant Interaction (FCI) in the upper plenum displaces and accelerates the surrounding liquid sodium. As a result, a significant mechanical energy may be released acting as a load on structures /vessel. The identification and evaluation of the main phenomena and event paths enhancing or mitigating the mechanical work potential during the PDE is essential to give evidence on the vessel/structures integrity with important design clues for the development of future SFRs. The present paper deals with PDE phenomenology and includes an overview of the quantitative evaluation of the work potential during the PDE of an Unprotected Loss of Flow (ULOF) on the basis of mechanistic SIMMER simulations. For assessing the important determining factors a large number of parametric analyses have been conducted at KIT for an SFR model case and, additionally, KIT simulations performed in previous years have been studied. A wide range of initial conditions and modelling options that may strongly impact the mechanical work potential has been investigated and are integrated in this work, i.e. different liquid fuel mass and temperature conditions, different steel contents in the melt pool, different structure conditions affecting the melt discharge into the sodium plenum, and different driving vapor pressures. The large amount of results has been also employed in the framework of the application of the Phenomenological Relationship Diagram (PRD) to perform a probabilistic evaluation of the work potential of an ULOF/PDE in a sodium small- to medium-sized reactor (SMR, 300 MWe). The present study has been conducted under the research contract between the KIT and the Regulatory Standard and Research Department of National Regulation Authority in Japan.

Speakers: Dr Barbara Vezzoni (Karlsruhe Institute of Technology (KIT)), Mr MIchael Flad (Karlsruhe Institute of Technology)

Material FR17_483.docx

55 An assessment of transient over-power accident in the PGSFR

KAERI (Korea Atomic Energy Research Institute) has been developing a preliminary specific design of the PGSFR (Prototype Gen-IV Sodium-cooled Fast Reactor), which is a pool-type sodium cooled fast reactor with a thermal power of 392.2 MW. The PGSFR has an inherent safety characteristic owing to the design to have a negative power reactivity coefficient during all operation modes and it has a passive safety characteristic due to the design of a passive decay heat removal circuit. For an evaluation of the safety features of the PGSFR, a sensitivity analysis has been performed for TOP(Transient Over-Power) which is one of most important DBEs in the PGSFR using MARS-LMR code. MARS-LMR contains the sodium property table including dynamic properties, heat transfer correlations for the liquid metal, and the models describing the flow resistance by wire-wrap spacer in the core, which shows a good agreement with the experimental data conducted in the EBR-II plant and the appropriateness of the models related to liquid metal reactor. For a sensitivity analysis, some design variables are applied to be conservative. An effect of uncertainties is evaluated on a Doppler reactivity and a sodium density. Conservative assumptions are applied to the analysis of the plant responses during the postulated DBAs, which are 102 % of power condition with ANS-79 decay power model, 5.0 seconds delay in opening of AHX and FHX dampers, and loss of off-site power (LOOP) is taken into account. Additionally, one PDHRS and one ADHRS are available in accordance with a single failure criterion and maintenance. As a result, the preliminary specific design PGSFR meets safety acceptance criteria with a sufficient margin during the TOP event and keep accidents from deteriorating into more severe accidents.

Speaker: Dr KWI LIM LEE (KAERI)

Paper FR17_KLLEE_DRAFT_2.pdf

4.1 Fuel Cycle Overview

Conveners: Prof. Christophe POINSSOT (French Nuclear and Alternative Energy Commission), Dr Vitaliy Vidanov (INSTITUTION «ITC «PRORYV» PROJECT»)

56 1992-2017: 25 years of success story for the Development of Minor Actinides Partitioning Processes

In the framework of the successive 1991 and 2006 Waste Management Act, French government supported a very significant R&D program on partitioning and transmutation in fast reactors of minor actinides (MA). This program aimed to study potential solutions for still minimizing the quantity and the hazardousness of final waste, by MA recycling. Indeed, MA recycling can reduce the heat load and the half-life of most of the waste to be buried to a couple of hundred years, overcoming the concerns of the public related to the long-life of the waste. Over the 20 years of development, different types of strategies were studied, from the early multi-stage DIAMEX-SANEX processes to the most recent innovative SANEX, from the grouped extraction of MA thanks to the GANEX process to the most recent sole Americium recycling thanks to the EXAm process. These developments were supported by a robust and long-standing approach allowing successively to screen the potential extractants, to quantify their extractive properties and develop a relevant chemical model to simulate it and to address their hydrolysis and radiolysis resistance. Finally, all these processes were qualified tested on a few kg of spent nuclear fuel within the Atalante CBP facility. This wide research program allows France to get in hand a flexible portfolio of MA recycling processes that could be implemented after industrial upscaling. More recently, CEA initiated a demonstration experiment, the so-called integral experiment, which aims to re-irradiate in a Material Testing Reactor some fuel pellet manufactured from recycled UAm. Most recent results on these key experiments will be presented. Finally, several European Research Projects were funded in parallel by the European Commission and allow studying alternative separation processes. A general overview of this 20 years of successful and innovative research history will be synthesised in this presentation.

Speaker: Prof. Christophe POINSSOT (French Nuclear and Alternative Energy Commission)

Material Full_paper_Poinssot_et_al_FR17_MA_transmutation_v3.docx

57 Comparison of fast reactors performance in the closed U-Pu and Th-U cycle

Breeding as well as burning capabilities of a reactor operated in the U-Pu or Th-U closed fuel cycle can be estimated from its equilibrium cycle parameters. In this study the equilibrium parameters were simulated for 8 selected fast reactors and both U-Pu and Th-U closed fuel cycles. For simplicity, the fission products were neglected and the reactors were represented only by infinite lattice. It was found that the mass flow is stabilized in equilibrium closed cycle. The fuel composition does not differ between two consecutive cycles and determines the excess reactivity. This reactivity can serve as a measure for breeding or burning capabilities of each reactor. For a breeder reactor it should be high enough to accommodate the expected fission products and the presumed neutron leakage. The study provided insight for the differences between the 8 fast reactors and also between the U-Pu and Th-U closed fuel cycles.

Speaker: Dr Jiri Krepel (Paul Scherrer Institut)

Paper IAEA-CN-245-52_rew.pdf

58 Reprocessing of fast reactors mixed U-Pu used nuclear fuel: studies and industrial test

Mixed U-Pu used fuel of fast reactors has a high Pu content, high burn-up (50 GW/days*ton and more), and short cooling time (no more than 3 years) as compared with the thermal reactors used fuel. Combined (pyro + hydro) and hydrometalurgy reprocessing technologies are developed in Russian Federation for the close nuclear fuel cycle. These technologies provide reprocessing of used fuel with 1-3 years cooling time, 10-15 % of Pu content and burn-up up to 100 GW/days*ton. The aim of reprocessing is production of purified mixture of actinides oxides. The purification of actinides from fission products should be around 10.000.000. The dry reprocessing technology based on pyroelectrochemical refining is under development as well. This technology should be suitable for reprocessing of fast reactors used fuel with burn-up more than 100 GW/days*ton and 1 year cooling time The main results of the studies are being discussed in a current paper. Within studying the analysis of products and operations of reprocessing 8 MOX irradiated assemblers from BN-600 reactor at RT1 plant with burn-up from 73 up to 89 GW/days*ton were performed. The technology steps were made under standard conditions. The increasing of Pu loses during MOX used fuel of BN-600 as compared with thermal reactor used fuel reprocessing was not found. The design of reprocessing facility for mixed U-Pu nitride fuel started at 2015. This facility is planed to be built at Siberian Chemical Combine as a part of experimental and demonstrating energocomplex with reactor BREST–ED-300.

Speaker: Andrei Shadrin (Bochvar Institute)

Material FR17_76_-_Corrected.pdf

59 Overview of the Nuclear Energy Agency Scientific Activities on Advanced Fuel Cycles

Activities related to the fast reactors systems at the Nuclear Energy Agency (NEA) mainly focus on scientific research and technology development needs and are carried out within the Nuclear Science Division. In particular, projects related to the advanced nuclear systems and fuel cycles are carried out through the Working Party of Scientific Issues of the Fuel Cycle (WPFC) and its five related experts groups covering all scientific aspects of the fuel cycle from front to back-end. Ongoing projects on advanced systems include fuel cycle scenarios, fuels, materials, physics and chemical separations. Members of the expert groups cooperate to share recent research advancements at an international level and help identify gaps and needs in the field. Current activities focus on nuclear systems in particular on the challenges associated with the adoption of new materials and fuels such as for example cladding materials, fuels containing minor actinides, or the use of liquid metal as coolants. The expert group on Innovative Fuels is conducting joint and comparative studies to support the development of innovative fuels in particular minor actinide bearing fuels. Ongoing work involves a benchmark study on fuel performance codes and experiments The new expert group on Liquid Metal Technology includes projects on liquid Na, lead or lead-bismuth) to support: (1) the development of construction codes used for design (design rules), (2) identify the key technical issues for licensing, (3) give recommendations for operation, inspection and handling. The report on the effects of uncertainties of input parameters prepared by the expert group on Advanced Fuel Cycle Scenarios will be published soon. The purpose of this study was to identify sources of uncertainty and use sensitivity studies to assess their impacts on system level results. Members of the expert group are currently working on a scenario study on transuranic management in order to assess the quantity materials contained in spent fuel that could be burnt using various fast burner fleets. In addition, a benchmark study on dose rate calculation for irradiated fuel assembly is being undertaken. At the back-end of the fuel cycle, separation technologies (aqueous and pyrochemical) are being assessed by the Expert Group on Fuel recycling Chemistry and a state-of-the art report on minor actinide separation chemistry is being finalised.

Speaker: Dr Stéphanie Cornet (Nuclear Energy Agency)

Material S._Cornet-FR17-IAEA-CN-245-527.docx

60 Assessment of the anticipated improvement of the environmental footprint of future nuclear energy systems

Environmental issues are nowadays a growing concern within most of the public opinion. It is therefore mandatory to propose relevant and qualified assessment of the overall environmental footprint of the different types of energy sources which are envisaged to be implemented. This question is specifically important for nuclear energy which suffers from a poor image in the public opinion due to the recent Fukushima accident. In this context, we developed a Life Cycle Assessment (LCA) tool, referred to as NELCAS, based on the current French nuclear energy system. Thanks to the Nuclear Safety and Transparency annual reports, detailed quantitative data were available for each of the fuel cycle plants. The whole fuel cycle from ore-mining to geological repository was considered as well as data for construction, deconstruction of any plants as well as the contribution of the transport. All the matter and energy fluxes were considered and normalised versus the electric production. Key environmental indicators, such as land use, water withdrawal and consumption, gaseous release, acidification, eutrophication, waste production … as well as potential impact indicators were hence assessed and validated with comparison with the few existing LCA results. This model was hence used to assess the respective figure of merits of the different generation of reactors and fuel cycles. In particular, it demonstrates that actinides recycling has a strong beneficial effect on the overall footprint due to the relative high impact of the front-end activities, specifically the ore mining. In the framework of a joint CEA-EDF-AREVA group, reference deployment scenario for the 4th generation reactors were developed for the French case based on both technical and economic considerations. The NELCAS tool was therefore used to assess the impact on the overall environmental footprint of this reference scenario.

Speaker: Prof. Christophe POINSSOT (French Nuclear and Alternative Energy Commission)

Material Full_paper_Poinssot_et_al_FR17_NELCAS_v3.docx

61 Concurrent Trends in Indian Fast Reactor Fuel Reprocessing Programme

The Indian fast reactor program, which began with the construction of the mixed carbide fuelled Fast Breeder Test Reactor (FBTR) at Kalpakkam, has reached a level of maturity with three decades of operating experience and is stepping into the realm of commercial operations with the construction of Prototype Fast Breeder Reactor (PFBR). The necessary technology for closing the fuel cycle, which is vital for the success of the fast reactor program, has been concurrently developed facing the unique challenges posed by the fast reactor fuel. CORAL (COmpact Reprocessing of Advanced fuels in Lead cells), a pilot facility, has been operating successfully since the year 2003, reprocessing the spent fuel discharged from FBTR, with burnup upto 155 GWd/t and very short cooling periods as low as 18 days. This facility has served the purpose of validation of the process as well as the equipments that were developed for fast reactor fuel reprocessing. Operating the facility has given valuable feedback for the Demonstration fast reactor Fuel Reprocessing Plant (DFRP), which will be a regular reprocessing plant for FBTR and also serve as a demonstration for the reprocessing of mixed oxide fuel from PFBR. The CORAL experience was also vital in designing the Fuel Reprocessing Plant (FRP) of Fast Reactor Fuel Cycle Facility (FRFCF), which would be a regular reprocessing plant for spent fuel discharged from PFBR. Considerable experiences gained and feedback obtained in design and operation of the reprocessing facilities provided vital inputs for achieving the required robustness in the fast reactor fuel reprocessing program. With the construction of FRFCF, the Indian fast reactor fuel reprocessing program would step in to the realms of commercial reprocessing. R&D efforts are also concurrently under progress to develop efficient processes and equipments, aqueous processes for reprocessing of the U-Pu-Zr metallic fuel as a fall-back for the pyro-chemical process.

Speaker: Mr KODANDARAMAN J (INDIRA GANDHI CENTRE FOR ATOMIC RESEARCH)

Material FR_17_Paper_IAEA-CN-245-297.docx

5.2 Advanced Fast Reactor Fuel Development II

Conveners: Dr Mikhail Veshchunov (IAEA), Dr Victor BLANC (French Atomic Energy Commission)

62 The IAEA Coordinated Research Project on Sodium Properties and Safe Operation of Experimental Facilities in Support of the Development and Deployment of Sodium-cooled Fast Reactors (NAPRO)

The International Atomic Energy Agency (IAEA) recently established a Coordinated Research Project (CRP) on “Sodium Properties and Safe Operation of Experimental Facilities in Support of the Development and Deployment of Sodium-cooled Fast Reactors - NAPRO”, to be carried out in the period 2013 – 2017. Eleven institutions from ten Member States participate in this CRP. The complete scope of this CRP is covered by three work packages. A specific work package (WP1), under the coordination of the Argonne National Laboratory (USA), is focused on the compilation and expert assessment of data sets of Na physical and chemical properties, as well as correlations for pressure drops and heat transfer in Na facilities. Identification of gaps in the data sets, and recommendations for their closure are included. A second work package (WP2), under the coordination of the Institute of Physics and Power Engineering – IPPE (Russian Federation), addresses the compilation, evaluation and development of best practices and guidelines for the design, operation and maintenance of Na facilities. Finally, Work Package 3 (WP3), coordinated by the French Alternative Energies and Atomic Energy Commission (CEA), concentrates in the compilation and development of guidelines and rules for the safe operation of Na facilities, including, among others, the prevention, detection and mitigation of Na leaks and fires. This work presents an overview of the compiled data bases and correlations of WP1, including recommendations for their use, as well as a summary of the guidelines and rules evaluated and developed in WP2 and WP3.

Speakers: Mr E. Vázquez (National Atomic Energy Commission of Argentina (CNEA)), Mr Osvaldo Azpitarte (National Atomic Energy Commission of Argentina (CNEA))

Material NAPRO-FR17_.docx

63 Effects of Oxygen Partial Pressure During Sintering at Laboratory and Industrial Scales on FR MOX Fuels

In order to prepare the industrial deployment of Sodium Fast Reactor (SFR), a French prototype is envisaged for 2025 (ASTRID: Advanced Sodium Technological Reactor for Innovative Demonstration). Its MOX (mixed oxide) fuels will be produced by a new industrial facility, currently under development and named AFC (for “Atelier de Fabrication des Coeurs”, core fabrication facility). The fabrication process of MOX fuel is based on powder metallurgy processes. The UO2 and PuO2 mixture is pelletized and then sintered at about 1700°C under reducing atmosphere of Ar/4%H2/H2O. Fuel has to be in compliance with specifications. In particular, the O/M (atomic oxygen to metal ratio) has to be hypostochiometric and close to 1.97 and the microstructure has to be dense, around 95 %ThD and free of cracks. The O/M and microstructure can affect numerous properties of the fuel during operation including thermal conductivity, mechanical properties and fuel-cladding interactions. To comply with these specifications, better knowledge of sintering at laboratory and industrial scale is needed. An original analysis method has been therefore developed for a better understanding of the O/M ratio evolution and of densification mechanisms during the sintering step. By coupling a dilatometer with an oxygen zirconia probe, it is possible to identify the different redox phenomena and to plot the evolution of the O/M of the oxides versus time during the densification process. This innovative method helps overcoming the obstacles in reaching the thermodynamic equilibrium between gas and fuel. Whereas it was difficult to predict a precise final O/M, this new method produces the expected ratio every time. This paper highlights the different final O/M values and microstructure, particularly in terms of microcracking, obtained during sintering in a continuous industrial or laboratory kiln. The impact of the evolution of moisture content in the gas is explained. Based on these results, recommendations can be made about the sintering atmosphere to improve industrial cycles and optimize fuel characteristics in order to obtain an O/M as close as possible to the target value and the right microstructure.

Speaker: Mr Stephane Vaudez (CEA)

Material Proceeding_FR17_Track_5.8_fuels_SV.pdf

64 Fuel Melting Margin Assessment of Fast Reactor Oxide Fuel Pins using a Statistical Approach

In the framework of the Basic Design of the Advanced Sodium Technological Reactor for Industrial Demonstration (ASTRID) project, the design margins have to be defined with accuracy. The design criterion considered here is the fuel melting margin during nominal operation conditions, which is given by the melting probability. Oxide fuel temperature and melting temperature being calculated with the CEA fuel performance code, GERMINAL, results could be depend on parameters from manufacturing processes, irradiation conditions and fuel behavior laws. The aim of this paper is to take into account uncertainties associated to these parameters in the melting margin evaluation and to quantify the sensitivity to these parameters. In a first approach, GERMINAL calculations of the temperature distribution given by analytic method is compared with direct Monte-Carlo sampling and with a reliability dedicated method by using the CEA uncertainty and sensitivity simulation platform, URANIE. First sensitivity analysis shows that linear heat rate, fuel stoichiometry and fuel clad gap are the first order parameters. Afterwards, in a detailed approach, pellet geometrical defects in fuel pellets are taken into account using a 3D finite element model based on CEA LICOS code. The maximum temperature being considerably dependent of these defects, a simple meta-model is built, and is linked with GERMINAL in order to build an artificial neural network. Using this global meta-model, first and second order reliability methods are finally compared with a large number of direct Monte-Carlo simulations in order to determine the fuel melting probability.

Speaker: Dr Victor BLANC (CEA)

Paper FR17_FuelMeltingMarginAssessmentUsingAStatisticalApproach_Paper333.pdf

65 New catalog on (U,Pu)O2 properties for fast reactors and first measurements on irradiated and non-irradiated fuels within the ESNII+ project

In order to develop the fast neutron systems, three prototypes of the Sodium Fast Reactor, the Gas-cooled Fast Reactor and a heavy liquid metal cooled Accelerator Driven System are studied in Europe: ASTRID (SFR prototype), ALLEGRO (GFR prototype) and MYRRHA (LBE-cooled ADS system related to the ALFRED LFR-demonstrator). The ESNII+ project with its workpackage 7-FUEL SAFETY aims to provide a set of oxide fuel properties needed for the fuel element design of each prototype. The improvement of fuel properties will also reduce uncertainties in safety behavior evaluations, in nominal conditions as well as during transients and will be achieved by the update of the European catalog on the MOX fuel properties. The uncertainties on the fuel properties have to be rigorously determined; the two main driver criteria for fuel element evaluation are the margin to melt for the fuel and the risk of clad failure. Property measurements are done on existing fresh and irradiated fuel samples, identified to cover the fuel characteristics for ESNII prototypes. The review of the state-of-the art has shown that the knowledge on the thermal conductivity of irradiated FBR MOX is currently very limited. Only one publication is available providing surprising experimental results: no degradation of thermal conductivity with burn-up was observed. The data and models available in the literature were reviewed and new experimental results are obtained in order to develop an updated recommendation. Fresh and irradiated fast reactor fuel was characterised and its thermal diffusivity was measured. The irradiated fuel has an average burn-up of 13.4 at.% and the thermal diffusivity was measured in 3 radial positions: 0.6 mm, 1 mm and 1.4 mm from the pellet cladding. The thermal diffusivity increases from the pellet periphery to the pellet centre and is significantly higher than for LWR UO2 or MOX fuels with similar burn-up. The impact of the main mechanisms is investigated in depth: radiation damage concentration as a function of the irradiation parameters, effect of the plutonium content, of microstructure, of fission gas atoms, of fission products and O/M. A new correlation for the conductivity is developed on the basis of the phenomena specific to FBR fuel: high irradiation temperature, restructuring, extensive fission gas release, diffusion of plutonium and fission products.

Speaker: Dr Kamil Tucek (European Commission, Joint Research Centre)

Material ESNII+_Contribution_template_893.784304819_c126.docx

66 Fission product and swelling behaviour in FBTR mixed carbide fuel

The advantages of a fast reactor, especially one that uses Uranium-Plutonium Carbide as its fuel is well documented. Irradiation performance assessment of carbide fuels began with experimental irradiations in EBR-II, FFTF, HFR, Rapsodie and Phenix etc. India has the extensive experience with this type of fuel at the Fast Breeder Test Reactor at Kalpakkam that has been operating for over 25 years. The fuel has attained a peak burn-up of 155 GWd/t at linear heat rating of 400 W/cm, in a large number of fuel pins. Comprehensive post-irradiation examinations (PIE) at various stages up to this high burn-up have yielded a wealth of information on behaviour of mixed-carbide fuel under steady state operations. In this paper, selected recent results on fission product migration, gas release, fuel swelling behaviour and microstructural evolution of the mixed-carbide fuel will be presented. Results of the PIE towards analysing the cause of failure in a fuel pin are also discussed. Axial distribution of fission products such as 137Cs and 106Ru in the fuel pins was assessed by gamma scanning. A steep increase in the fuel stack length was observed beyond 100GWd/t burn-up indicating onset of FCMI. Fission gas release in fuel pins after a burn-up of 155 GWd/t indicated relatively low gas release of 16%. Systematic change in the fuel-clad gap and cracking pattern was observed with increasing burn-up. Fabrication porosities present in the fuel was found to decrease with increasing burn-up indicating that the fuel swelling is being accommodated in the porosities. Caesium axial distribution in the failed pin and some of the fuel pins adjacent to it in a failed sub-assembly irradiated at a lower linear power of 260 W/cm. Gas release behaviour showed contrasting trends with higher gas release in the pins adjacent to the failed pin and lower gas release in pins located farther away from the failed pin. The micrograph of the failed pin cross-section at the location of failure showed highly densified fuel region. Asymmetric circumferential cracking was observed, indicating non-uniform temperature around the pin resulting from the diameter increase and local bowing in the fuel pins. Clad carburisation was not observed. The performance assessment through PIE has provided valuable insights into the behaviour of the mixed carbide fuel and cause of failure in a fuel pin.

Speaker: Mr Suresh Kumar Kannankara Vasudevan (IGCAR, Dept. of Atomic Energy, India)

Material FR_17_paper_-_Swelling_and_FP_behaviour.docx

67 Analysis of experimental data on fission gas release and swelling in mononitride fuel irradiated in BR-10 reactor

Uranium mononitride fuel was used in the fourth and fifth BR-10 reactor core loadings. The total number of irradiated fuel pins was 1250 (660 and 590 fuel pins in fuel loadings IV and V respectively). Most fuel pins were irradiated up to design burnup (8%) without cladding failure. In addition to standard FAs, some experimental FAs were irradiated in BR-10 reactor. The post irradiation examination (PIE) of 8 standard and 3 experimental fuel assemblies (FA) was done in the IPPE hot lab. The paper presents the results of study of fission gas release and nitride fuel swelling in standard and experimental fuel pins irradiated in BR-10 reactor. These two phenomena have a significant impact on cladding stress level and therefore on the fuel life time. Substantial fission gas release from BR-10 nitride fuel starts at a burnup of more than 3at%. Irradiation temperature increase and fuel density decrease both lead to increase of gas release rate. N14(n,α)B11 nuclear reaction in nitride fuel causes formation of quite big amounts of helium. This fact should be taken into account in computer codes used for nitride irradiation behavior modeling. The paper presents the measured nitride swelling rate values in the temperature range from 760 to 1115 C. Increase of fuel temperature leads to increase of fuel swelling rate.

Speaker: Mr Sergey Porollo (IPPE)

Material IAER-CN-245-81.docx

6.11 IAEA Benchmark on EBR-II Shutdown Heat Removal Tests

Conveners: Dr Ethan Bates (TerraPower LLC.), Dr Vladimir Kriventsev (IAEA)

68 IAEA’s Coordinated Research Project on EBR-II Shutdown Heat Removal Tests: An Overview

A Coordinated Research Project (CRP) on “Benchmark analysis of EBR-II Shutdown Heat Removal Tests (SHRT)” was launched by the International Atomic Energy Agency (IAEA) in 2012. A series of transient tests were conducted on the EBR-II reactor at Argonne National Laboratory (ANL) to improve the understanding of thermal hydraulics and neutronics of fast reactors. Shutdown heat removal tests conducted in 1984 and 1986 demonstrated mechanisms by which fast reactors can survive severe accident initiators with no core damage. Two SHRT tests, SHRT-17 representing Protected Loss of Flow (PLOF) transient and SHRT-45R representing Unprotected Loss of Flow (ULOF) transients have been studied in the IAEA CRP. The objectives of the CRP were to improve design and simulation capabilities in fast reactor thermal hydraulics, neutronics and safety analyses through benchmark analysis of these two important tests. At the first stage of the benchmark, ANL provided the input data on EBR-II geometry, as well as initial and boundary conditions for the SHRT-17 and SHRT-45R tests to perform “blind” calculations. At the second stage, ANL released the experimental observations and participants had the chance to discuss the difference and refine the models. Nineteen organizations from eleven countries participated in the CRP making it one of the largest CRP coordinated by the IAEA fast reactor team. The papers provides a general CRP overview while the companion papers presented both on this session and at the poster session give the details of the EBR-II reactor design, describe the shutdown heat removal tests, the benchmark setup, results of numerical simulations, and the detailed discussion on this CRP.

Speaker: Dr Vladimir Kriventsev (IAEA)

Material FR17-IAEA-CN-245-361-EBR-II.docx

69 EBR-II Passive Safety Demonstration Tests Benchmark Analyses

In 2012, the International Atomic Energy Agency (IAEA) established a coordinated research project (CRP) on EBR-II Shutdown Heat Removal Tests (SHRT). The objectives of the CRP, which concluded in 2016, were to improve design and simulation capabilities in fast reactor neutronics, thermal hydraulics, plant dynamics, and safety analyses through benchmark analysis of two landmark tests from the EBR-II SHRT program: SHRT-17 and SHRT-45R, the most severe protected and unprotected loss-of-flow transients, respectively. Nineteen organizations representing eleven countries participated in the CRP. The benchmark was performed in two phases. During phase 1, participants had no access to recorded data from either test. Once all phase 1 calculations were completed in February 2014, phase 2 was initiated with participants receiving the experimental data. This paper will summarize the SHRT-17 and SHRT-45R tests and the benchmark specifications for each test that were developed for the CRP. An overview of the major simulation results from all CRP participants will be presented, and lessons learned from the CRP will be discussed. Two companion papers in this conference will cover in detail the comprehensive analysis results for SHRT-17 and SHRT-45R, respectively. Two additional conference papers will address 1) the optional neutronics benchmark that was conducted for the SHRT-45R test and 2) the computational fluid dynamics and subchannel models that were used by some of the CRP participants. **Note: EBR-II Benchmarks CRP Invited Session**

Speaker: Ms Laural Briggs (Argonne National Laboratory)

Material IAEA-CN-245-4_FR-17.docx

70 IAEA NEUTRONICS BENCHMARK FOR EBR-II SHRT-45R

A Coordinated Research Project (CRP) initiated by the International Atomic Energy Agency (IAEA) aimed to benchmark Shutdown Heat Removal Tests (SHRT) conducted at Experimental Breeder Reactor II (EBR-II). Two SHRT tests (SHRT-17 and SHRT-45R) representative, respectively of Protected Loss of Flow (PLOF) and Unprotected Loss of Flow (ULOF) transients were considered. The SHRT-45R benchmark included both safety analyses and an optional neutronics benchmark for SHRT-45R. Only the activities carried out for the neutronics benchmark are described in this paper. The objective of the neutronics benchmark was to provide reactivity feedback coefficients for the thermal hydraulic analysis of SHRT-45R. Several institutes participated in this benchmark, including: Karlsruhe Institute of Technology (KIT), University of Fukui, Paul Scherrer Institute (PSI), and Argonne National Laboratory. The parameters compared code-to-code were keff, βeff, reactivity feedback coefficients (axial, radial and control rod expansion, sodium density, and Doppler) and the power distribution in each subassembly, including fission and gamma heat. The fission and decay heat power for 15 minutes after a postulated scram at the beginning of SHRT-45R were also calculated. Several stochastic and deterministic codes were used: MC2-3/TWODANT, DIF3D, VARI3D, and PERSENT by Argonne, SERPENT by PSI, and the ECCO/ERANOS codes by the University of Fukui and by KIT. KIT also used the PARTISN code. Results obtained for keff and βeff were in good agreement (1.2% maximum difference) among the participants. The reactivity feedback coefficients initially showed a large spread that was reduced by establishing consistency among the definitions used by the participants. However, some spread remains, partially due to the different linear thermal expansion coefficients used in converting the change in reactivity (pcm) to change in reactivity per change in temperature (pcm/K), and will be discussed in the full paper. Differences due to core modeling options (detailed fuel pin modeling vs. homogenized subassembly modeling) and neutron cross-section preparation were also analyzed. Differences among the calculated power distributions were large (up to 80%) in the non-fueled subassemblies, where photon heating dominates, while differences were less than 5% in the fueled subassemblies. No recorded data are available for the detailed power distribution.

Speakers: Ms Barbara Vezzoni (Karlsruhe Institute of Technology (Germany)), Mr Wilfred Van Rooijen (Research Institute of Nuclear Engineering (RINE), University of Fukui, (Japan))

Material IAEA_NEUTRONICS_BENCHMARK_FOR_EBR-II_SHRT-45R.docx

71 Conclusions of a Benchmark Study on the EBR-II SHRT-45R Experiment

This paper presents the conclusions of a 4 year benchmark study on the simulation of the EBR-II SHRT-45R experiment. The SHRT-45R experiment was an unprotected loss of flow transient where pump dynamics, natural convection, core and mechanical behavior played a large role in passively and safely limiting the power and temperature rise of the fuel assemblies. Participants from China, Germany, Japan, Korea, Netherlands, Russia and the U.S. presented transient reactor system modeling results for a variety of instrumented parameters, including core outlet temperatures, pump flow rates, and fission power. Detailed pin-level experimental data of the instrumented XX10 (non-fueled) and XX09 (fueled) subassemblies were also assessed. Code-to-code comparisons were made for other non-measured parameters, such as decay heat and peak cladding and fuel temperatures. A small subset of participants presented code predictions of the negative expansion feedbacks (coolant, axial, radial) and Doppler feedback inherent to the EBR-II core. The final meeting held April 2016 in Vienna summarized key findings and sensitivity studies completed after the experimental data was released and the benchmark study converted from blind to open. The fidelity and methodology of core and system models varied greatly between participants. It was found that accurate simulation of the pump coastdown, system pressure drop, and coolant and radial expansion feedbacks strongly influenced the fission power and temperatures in the core during the transient. Relatively simple models for radial expansion were sufficient to capture the behavior during the transient, in part due to the simpler mechanical dynamics of EBR-II’s core and the applicability of the point-kinetics model. Reactor core outlet (Z-pipe and IHX) temperatures were somewhat difficult to match due to the high fidelity required to capture the temperature at the specific thermocouple location. Faulty subassembly flow meter data from XX09 and XX10 prevented a more accurate study of the core flow redistribution occurring during the pump coastdown. Uncertainties and variations in heat transfer and subassembly pressure drop correlations, and fuel expansion assumptions were found to have little effect on the prediction of fission power and temperature. Overall, the benchmark of the SHRT-45R was a valuable exercise that facilitated the development of state-of-the-art models for sodium fast reactor system and neutronic reactivity feedbacks. “Note: EBR-II Benchmarks Invited Session”.

Speakers: Ms Dalin Zhang (Xi'an Jiaotong University), Dr Ethan Bates (TerraPower)

Material Bates_shrt45r_paper_FR17_1_24_final_draft.docx

72 Thermal Hydraulic Investigation of EBR-II Instrumented Subassemblies during SHRT-17 and SHRT-45R Tests

Experimental Breeder Reactor (EBR-II) was a U-Pu-Zr metal-alloy fuelled liquid-metal-cooled fast reactor, extensively used for conducting safety experiments. EBR-II was heavily instrumented to measure sodium flows and temperatures at various locations in the primary circuit including the temperature distribution inside the subassemblies (SA). Several transient tests were conducted on the reactor to improve the understanding of thermal hydraulics and neutronics of fast reactors. The shutdown heat removal tests (SHRT-17 & SHRT-45R) conducted in 1984 and 1986 demonstrated mechanisms by which fast reactors can survive severe accident initiators with no core damage. In order to utilize the data recorded during these tests and facilitate computer code validation, IAEA has initiated a coordinated Research Project (CRP) wherein 19 organizations representing eleven countries participated. Several participants simulated parts of the primary heat transport system using CFD codes. Amongst theses studies, the sub-channel/CFD analysis of the instrumented SA (XX09 & XX10) are very important. XX09 was a 61-pin (59 fueled) SA with helically wound spacer wire over each pin and XX10 was a 19-pin non-fuelled SA without spacer wire. The instrumented SA are additionally cooled by a small amount of thimble flow around the SA. These SA were instrumented with wire wrap thermocouples, flow meters (below the core) and thermocouples at the SA inlet and outlet. Participants used sub-channel analysis codes and CFD codes for predicting the thermocouple temperatures at various locations. It is seen that the CFD studies are computationally intensive and transient studies could not be continued for long duration. The core top and SA top temperatures predicted by the sub-channel analysis codes and CFD codes are in reasonably good agreement with the measured values. The temperature distributions at the middle of the core predicted by CFD codes are in closer agreement with the measured values as compared to the predictions by sub-channel studies. The studies brought out the importance of thimble flow, inter-subassembly heat transfer, the effect of spacer wire and the power distribution inside the SA. The full length paper gives modeling details of the SA with various codes and the comparison of the results obtained. “Track 3: Fast Reactor Safety”: EBR-II Benchmarks Invited Session

Speaker: Mr Partha Sarathy UPPALA (Indira Gandhi Centre for Atomic Research)

Material FR17-Paper_ID_118.pdf

73 Final Results and Lessons Learned from EBR-II SHRT-17 Benchmark Simulations

In 2012 the IAEA initiated a 4-year coordinated research project (CRP) “Benchmark Analyses of EBR-II Shutdown Heat Removal Tests”, with Argonne National Laboratory serving as the lead technical institution. Nineteen participants from eleven countries were involved in the project. The overall purpose of the CRP was to improve validation of state-of-the-art sodium-cooled fast reactor (SFR) computer codes through comparisons of the analytical predictions against whole-plant recorded test data. A secondary purpose was training of the next generation of SFR analysts and designers through participation in international benchmark exercises. Numerical simulations were performed for the two most severe experiments conducted in the 1980s during Argonne’s EBR-II Shutdown Heat Removal Tests program. The first test was SHRT-17, where a PLOF (Protected Loss Of Flow) accident scenario was performed, and the second – SHRT-45R, where a ULOF (Unprotected Loss Of Flow) scenario was performed. This paper describes the results (blind and final) of the SHRT-17 experiment simulation, findings of the CRP benchmark exercise associated with the EBR-II SHRT-17 test, improvements proposed by the participants, and the lessons learned within the project. **Note: EBR-II Benchmarks Invited Session**

Speaker: Mr Nikita Rtishchev (IBRAE RAN)

Material IAEA-CN-245-84_FR-17.docx

6.2 Thermal Hydraulics Calculations and Experiments

Conveners: Mrs Nastasia Mosunova (Nuclear Safety Institute of the Russian Academy of Sciences (IBRAE RAN)), Mr Selvaraj Perumal (Indira Gandhi Centre For Atomic Research, Department of Atomic Energy, Kalpakkam, India)

74 Thermal Hydraulic Study of Steam Generator of PGSFR

In Prototype Gen-IV Sodium cooled Fast Reactor (PGSFR), integral once-through type, counter-flow shell-and-tube heat exchanger with straight vertical tubes was adopted for steam generators. Reliable operation of the steam generators has been a key issue through operating experience of foreign sodium cooled fast reactors because it is one of the most important components deciding the plant availability and reliability. Non-uniformity in sodium flow and temperature distributions might cause mechanical integrity problems such as tube buckling and tube-to-tube sheet junction failure in straight tubes. This work reports thermal hydraulic study on the sodium flow at the inlet plenum and the temperature distribution in the sodium-side of the PGSFR steam generator based on multidimensional numerical analysis. Optimization of porosity of distributers for achieving circumferentially uniform flow at the inlet plenum was carried out with the STAR-CCM+ CFD package. Then, the multidimensional sodium temperature distribution at tube bundle region was also calculated by the STAR-CCM+ CFD package. The heat flux from the sodium-side to the water-side was estimated using 1-D in-house code and supplied as boundary conditions at tube walls in the multidimensional CFD simulation. Iterative calculations between the STAR-CCM+ and 1-D in-house code were successfully conducted to acquire the radial and axial sodium temperature distributions under normal operation condition. The thermal hydraulic analysis results would be provided as input data to evaluate the mechanical structure integrity of the steam generator of the PGSFR.

Speaker: Dr Jonggan Hong (KAERI)

Material FR17-Full_Paper-PDF_revised.pdf FR17_List_of_revision.pdf

75 EFFECT OF INLET TEMPERATURE AND OPERATING LINEAR HEAT RATING (LHR) ON THE MAXIMUM ACHIEVABLE BURNUP OF MK-1 CARBIDE FUEL IN FBTR

India had constructed and is operating Fast Breeder Test Reactor (FBTR) with Mixed Carbide Fuel, a first of its kind in the world, as driver fuel. Mixed Carbide was chosen as 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. The Linear Heat Rating (LHR) and burnup of the fuel was initially set at 250 W/cm and 25 GWd/t respectively. Based on rigorous theoretical analysis and Post Irradiation Examination (PIE) done at 25 GWd/t, 50 GWd/t and 100 GWd/t burnup intervals, the LHR limit was raised to 400 W/cm and allowable burn-up was raised to 155 GWd/t. The burnup limit of the fuel SA comes from the following factors: Wrapper dilation; Wrapper residual ductility; Fission gas pressure and Fuel Clad Mechanical Interaction (FCMI) induced stress in pin; Clad strains; Clad residual ductility; Clad Cumulative Damage Fraction (CDF); Subassembly flow reduction, etc. Presently, the operating parameters like inlet temperature and the peak LHR of the FBTR of MK-1 fuel SA are 400°C and 400 W/cm respectively which may result in different limits on the achievable burnups. In this work, the effect of LHR & inlet temperature have been comprehensively studied on the achievable burnup of the MK- 1 fuel SA. From the analysis, it is observed that the two enveloping parameters that govern the SA life are wrapper dilation and pin CDF. The maximum burnup achievable with an operating LHR of 400 W/cm is 85 GWd/t and 114 GWd/t for inlet temperatures of 400°C and 380 ° C respectively. The reduction in the inlet temperature by 20 °C not only decreases the fuel swelling but also helps in increasing the free swelling phase without FCMI. Thus, this study gives an insight on the behaviour of the MK-1 carbide fuel in FBTR for the present operating conditions of the FBTR and the influence of inlet temperature and operating LHR on the achievable burnup.

Speaker: Mr CLEMENT RAVI CHANDAR SOWRINATHAN (IGCAR, INDIA)

Paper IAEA-CN-245-310.pdf

76 Extension to Heavy Liquid Metal coolants of the validation database of the ANTEO+ sub-channel code

Among the numerous numerical methods available for preliminary design verification purposes, the sub-channel one has historically been the reference, thanks to its ability to cover the scale between CFD and system codes which is the one of particular interest for the core designer. Recently, the sub-channel code for liquid metal applications ANTEO+ has been developed by ENEA and a comprehensive validation performed, covering all the salient aspects of the fuel assembly thermal-hydraulic analysis like pressure drops, sub-channel and outer clad temperatures. Due to the database available at the time, the focus for the sub-channel temperatures validation was mainly related to sodium and sodium eutectics coolants. Thanks to the increasing interest in heavy liquid metal coolants for fast reactor applications, numerous experimental activities have been very recently performed (CLEAR-S, SEARCH) and many are still ongoing, enabling the extension of the previous ANTEO+ validation database so to make this tool even more persuasive for Generation IV reactor concepts applications and mostly to estimate the uncertainty to recon to the code results. In the present work, ANTEO+ validation against the most recent experiments with heavy liquid metal coolant is presented: several tens of new experimental points have been considered in this campaign, covering a broad range of configurations which spans over the one of anticipated interest. The results of this validation activity have confirmed the good predictive ability of the code, notably when compared to other state of the art tools. Some criticalities have also emerged, especially to what concerns the sub-channels and pins close to the wrapper, which significantly modifies their thermal field; this has a particular impact on the Nusselt number, highlighting the lack, in the open literature, of a reliable correlation for the outer row of pins.

Speaker: Mr Francesco Lodi (University of Bologna)

Material IAEA-FR17_Paper-CN-245-232.docx (Protected)

77 Density of sodium along the Liquid-Vapor Coexistence Curve, including the Critical Point

Sodium densities along the whole liquid-vapor coexistence curve are reanalyzed using the equation proposed by Apfelbaum and Vorob’ev(1). The formulation has built-in the correct behavior for liquid and vapor densities, both at low temperature and in the near-critical region. Thus, it satisfactorily represents the available experimental data in the low and intermediate temperature range, while providing a sound density extrapolation to the critical point: in reduced units, the calculated values for sodium are consistent with those measured for Rubidium and Cesium(2), as required by the principle of Corresponding States. The enthalpy of vaporization, calculated via Clausius-Clapeyron relation, is also correctly described. The main differences between our results and those from the previous formulation by Finck and Leibowitz(3) are found in the high-temperature region (2300 K – Tc), where the coexistence curve predicted by the latter exhibits an unusual shape. Our results indicate that the value for the critical density, (180 ± 10) kg/m3, is 20 % lower than the one recommended before (219 ± 20) kg/m3. (1) E. M. Apfelbaum and V. S. Vorob’ev, The Wide-Range Method to Construct the Entire Coexistence Liquid−Gas Curve and to Determine the Critical Parameters of Metals, J. Phys. Chem. B, 2015, 119, 11825–11832. (2) S. Jüngst, B. Knuth and F. Hensel, Observation of Singular Diameters in the Coexistence Curves of Metals, Phys. Rev. Lett. 1985, 55, 2160-2163. (3) J. K. Fink and L. Leibowitz, Thermodynamic and Transport Properties of Sodium Liquid and Vapor, ANL/RE-95/2 (1995).

Speaker: Dr M. Laura Japas (CNEA - Argentina)

Paper IAEA-CN-245-440_vf.pdf

78 Experimental investigations of velocity and temperature fields, stratification phenomena in a integral water model of fast reactor in the steady state forced circulation

The results of experimental investigations of local velocities for height and radius of the top of the camera in the plane in the direction from the core center to the intermediate heat exchanger and the temperature of the coolant in the upper (hot) chamber, and other elements of the circulation circuit on the integrated water model of the reactor on fast neutrons (scale ~ 1:10) for the stationary forced circulation mode, simulating a nominal operation regime. The data obtained on stand V-200 using a specially designed and implemented system of measurement that provides high measurement accuracy and speed of registration. The results show that the structure of nonisothermal motion of the coolant in the top chamber model is defined by the action of lifting forces: hot coolant from the core rises up through the Central column to the surface section and forms a vast vortex nearly isothermal zone in the upper region of the chamber from which flows into the intermediate heat exchangers. Above the side screens formed of insulated cold zone of the heat carrier, the size of which increase with the overall consumption increase. On stratified horizontal boundary insulated zones across the cross-section model of the reactor tank there are internal waves which cause temperature pulsations in the material of the walls of the equipment. There is a significant and stable thermal stratification of the coolant not only in the peripheral area of the top chamber of the reactor above the side screens, but in the cold and the pressure chambers, elevating the enclosure, the cooling system of the reactor, at the outlet of the intermediate heat exchangers. At the boundaries of stratified and recycling entities recorded strong gradients and temperature pulsations, allowing to judge about the amplitude and frequency characteristics of temperature pulsations in these potentially hazardous areas. The data obtained indicate the necessity of taking into consideration the stratification phenomena in justifying reliability management, security, design terms of operation of fast reactors.

Speaker: Mr Aleksandr Trufanov (SSC RF-IPPE)

Paper Paper_439_(IPPE_Trufanov)-updated.pdf

79 Development and Validation of Multi-scale Thermal-Hydraulics Calculation Schemes for SFR Applications at CEA

In the framework of the ASTRID Gen4 SFR project, extensive R&D efforts are under way to improve and better validate the SFR thermal-hydraulics codes available at CEA. These efforts include : - The development and validation of SFR-specific models in CATHARE. Developed at CEA, CATHARE is the reference STH code for French LWR safety studies : SFR developments are being integrated and validated into the lastest version of the code, CATHARE3. - The development and validation of TrioMC, a subchannel code specific to SFR core TH. Initially created for design studies (with the aim of computing the maximum cladding temperature of a given core flowrate), TrioMC has been upgraded in order to compute the local behavior of the core during accidental transients. - The application and validation of TrioCFD, a 3D CFD code developed at CEA, to SFR studies. TrioCFD is being used to compute flow behavior in the large plena of pool-type SFRs (hot and cold pools), as well as in the IHX primary side and in the in-core inter-wrapper gap regions. In most cases, these codes are used independently. However, in some cases, local phenomena may have a strong feedback effect on the global behavior of the reactor : for instance, during passive decay-heat removal by natural convection, inter-wrapper flows may contribute to up to 30% of the overall DHR if the heat sink is provided by DHXs in the hot pool. The strength of this contribution leads to a feed-back effect from a local (subchannel/CFD) phenomenon) to the global (system) scale. In order to model such effects, a coupling between CATHARE, TrioMC and TrioCFD has been developed at CEA and integrated into a new code : MATHYS (Multiscale ASTRID Thermal-HYdraulics Simulation). Within MATHYS, TrioMC and TrioCFD are coupled at their boundaries (core outlet and hex-can sides), using a domain-decomposition approach : then, the two codes are coupled with a CATHARE simulation of the complete system using a domain-overlapping method. The resulting multi-scale simulation is able to account for feedback effects between all three scales. This paper first outlines the development and validation efforts related to CATHARE, TrioMC and TrioCFD; then, the coupling algorithm underlying MATHYS is described. The final section discusses the validation of MATHYS : overall approach, validation of coupled effects on existing experiments (TALL-3D for STH/CFD, PLANDTL-DHX for subchannel/CFD, PHENIX at the integral scale).

Speaker: Mr Yannick Gorsse (CEA-Saclay)

Paper mathys_rev.docx

5:10 PM Coffee Break

Panel 1: Development and Standardization of Safety Design Criteria (SDC) and Guidelines (SDG) for Sodium Cooled Fast Reactors

Convener: Dr Vladimir Kriventsev (IAEA)

80 The Safety Design Criteria Development and Summary of Its Update for the Generation-IV SFR Systems (USA/Japan/GIF)

The Generation-IV International Forum (GIF) Task Force completed development of Safety Design Criteria (SDC) for the Generation-IV SFR systems in May, 2013. SDC reflects high level GIF safety and reliability goals (excellence in operational safety and reliability, and reduced likelihood and degree of core damage) and follows GIF basic safety approach (application of defense-in-depth and emphasis on inherent and passive safety features so that safety is built-in to the design, not added-on). The SDC report aimed to establish reference criteria for safety design of structures, systems and components and achieve harmonization of safety approaches among GIF member states. Following its public release, SDC report was distributed to international organizations and national regulatory bodies for review and feedback. Based on comments received during the following two year period, SDC report underwent a revision reflecting important feedback received from IAEA, NRC (USA), IRSN (France), and NNSA (China). This paper will provide an overview of SDC development effort, and summarize the comments/suggestions received from its international review along with their resolutions by the GIF Task Force.

Speaker: Dr Yasushi OKANO (Japan Atomic Energy Agency)

Material FR17-GIF-SDC.pdf

81 The Safety Design Guideline Development for Generation-IV SFR Systems (Japan/GIF)

The Generation-IV International Forum (GIF) Safety Design Criteria Task Force (SDC TF) has been developing a set of safety design guidelines (SDG) to support practical application of SDC since completion of the “SDC Phase I Report” that clarifies safety design requirements for Generation IV SFR systems. The main objective of the SDG development is to assist SFR developers and vendors to utilize the SDC in their design process for improving the safety in specific topical areas including the use of inherent/passive safety features and the design measures for prevention and mitigation of severe accidents. The first report on “Safety Approach SDGs” aims to provide guidance on safety approaches covering specific safety issues on fast reactor core reactivity and on loss of heat removal. The second report on “SDGs on key Structures, Systems and Components (SSCs)” focuses on the functional requirements for SSCs important to safety; reactor core system, reactor coolant system, and containment system.

Speaker: Mr Ryodai Nakai (Japan Atomic Energy Agency)

82 Considerations on GEN IV safety goals and how to implement them in future Sodium-cooled Fast Reactors (France)

From a general perspective, generation IV (GEN IV) reactors should excel in safety, and as part of a continuous improvement process, provide safety enhancements with respect to GENIII reactors. GENIII safety objectives are already very ambitious, notably regarding: - Prevention of the severe accident; - Mitigation of the severe accident in the frame of the fourth level of defense in depth; - Response to external hazards, including natural hazards of extreme intensity. Concerning GENIV sodium-cooled fast reactors (SFR), the achievement of these ambitious safety objectives and the reinforcement of the robustness of the safety demonstration, will be ensured: - Firstly, by mastering the sensitive points of the SFR such as neutron reactivity potential of the core, chemical reactivity of sodium, inspection of structures under sodium. - Secondly, by taking full benefit in the design of the favorable characteristics of the SFR such as large thermal inertia, large margin to boiling, natural convection capabilities and by providing high diversification and independence between safety systems associated to different levels of defense in depth. The paper presents some of these possible ways of safety improvement for the future SFR.

Speaker: Mr Paul Gauthé (CEA)

Material fr17_papier_sur_GEN_IV_safety_goals.docx

83 Safety criteria for future Indian SFRs (India)

to be provided later by the author

Speaker: Mr Subhash Chander CHETAL (India)

84 Compliance of Korean SFR Safety Design Approaches with Generation-IV Safety Design Criteria (Korea, R. of)

Korea Atomic Energy Research Institute (KAERI) is developing Prototype Generation-IV SFR (PGSFR). The first design stage has been completed at the end of 2015 and the preliminary safety information document (PSID) has been issued as a main outcome of the phase. The safety design approach of PGSFR is compliance of defense-in-depth and to enhance the inherent and passive safety design features of a metal fuel and pool type sodium system. Additional design measures against the severe accident mitigation has been implemented into the PGSFR design. Inherent reactivity feedback resulting from the metal fuel properties plays a positive role during design basis accidents and design extension conditions, which is basic mechanism to prevent the severe accident as well as the fully passive safety grade decay heat removal systems. The ex-vessel cooling and self-actuated shutdown system provides additional design margin against severe accident propagation toward the goal of molten-fuel in-vessel retention. The compliance of PGSFR with Generation-IV SDC will be explained in more details during panel discussion.

Speaker: Dr JAEWOON YOO (Korea Atomic Energy Research Institute)

85 Russian SFR Safety Requirements and Approaches and Their Correspondence to Generation-IV SFR Safety Design Criteria (Russia)

Speaker: Iurii Ashurko (IPPE)

86 Panel Discussion

Tuesday, June 27

Plenary Session 27 June

Conveners: Dr Hideki Kamide (Japan Atomic Enery Agency), Dr Vladimir Kriventsev (IAEA)

87 Indian Fast Reactor Programme : Status and R&D Achievements

Fast Breeder Reactors form the second stage of India’s three stage Nuclear Power Programme based on the domestic nuclear resources. Indira Gandhi Centre for Atomic Research (IGCAR) is primarily dedicated for the broad based research & development of sodium cooled fast reactors, fuel cycle and associated technologies. India is operating a Fast Breeder Test Reactor (FBTR) since 1985, fuelled with a unique Pu rich mixed carbide fuel (70% PuC + 30% UC). It has so far completed 24 irradiation campaigns in its successful operation over thirty years. Fuels of all types viz. carbide, oxide as well as metal fuels (both binary and ternary) are currently under irradiation. FBTR has served as a test bed for various experiments, fuel and structural material irradiation, isotope generation programs. The mixed carbide fuel has demonstrated a record burnup of 165 GWd/t and it has been operated at 400 W/cm peak LHR and at higher operating temperatures. Currently, a 500 MWe Prototype Fast Breeder Reactor (PFBR) designed and developed by IGCAR, is in an advanced stage of commissioning. The design of PFBR incorporates several state-of-art features and is foreseen as an industrial scale techno-economic viability demonstrator for India’s FBR program. IGCAR is presently engaged in the design of 600 MWe oxide fuelled FBRs incorporating many advanced features. CORAL (COmpact Reprocessing of Advanced fuels in Lead cell) facility has reprocessed spent fuel discharged from FBTR with burnup up to 155 GWd/t and adequate decontamination has been demonstrated. Currently, a Demonstration fast reactor Fuel Reprocessing Plant (DFRP) is being established to process both MOX and mixed carbide fuels. A dedicated co-located Fast Reactor Fuel Cycle Facility (FRFCF) for PFBR is under construction. For the future, IGCAR has initiated development program on metallic fuel. Demonstration of fuel fabrication and pyroprocessing / aqueous technologies for metal fuels on an engineering scale is being pursued. The R & D areas address all domains of fast reactor science and technology, including sodium technology, safety, materials development, fuel cycle, chemistry, sensors, advanced instrumentation and inspection. This paper presents an overview of the broad based R&D carried out by IGCAR in the domain of reactor technology, fuel cycle technology, materials development, basic sciences in support of fast reactor program, fuel chemistry, sodium technology, engineering development etc.

Speaker: Mr ARUN KUMAR BHADURI (INDIRA GANDHI CENTRE FOR ATOMIC RESEARCH)

Paper CN245-420_Keynote_Bhaduri.docx

88 Current status and future view of the fast reactor cycle technology development in Japan

As stated in the “Fourth Strategic Energy Plan”, which was approved by the Cabinet in April 2014, Japan continues to position nuclear energy as a major base-load power source even after the TEPCO’s Fukushima Dai-ichi Nuclear Power Station accident. Its basic policy in the plan is to promote nuclear fuel cycle in terms of the efficient use of resources and reduction in volume and toxic level of high-level waste, and carry out fast reactor (FR) cycle R&D for the commercialization, taking advantage of international cooperation. For the commercialization of FR cycle, Japan Atomic Energy Agency (JAEA) is conducting several R&D activities primarily focusing on 1) the reduction in volume and toxic level of radioactive waste and 2) the improvement of the safety of FRs and FBRs, as mentioned in the Fourth Strategic Energy Plan, in parallel with R&D utilizing international cooperation with bilateral frameworks such as ASTRID program with France and multilateral frameworks such as the Generation IV International Forum (GIF). In the nuclear fuel cycle R&D, SmART cycle project to conduct small-scale minor actinide (MA) recycling using existing facilities is in progress. The prototype FBR Monju will be subject to a fundamental review, and the government’s official policy on Monju together with FR development in concrete terms will be presented by the end of this year. As for the experimental FR Joyo, JAEA completed the upper core structure (UCS) replacement work and is preparing to make an application for earlier restart under the new regulatory requirements.

Speaker: Mr Yutaka Sagayama (Japan Atomic Energy Agency)

Material 201705011__FR17_No.571(Yutaka_Sagayama,_Japan).doc

89 Status of Sodium Cooled Fast Reactor Development Program in Korea

The Korea Atomic Energy Commission (KAEC) authorized the R&D action plan for the Advanced SFR (sodium-cooled fast reactor) and the pyro-process to provide a consistent direction to long-term R&D activities in December, 2008. This long-term advanced SFR R&D plan was revised by KAEC in November 2011 in order to refine the plan and to consider the available budget for SFR. The revised milestones include specific design of a prototype SFR by 2017, specific design approval by 2020, and construction of a prototype SFR by 2028. The prototype SFR program includes the overall system engineering for SFR system (NSSS and BOP) design and optimization, integral V&V tests, and major components development. Based upon the experiences gained during the development of the conceptual designs for KALIMER, the conceptual design of SFR prototype plant (PGSFR) had been carried out in 2012 and has been performing a preliminary design since 2013. The first phase of the development of PGSFR has been completed at the end of February 2016 and now going toward the second design phase in 2016. All the design concepts of systems, structures and components (SSCs) have been determined and incorporated into the preliminary safety information document (PSID), which includes basic design requirements, system and component descriptions, the results of safety analysis for the representative accident scenarios. The PSID will be a base material for a pre-review of the PGSFR safety. The target of the second phase of PGSFR design is to prepare a specific design safety analysis report (SDSAR) by the end of 2017. The specific safety analysis report is equivalent to the conventional preliminary safety analysis report (PSAR) but without the specific site information of the plant. The design activities are being carried out to freeze the design details of PGSFR by the end of 2016. To support the design, various R&D activities are being performed in parallel with design activities, including V&Vs of design codes and system performance tests. The details on the design status and plan will be presented in the conference

Speaker: Dr JAEWOON YOO (Korea Atomic Energy Research Institute)

Material FR17_Full_Paper_Rev01.docx

90 Research and Pilot Fast Neutron Reactors in Russian Federation as the Ground for Development of Worldwide Commercial Technologies

As early as the exploration of nuclear energy based on the fission of heavy atoms, it became obvious that one of the key conditions for its future wide-scale application is nuclear breeding. For this purpose, in the middle of the last century, the Soviet Union started implementing an experimental infrastructure to develop and construct fast reactors. In a quick period of time the following test reactors were developed and commissioned: BR-1 (1955), BR-2 (1956) and BR-5 (1958). In 1961, a critical assembly BFS-1 was put into operation to simulate neutronic characteristics of fast reactors. In 1969, a fast test sodium-cooled reactor BOR-60 was commissioned having a steam turbine to produce electricity. The reactor is intended to test all the fast sodium reactor technologies. The same year, the world’s largest critical assembly BFS-2 was constructed. For the next eleven years, commercial power reactors BN-350 (1973) and BN-600 (1980) were commissioned. After the severe accident at the Chernobyl’s 4th unit in 1986, the Soviet Union’s intensive nuclear energy development program was suspended and the next following decades were devoted to the fundamental research in the reactor feasibility and safety as well as to the development of new reactor materials and design concepts. Since the beginning of 1990s, the Russian Federation has conducted R&D and design activities to develop lead-bismuth- and lead-cooled fast reactors with inherent safety. The development activities related to the fast sodium reactors have been continued and, so far there was put into operation the BN-800 commercial power reactor with a hybrid core operating oxide and MOX fuel; the BN-1200 commercial fast sodium reactor project was developed as well. Speaking about the test reactors, the BOR-60 lifetime has been extended till 2020 to continue the in-pile testing of the structural and fuel materials; a design of a new fast test reactor MBIR was developed and its construction has been started to further develop and experimentally support the wide-scale program for commercial power reactors of the next generation.

Speaker: Mr Alexander Tuzov (JSC "SSC RIAR")

10:00 AM Coffee Break

1.3 SYSTEM DESIGN AND VALIDATION

Conveners: Mr RAGHUPATHY S. (Indira Gandhi Centre for Atomic Research, Kalpakkam), Mr Sergey Ruhlin

91 EXPERIMENTAL SEISMIC QUALIFICATION OF DIVERSE SAFETY ROD AND ITS DRIVE MECHANISM OF PROTOTYPE FAST BREEDER REACTOR

Prototype Fast Breeder Reactor (PFBR) has two independent and diverse fast acting shutdown systems. The mechanisms that handle Control & Safety Rods (CSR) are called Control and Safety Rod Drive Mechanism (CSRDM). CSRDM & CSR are for start up, control of reactor power, controlled shutdown and SCRAM of the reactor. The mechanisms that handle Diverse Safety Rods (DSR) are called Diverse Safety Rod Drive Mechanism (DSRDM). There are 9 CSRDMs and 3 DSRDMs provided. DSR serves to shutdown the reactor on demand and are in fully raised position during normal reactor operation. CSRDM/DSRDM consists of independent sets of sensors connected to two reactor protection logics of different designs. The output of either of the reactor protection logic system is capable of ordering safety actions through SCRAM signal by de-energizing electromagnet of CSRDM & DSRDM. As part of seismic qualification, full scale DSRDM along with DSR was extensively tested at room temperature in water for two earthquake levels, namely Operation Base Earthquake (OBE) and Safe Shutdown Earthquake (SSE). Drop time of DSR and its mobile assembly at different instant of dropping from the beginning of shaking were obtained. Full insertion of DSR within the stipulated time and healthy functioning of DSRDM during and after seismic testing have been demonstrated. The details of seismic testing carried out for DSRDM is presented in this paper.

Speaker: Mr RAGHUPATHY S. (Indira Gandhi Centre for Atomic Research, Kalpakkam)

Material FR17_344.pdf

92 Progress in the ASTRID Gas Power Conversion System development

Within the framework of the French 600 MWe Advanced Sodium Technological Reactor for Industrial Demonstration project (ASTRID), two options of Power Conversion System (PCS) were studied during the conceptual design phase (2010-2015): - the use of a classical Rankine water-steam cycle, similar to the solution implemented in France in Phenix and Superphenix , but with the goal of greatly reducing the probability of occurrence and limiting the potential consequences of a sodium-water reaction; chosen as the reference for the ASTRID Plant Model during the conceptual design phase due its high level of maturity, - an approach which has never been implemented in any Sodium Fast Reactor using a Brayton gas cycle. Its application is mainly justified by safety and acceptance considerations in inherently eliminating the sodium-water and sodium-water-air reaction risk existing with a Rankine cycle. The ASTRID conceptual design phase period allowed to greatly increase the maturity level of a standalone Gas Power Conversion System option. It has been thus decided to lay during the 2016-2017 phase the ASTRID Gas PCS integration studies at the same level as that achieved by ASTRID Water based PCS at the end of 2015. The 2016-2017 period, in which the Gas PCS is integrated in the overall layout of the reactor, will allow to better specify the technical and economic implications of the selection of gas PCS taking into account all the aspects of the integration of such an option. A well-documented comparison between the two systems will be therefore facilitated. This paper resumes progress in the integration of the Gas Power Conversion System in the Astrid Reactor Plant Model. It describes the main characteristics defined particularly on the Balance of Plant (BOP), the turbomachinery, the Sodium Gas Heat Exchangers (SGHE) as well as expected performances, operability and safety analysis.

Speaker: Mr david plancq (CEA)

Material Paper_IAEA-CN-245-285.docx

93 ASTRID FUEL HANDLING ROUTE FOR THE BASIC DESIGN

At the beginning of the Basic Design phases of ASTRID starting in 2016, the entire fuel handling route has been challenged in order to improve some aspects like availability and investment cost. Especially, studies performed at the end of preliminary design (2010 – 2015) phase show that the availability target will be difficult to obtain with a significant risk of malfunction because of multiple handling operations in series. Two main changes have been then decided: the implementation of an in-sodium external buffer zone, similar with an in-sodium external storage but associated with an in-primary vessel storage to reduce its size and its allowable residual power, and the merging of the fresh assembly storage with the spent fuel assembly storage that allows both reduction of size and equipment. These new options have the first advantages to reduce drastically the number of operation that has to be done during scheduled outage and minimizing the overall Balance Of Plant. After description of the entire fuel handling route adopted for ASTRID, this paper aims at identifying the advantages of this new option and points the remaining issues or questions that will be studied in details during the current Basic Design phase.

Speaker: Mr FRANCK DECHELETTE (CEA)

Material FR17_-_IAEA-CN-245-395_-_Fuel_handling.pdf

94 COMPONENT HANDLING SYSTEM : PFBR AND BEYOND

Component handling system deals with the handling of fresh and spent subassemblies (fuel handling) and irradiated primary system components using special flasks (special handling). In FBRs, design of fuel handling machines is very important considering the fact that in-vessel handling is a blind operation due to opacity of sodium and most of the fuel handling operations are carried out remotely. Special features are provided in the design of hoisting system of fuel handling machines like single failure proof design features in order to avoid fall of subassembly during handling. Incidents on component handling system have a serious impact on plant availability and hence utmost care is taken in the design to avoid wrong operations of fuel handling machines. PFBR in-vessel handling utilises two rotatable plugs and an offset arm type machine (Transfer Arm). For ex-vessel handling, an A-frame type machine called Inclined fuel transfer machine (IFTM) is used. Several other machines are used as part of the fresh and spent fuel handling chain. A water pool type storage is provided for ex-vessel storage before the subassemblies are transferred to the reprocessing plant. Critical primary fuel handling machines namely Transfer arm and IFTM were qualified by cyclic testing in air and in sodium in dedicated test facilities. The design of PFBR fuel handling system and the design validation of the critical fuel handling machines are described in this paper. The design, manufacturing and testing of fuel handling machines of PFBR have given valuable feedback for future FBRs. Beyond PFBR, six more oxide fuelled FBRs are planned as twin units. Refuelling in fast reactors being done off-line, gives opportunity to evolve a fuel handling system shared between multiple units for improved economy. The design of fuel handling system for the twin unit 600 MWe future FBRs is described. The rationale behind the changes proposed with respect to PFBR is brought out. Most of the fuel handling equipment is shared between the twin units and a unique twin unit layout has been evolved which is also covered in this paper. In the future, it is planned to deploy metal fuelled based reactors for achieving faster growth through rapid deployment of FBRs. The details of fuel handling system conceived for future metal fuelled FBRs is also brought out.

Speaker: Mr RAGHUPATHY S. (IGCAR, KALPAKKAM)

Material FR_17_314corr.doc

95 Main operation procedures for ASTRID gas power conversion system

Until the end of the first part of the basic design phase (2017), the ASTRID project has made the choice of studying a power conversion system (PCS) based on a Brayton cycle with nitrogen as coolant. The justification is related to a safety and public acceptance considerations in order to inherently eliminate the sodium-water and sodium-water-air reactions risks. The objective of the studies engaged is to enhance the level of maturity of the gas PCS as close as possible to the classical Rankine cycle. The choice of two PCS of 300 MWe each has been made in order to limit the gas inventory, the size and length of gas pipes as well as maintaining a high level of availability. This paper presents the current architecture of the gas PCS, the layout of the tertiary circuits and will also deals with specific operating procedures as start-up of the plant, scram, normal shutdowns and grid frequency control. The current procedures in the three circuits of the plant and the expected regulation will be presented. A focus will be made on the nitrogen inventory control which takes part of the electric power regulation provided to the grid. When possible, the comparison with the vapor PCS will be shown in terms of impact of thermal transients on structures. Finally some perspectives of the gas PCS use for the future of the sodium fast reactors will be drawn in terms of better cost-effectiveness of operation through optimization of its Brayton cycle.

Speaker: Mr dominique barbier (CEA)

Paper FR17_Paper_468_full_paper_operation_procedures_for_gas_PCS_final.pdf

3.3 Probabilistic Safety Assessment

Conveners: Mr Pavel Antipin (JSC “Afrikantov OKBM”), Dr Takashi Takata (Japan Atomic Energy Agency)

96 Development of Smart Component Based Framework for Dynamic Reliability Analysis of Nuclear Safety Systems

Dynamic reliability methodologies account for the safety system's time dependent characteristics while estimating the reliability. Time dependence can arise due to interaction of process variables with the hardware and hardware failure on process conditions. Though static reliability models often capture the average behavior and try to make conservative estimates, it is inadequate from a number of perspectives. First, this requires that the analyst needs to establish that the model is conservative. Second, such modeling requires more expertise and experience in the appropriate domain of the problem, rather than in the reliability methods. Third, approximate methods may be inadequate to establish reliability enhancements or degradations due to subtle alterations in the system design. In spite of the significant effort in the reliability community to establish dynamic reliability analysis methods, there are no general purpose tools similar to that available for fault tree event tree modeling. A methodology based on 'Smart Components' is being developed for dynamic reliability evaluation of safety systems involving digital IC systems interacting with process and hardware. Smart Component based dynamic method uses elements of object oriented and relational data base architecture and is suitable for being developed into a general purpose tool. The paper demonstrates the capability of the method to evaluate reliability of systems having various types of time dependence, interaction between hardware failure and process evolution and complexity by means of few case studies. The method is found to be promising for accurate modeling of dynamic as well as static scenarios.

Speaker: Mr Puthiyavinayagam Pillai (Indira Gandhi Centre for Atomic Research)

Material IAEA_189_Smart Component Methodology

97 Dynamic probabilistic risk assessment at a design stage for a sodium fast reactor.

The design of a new reactor is an iterative process. At a design stage, each reactor system may exist in several variants : combined together, these variants give different reactor designs. Probabilistic risk assessment is a tool that gives a measure (risk, and risk space) that may help to compare different reactor designs. Conservative or macroscopic way to perform probabilistic risk assessment at a design stage may be insufficient to provide enough information to distinguish between different design variants. The classical way to construct a PRA model with boolean Event Trees/Fault trees (ET/FT) is well applicable for a PWR type reactors at a design stage. Nevertheless ET/FT formalism finds its limits for a Sodium Fast Reactor if one considers long mission times and possible system recoveries. Due to thermal inertia and simplicity of thermal-hydraulic behavior, the decay heat removal function of a Sodium Fast Reactors (SFR) is a good candidate for dynamic probabilistic risk assessment. In this paper, we present a new approach to construct the dynamic probabilistic model of the Decay Heat Removal (DHR) system of a Sodium Fast Reactor using Stochastic Hybrid Automata with PyCATSHOO modeling tool.The proposed approach allows to construct a dynamic probabilistic model of a SFR DHR system that incorporates the time evolution of physical parameters and dynamic changes of DHR system state (failure or recovery of DHR system components) and presents enough flexibility to be applied at the design stage. Below, we list the important properties of a dynamic models in more details: - We define core damage as function of primary sodium temperature during an accidental sequence. The evolution of primary sodium temperature follows the simplified equations that can be validated by a qualified thermohydraulic code. Evolution of primary sodium temperature is automatically calculated for every accidental sequence to predict the end state of the sequence (OK/Core Damage). - We explicitly treat the dependency of DHR system trains on different support systems: electrical, I&C, ventilation etc. - We explicitly model Common Cause Failures between different DHR components. - We explicitly model component recoveries. Our modeling approach allows to: - have a detailed model e.g. it can be as detailed system components, - easily change the system architecture to test and compare in a realistic way different design variants, - perform an uncertainty analysis of the risk as a function of uncertainty in reliability and physical parameters

Speaker: Dr VALENTIN RYCHKOV (EDF R&D)

Material IAEA_DynamicReliability.pdf

98 PROBABILISTIC SAFETY ANALYSIS RESULTS FOR BN REACTOR POWER UNITS

Probabilistic safety analysis (PSA) is a constituent part of range of works aimed at BN power units safety assessment during operation (BN-600 and BN-800 reactors power units), lifetime extension (BN-600 reactor power unit), designing (BN-1200 reactor power unit). PSA reports are part of the document sets required to obtain appropriate Rostechnadzor licenses. JSC “Afrikantov OKBM” together with General Designer and Scientific Supervisor has performed the following PSA Level 1 (PSA-1): а) BN-600 and BN-800 reactors power units: - PSA-1 for internal initiating events for power operation mode, - PSA-1 for internal initiating events for shutdown reactor modes, - PSA-1 for internal fires, - PSA-1 for internal floods, - PSA-1 for external hazards. б) BN-1200 reactor power unit: - PSA-1 for internal initiating events for power operation mode (preliminary). General PSA goals are the following ones: – power unit safety level assessment; – recommendation development for power unit safety measures improvement. First of all, PSA-1 was developed for internal initiating events for power operation mode. All following studies are based on models prepared within that PSA-1. Within each of the PSA-1 studies, systems reliability analysis was performed, accident sequences were developed, human reliability analysis was implemented, database on initiating event frequencies and system component reliability indices was developed, integral probabilistic reactor power unit model was formed, quantitative analysis was performed including of importance, sensitivity and uncertainty analysis. The PSA database on initiating event frequencies and component reliability indices is developed and updated based on the analyzed experience of BN-600 reactor power unit. For BN-800 and BN-1200 reactors power units PSA data analysis takes into account power units design distinctions. At the present time JSC “Afrikantov OKBM” is continuing to improve all PSA models. Among other actions BN-600 and BN-800 reactors power units safety measures improvement based on Fukushima Daichi accident lessons learned and power units operating experience feedback update are taken into account. Probabilistic safety analysis Level 2 is being performed.

Speaker: Mr Pavel Antipin (JSC “Afrikantov OKBM”)

Paper IAEA-CN-245-419.pdf

Slides IAEA-CN-245-419.pptx

5.3 Advanced Fast Reactor Cladding Development I

Conveners: Dr Jeong-Yong Park (Korea Atomic Energy Research Institute), Dr Liudmila Zabudko (Innovative &Technology Center by "PRORYV" Project)

99 Development of core and structural materials for fast reactors

This paper summarizes ongoing efforts in Japan Atomic Energy Agency on the development of core and structural materials for sodium-cooled fast reactors. For core materials, oxide dispersion strengthened (ODS) steels and 11Cr ferritic steel (PNC-FMS) will be applied to fuel pin cladding and wrapper tube, respectively. As for ODS steel, 9Cr,11Cr-ODS steels have been extensively developed. Their laboratory scale manufacturing technology has been developed including reliability improvement in tube microstructure and strength homogeneity. Large scale manufacturing technology development and mechanical testing for codification of material strength standard are on-going. As for PNC-FMS wrapper tube, development of dissimilar joining technique with type 316 steel and properties evaluation of dissimilar welds have been carried out. For structural materials, 316FR stainless steel and Modified 9Cr-1Mo steel are being code qualified. Long-term data have been accumulated and the properties are analysed to establish a technical basis for 60-year design. Also described is the current status of codification of structural materials standards in the design code of fast breeder reactors published form the Japan Society of Mechanical Engineers.

Speaker: Dr Tai Asayama (Japan Atomic Energy Agency)

Material ID_77_FR17.docx

100 Results of monitoring, using high-resolution neutron diffraction, of radiation-induced damages in claddings of fuel pins after their performance in the reactor BN-600 as a ground for prolongation of their life expectancy

Fast neutron irradiation gives rise to rather complex processes developing in the fuel element claddings that lower their technical characteristics and restrict time of safe exploitation. Neutron diffraction studies open up a possibility to monitor them even at a stage of incubation period and thereby to promote development of reliable methods for life expectancy prolongation of the reactor components. An important advantage of these methods is minimal manipulations needed for work with high radioactive samples. We carried out neutron-diffraction studies of the samples prepared from the fuel elements claddings made of austenitic steel EK-164 both in the initial state and after real exploitation in different parts of the reactor BN-600 core at temperatures (360 – 630) C up to the dose of 75 dpa. It was found that the samples cut from the fuel elements of different lots had small distinctions in their structure state (inner stresses, texture, dislocation density). It was confirmed that up to the highest fluences in the study, the FCC lattice is retained, with the microstate depending of three parameters: fluence, neutron flux density and temperature. Our neutron diffraction data saying that maximal concentration of defects takes place at high fluence within the core part with the temperature of (400-550) C are in agreement with data on the fuel element claddings swelling. At the same time low neutron flux densities, temperature about 375 C and dose up to 10 dpa result in the annealing of initial defects and decrease of microstresses (dislocation densities). It is interesting that within the core part at a temperature of 628 C and 75 dpa defect concentration is shown to decrease again, down to the level being lower initial. Now we continue our studies of the claddings materials up to 105 dpa. The research was carried out at IMP Neutron Material Science Complex within the state assignment of FASO of Russia (theme “Flux” No. 01201463334), supported in part by Ural Branch of Russian Academy of Sciences (Project № 15-17-2-3).

Speaker: Dr Vladimir Bobrovskii (M.N. Miheev Institute of Metal Physics of Ural Branch of Russian Academy of Sciences)

Material IAEA-CN-245-32.docx

101 FRACTURE STRAIN AND FRACTURE TOUGHNESS PREDICTION FOR IRRADIATED AUSTENITIC STEELS OVER WIDE RANGE OF TEMPERATURES TAKING INTO ACCOUNT THE EFFECT OF SWELLING AND THERMAL AGEING

Fracture toughness and fracture strain of austenitic steels are the important performance properties, which control serviceability of the components of fast reactors including fuel assemblies. It is known that the above properties decrease under irradiation and thermal ageing. Especially strong decrease of fracture toughness occurs under irradiation accompanied by swelling. It is necessary to note that ductile transgranular fracture mechanism dominates at temperature less than 500 degrees Celsius even for highly embrittled material with high swelling. Such type of fracture can be predicted by the ductile fracture model proposed by authors early. At the temperature higher than 500 degrees Celsius fracture of irradiated austenitic steels occurs by intergranular mode. Such embrittlement is known as high temperature radiation embrittlement (HTRE). There are only few experimental data on HTRE, and models are practically absent for prediction of fracture toughness and ductility for different levels of stress triaxiality. The present work considers the features of prediction of fracture strain and fracture toughness for irradiated austenitic steels over wide range of temperatures with regard for swelling and thermal ageing. Model is developed for prediction of both quasi-brittle intergranular and ductile transgranular fracture and the fracture mechanisms transition. The model allows one to predict fracture strain and fracture toughness of material for different stress triaxiality taking into account the influence of neutron irradiation, swelling and grain boundary damage by He. The data are represented for fracture modeling of the material of decommissioned fuel assemblies.

Speaker: Dr Alexander Sorokin (Central Research Institute of Structural Materials “Prometey”)

Paper IAEA-CN-245-153.docx IAEA-CN-245-153.pdf

102 IAEA activities in the area of Nuclear Power Reactor Fuel Engineering

The main IAEA program implementation tools in the area of fuel engineering are Coordinated Research Projects (CRP), Technical Meetings (TM), Expert Reviews, and NEA-IAEA International Fuel Performance Experiments (IFPE) Database. This report provides information about organization and implementation practices of these activities, and summarizes their major outputs including ongoing CRPs and TMs in the area of Fuel Engineering. The first announcement and preliminary information on the new CRP “Fuel Materials for Fast Rectors” to be launched in 2018 will be presented, including main topics for International collaboration within the CRP.

Speaker: Dr Mikhail Veshchunov (IAEA)

Paper Paper_Veshchunov-2.docx

103 Examination of Fast Reactor Materials and Structural Elements at JSC “INM” Premises

At the Institute of Nuclear Materials post irradiation examination has been carried out since BN-600 reactor commissioning to justify safety and reliability of different core elements during routine operation and to search for new ways of extending their service life. The examination is carried out in close cooperation with the design bureau (Afrikantov OKBM), nuclear operator (Beloyarsk NPP), materials testing enterprise (VNIINM), fuel assembly manufacturer (MSZ) and other enterprises. It helps to use post irradiation examination results promptly to advance reactor structural elements and improve economic efficiency. Main aspects of the examination are as follows: - examination of fuel elements and shroud tubes of standard, trial and test fuel assemblies; - examination of reactor control and safety units (control rods including absorber elements and shroud tubes); - examination of materials science assembly samples irradiated in BN-600 reactor; - investigation of possible service life extension from 30 to 45, and then to 60 years. The examination carried out at INM is unique because it is not limited with statement of fact of changes in structural elements and their material properties. The aim of the examination is to predict their further behaviour and find out the cause of the changes. It is not sufficient to carry out separate post irradiation examinations, there should be a systematic result set based upon theoretical concepts on the process mechanisms, descriptive modeling of structural evolution processes and corresponding changes in physical and mechanical properties. It is also necessary to improve existing examination techniques and develop new ones to obtain characteristics used to predict residual and limited life for core elements and the reactor as a whole. The paper aims to show the main results of BN-600 structural element examination at INM, demonstrate their practical application, and make a review on the developed theoretical concepts and the development of the techniques correlating with the examination.

Speaker: Mr Alexander Kozlov (Joint Stock Company "Institute of Nuclear Materials")

Material IAEA-CN-245-94_Rev.pdf

6.10 Other issues of code development and application

Conveners: Mrs Nastasia Mosunova (Nuclear Safety Institute of the Russian Academy of Sciences (IBRAE RAN)), Mr Tanju Sofu (Argonne National Laboratory)

104 U.S. Sodium Fast Reactor Codes and Methods: Current Capabilities and Path Forward

The United States has extensive experience with the design, construction, and operation of a variety of sodium cooled fast reactors (SFRs) over the last six decades. Despite the closure of various facilities, the U.S. continues to dedicate research and development (R&D) efforts to the design of novel experimental, prototype, and commercial facilities. Accordingly, in support of the rich operating history and ongoing design efforts, the U.S. has been developing and maintaining a series of tools with capabilities that envelope all facets of SFR design and safety analyses. This paper will provide an overview the current U.S. SFR analysis toolset, including codes such as SAS4A/SASSYS-1, MC2-3, SE2-ANL, PERSENT, NUBOW-3D, and LIFE-METAL, as well as the higher-fidelity tools (e.g. PROTEUS) being integrated into the toolset. Current capabilities of the codes will be described, and key ongoing development efforts will be highlighted.

Speaker: Dr Acacia Brunett (Argonne National Laboratory)

Material 41_Final.pdf

105 Validation of Advanced Metallic Fuel Models of SAS4A using TREAT M-Series Overpower Test Simulations

The SAS4A safety analysis code has been extended to include mechanistic and physics-based models of U-Pu-Zr and U-Zr metallic fuel pins. The simulation of various phenomena such as metal fuel component migration, fission gas behavior, clad wastage formation, gas swelling induced axial fuel expansion, in-pin and ex-pin molten fuel relocation, and clad failure models has been significantly enhanced. The integrated code is validated through analyses of eight metal fuel TREAT M-Series overpower experiments. In this study, the SAS4A calculated fuel reactivity and clad failure data are compared with the corresponding experimental data. The results show that the code satisfactorily predicts solid fuel axial expansion, molten fuel in-pin relocation, cladding loss due to rapid eutectic penetration, cladding creep fracture and molten fuel ejection to the coolant channel. The study shows that the uncertainties in transient response tend to be higher for the lower burnup fuel.

Speaker: Dr Aydin Karahan (Argonne National Laboratory)

Paper Revised_paper_FR17_11-4-2017.pdf

106 Review of Transient Testing of Fast Reactor Fuels in the Transient REActor Test Facility (TREAT)

The restart of the Transient REActor Test (TREAT) facility provides a unique opportunity to engage the fast reactor fuels community to reinitiate in-pile experimental safety studies. Historically, the TREAT facility played a critical role in characterizing the behavior of both metal and oxide fast reactor fuels under off-normal conditions, irradiating hundreds of fuel pins to support fast reactor fuel development programs. The resulting test data has provided validation for a multitude of fuel performance and severe accident analysis computer codes. This paper will provide a review of the historical database of TREAT experiments including experiment design, instrumentation, test objectives, and salient findings. Additionally, the paper will provide an introduction to the current and future experiment plans of the U.S. transient testing program at TREAT.

Speaker: Dr Colby Jensen (Idaho National Laboratory)

Material CJensen_FR17_submit.pdf

107 USDOE NEAMS Program and SHARP Multi-Physics ToolKit for High-Fidelity SFR Core Design and Analysis

Under the Reactors Product Line of U.S. DOE’s Nuclear Energy Advanced Modeling and Simulation (NEAMS) program, 3-D, high-fidelity multi-physics simulation capabilities are being developed to address the needs of designers and analysts in studying advanced, non-water reactor systems in general, and SFRs in particular. Simulation-based High-fidelity Advanced Reactor Prototyping (SHARP) toolkit is a suite developed under the Reactors Product Line of NEAMS, and it consists of pin-by-pin neutronics, thermal hydraulic, and structural mechanics modules, as well as the capabilities to integrate these modules for multi-physics analyses. Physics modules currently include the PROTEUS neutronics code, the Nek5000 computational fluid dynamics (CFD) code for thermal-hydraulics, and the DIABLO implicit finite element analysis code for structural mechanics. Each module can be utilized as a standalone code component or as part of an integrated analysis. The development philosophy for the modules is to incorporate as much fundamental physics as possible in order to extend functionality to general reactor types, rather than developing tools for a limited set of specific reactor analysis applications. This paper summarizes the initial efforts focusing on SFR design and analysis in demonstration of the inherent and passive safety characteristics resulting from multi-physics thermal-structural-neutronics phenomena.

Speaker: Mr Tanju Sofu (Argonne National Laboratory)

Material FR17-NEAMS4[1].docx

6.3 Neutronics - 1

Conveners: Dr Andrei Rineiski (Karlsruhe Institute of Technology (KIT)), Mr Dmitry KLINOV (IPPE)

108 The APOLLO3 scientific tool for SFR neutronic characterization: current achievements and perspectives

ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration) is a Sodium Fast Reactor design that will be France's Flagship 4th Generation Reactor. Its innovative core contains many axial and radial heterogeneities (in order to obtain a negative void coefficient) and interfaces that are challenging for current deterministic codes to simulate correctly. Hence there is the need for new improvements in modeling (3D simulations, parallel processing) like those being elaborated within the APOLLO3 platform. The APOLLO3-SFR package built with APOLLO3 solvers defines reference calculation schemes associated with a nuclear data library to calculate all neutronic parameters (critical masses, sodium void, Doppler coefficient, eff, etc… ) together with certified biases and uncertainties derived from the VV&UQ process. This VV&UQ process incorporates numerical validation, a-priori uncertainties based on nuclear data covariances as well as experimental validation mainly from MASURCA, a fast mock-up reactor, located at CEA Cadarache. A future programme called GENESIS will be performed in support to the prototype ASTRID to validate the CFV core specificities. In addition, a part of the GENESIS experimental program contains integral experiment underway at the BFS facility. The paper presents the various VV&QU activities which are currently conducted to derive all neutronic characteristics with a certified uncertainty.

Speaker: Prof. Gerald Rimpault (CEA)

Paper APOLLO3_FR17_216.pdf

109 INTEGRAL EXPERIMENTS WITH MINOR ACTINIDES AT THE BFS CRITICAL FACILITIES: STATE-OF-THE-ART SURVEY, REEVALUATION AND APPLICATION

The paper presents the results of a computational and experimental analysis of a systematized and revised series of experiments carried out between 1990-2013 on measurements of absolute fission rate of minor actinides (from 237Np to 245Cm) in different neutron spectra at the BFS-1,2 facilities. A total of 25 critical configurations, i.e., reactor core models with different fuels and coolants were examined. The earlier experimental data have been revised according to more accurate data processing methods with account of permanent chamber deformations and introduced corrections to the efficiency of detecting fragments (fission events in the chambers). The computational models of assemblies presented in the international handbooks were supplemented with evaluated fission rate ratio models using non-analog algorithms. The resulting consistent set of experimental data and computational models can be used in solving various applied and fundamental problems. A generalization and re-evaluation of a series of integral experiments at the BFS facilities can serve as an information base for the verification and refinement of evaluated minor actinides neutron data. The analysis of all the available set of data on minor actinides fission rate ratio measurements can be used for supporting rationale and planning of research programs for critical assemblies.

Speaker: Ms Olga Andrianova (IPPE JSC “SSC RF –IPPE”, 1, Bondarenko sq., Obninsk, Kaluga reg. 249033)

Material _Andrianova_IAEA_FR-17_revised.docx

110 International research center based on MBIR reactor – cornerstone for Generation 4 technologies development

This report intended to provide an update on the International research center (IRC) based on the fast sodium research reactor MBIR development. The report will include the proposed IRC structure, key terms of participation, proposed areas for multilateral research, etc. It will also present the R&D possibilities that IRC members will have on the closed nuclear fuel cycle due to the Multi-functional radio-chemistry research facility, which is also being constructed at RIAR site. MBIR reactor technical parameters (very high flux, up to 3 simultaneously working independent loops, horizontal and vertical channels, high experimental capacity and other features) ensure the needed experimental support for the R&D conducted to create the new generation innovative nuclear energy facilities. MBIR and Multi-functional radio-chemistry facility at one site will provide an opportunity to execute and perfect the closed fuel cycle and radioactive waste utilization. In addition the combination of those facilities will allow to conduct the complex material testing research including creation of the new constructive materials, fuel and absorbing materials as well as to perform complex experimental tasks with the use of neutrons for fundamental studies. Both MBIR reactor and the radio-chemistry facility were awarded the ICERR status as part of RIAR facilities in September 2016. The high flux neutron fast reactor facility is a powerful instrument and cannot be realized small scale or as a modular complex which leads to high capital costs and overcapacity for a single user. This is one of the reasons behind the idea of the international partnership where one reactor can be used by multiple international users and research can be conducted both on bilateral and multilateral basis.

Speaker: Mr Vyacheslav Pershukov (ROSATOM)

Material FR17-578-Pershukov-Empty-Nopaper.docx

111 Verification of the neutron diffusion code AZNHEX by means of the Serpent-DYN3D and Serpent-PARCS solution of the OECD/NEA SFR Benchmark

AZNHEX is a neutron diffusion code for hexagonal-z geometry currently under development as part of the AZTLAN project in which a Mexican platform for nuclear core simulations is being developed. The diffusion solver is based on the RTN0 (Raviart-Thomas-Nédélec of index 0) nodal finite element method together with the Gordon-Hall transfinite interpolation which is used to convert, in the radial plane, each one of the four trapezoids in a hexagon to squares. The main objective of this work is to test the AZNHEX code capabilities against two well-known diffusion codes DYN3D and PARCS. In a previous work, the Serpent Monte Carlo code was used as a tool for preparation of homogenized group constants for the nodal diffusion analysis of a large U-Pu MOX fueled Sodium-cooled Fast Reactor (SFR) core specified in the OECD/WPRS neutronic SFR benchmark. The group constants generated by Serpent were employed by DYN3D and PARCS nodal diffusion codes in 3D full core calculations. A good agreement between the reference Monte Carlo and nodal diffusion results was reported demonstrating the feasibility of using Serpent as a group constant generator for the deterministic SFR analysis. In order to verify the under development solver inside AZNHEX, the same Serpent generated cross sections sets for each material were exported to AZNHEX format for four different states (as in DYN3D and PARCS): a) a reference case in which the multiplication factor (keff) is the compared value, b) the Doppler constant (KD), c) the sodium void worth, and d) the total control rod worth. Additionally, the radial power distribution was also calculated. The results calculated with AZNHEX showed also a quite good agreement in the direct comparison with DYN3D (-66 pcm in keff) and PARCS (-109 pcm in keff) and therefore against the Serpent reference solution (-194 pcm in keff). As AZNHEX is still under development further improvements will be implemented and new tests will be carried out, but so far the results presented here give confidence in the development.

Speaker: Dr Armando Miguel Gomez Torres (Instituto Nacional de Investigaciones Nucleares)

Material AZNHEXvsDYN3D_final_rev.pdf

112 Benchmark Evaluation of Dounreay Prototype Fast Reactor Minor Actinide Depletion Measurements

Historic measurements of actinide samples in the Dounreay Prototype Fast Reactor (PFR) are of interest for modern nuclear data and simulation validation. Samples of uranium, neptunium, plutonium, americium, curium, and californium isotopes were irradiated for 492 effective full-power days and radiochemically assayed at Oak Ridge National Laboratory (ORNL) and Japan Atomic Energy Agency (JAERI). Limited data were available regarding the PFR irradiation; a six-group neutron spectra was available with some power history data to support a burnup depletion analysis validation study. Under the guidance of the Organisation for Economic Co-Operation and Development Nuclear Energy Agency (OECD NEA), the International Reactor Physics Experiment Evaluation Project (IRPhEP) and Spent Fuel Isotopic Composition (SFCOMPO) Project are collaborating to recover all measurement data pertaining to these measurements, including collaboration with the United Kingdom to obtain pertinent reactor physics design and operational history data. These activities will produce internationally peer-reviewed benchmark data to support validation of minor actinide cross section data and modern neutronic simulation of fast reactors with accompanying fuel cycle activities such as transportation, recycling, storage, and criticality safety.

Speaker: Dr Tatiana Ivanova (OECD Nuclear Energy Agency)

Material FullPaper-PFR-MA-Benchmark.pdf

7.2 Economics of Fast Reactors

Conveners: Mr Dmitriy Tolstoukhov (ITCP «PRORYV»), Dr GILLES MATHONNIERE (CEA)

113 Fast Reactors and Nuclear Cogeneration: A Market and Economic Analysis

Fast reactors are typically considered for their potential to make optimal use of natural resources or for their potential to minimize the amount and level of nuclear waste. The additional opportunity of fast reactors designed for cogeneration applications (i.e., production of electricity and process heat), which can bring an enormous reduction in CO2-emissions, is made possible by the elevated temperatures characterizing the primary circuit of such reactors, compared to traditional light water reactors. This article will provide a state-of-the-art overview on the cogeneration market with emphasis on opportunities for lead, gas, and sodium fast reactors, summarize recommendations for these fast reactor systems and their interfaces with a cogeneration application, and discuss the results of a top down cost estimate for a lead fast reactor system with a typical cogeneration application. The economic analysis clearly shows that coupling a small or medium sized fast reactor to a cogeneration application seems attractive.

Speaker: Mr Michele Frignani (ANN)

Material Paper_82_IAEA_FR17_Cogeneration_reviewed.docx

114 Providing the competitiveness of nuclear energy in the implementation of PRORYV project

In the report said about reducing the share of nuclear power generation in the world. In conditions inter-fuel competition highest growth rates show renewables. The key to advancing the development of nuclear energy is to ensure the competitiveness of NPPs whith solving systemic problems. Modern NPPs with LWR in an open NFC, practically exhausted the potential of improving the competitiveness. In the report considered the comparative competitiveness of NPPs and power plants whith fossil fuel, renewable energy sources. All calculations are made for different countries by monitoring data of technical and economic. Singly considered the comparison conditions for the Russian Federation. Considered criteria of competitiveness for NPPs, allowing to ensure the effective development of nuclear power, taking into account of improving the technical and economic performance of alternative generation. Fixed the requirements for the technical and economic parameters of NPP FRs and closed NFC. The data on the assessment of the achievement the required indicators for NPPs with FRs and closed NFC on the basis of actual work within the PRORYV project. Considering the model of PRORYV project management, identify areas to improve efficiency of the risk-management during creating innovative facilities.

Speaker: Mr Dmitriy Tolstoukhov (ITCP «PRORYV», Russian Federation)

Paper IAEA-CN-245-98.docx IAEA-CN-245-98.pdf

115 COMPARATIVE ANALYSIS OF ELECTRICITY GENERATION FUEL COST COMPONENT AT NPPs WITH WWER AND BN-TYPE REACTOR FACILITIES

The fuel cost component (FCC) of electricity generation is defined as a specific indicator - the cost of 1 kWh of electricity produced. This value is obtained as the levelized (discounted) nuclear fuel cost value, generally beginning with natural uranium procurement and ending with spent fuel management, normalized to the total electric energy generated over the nuclear power plant lifetime. I.e. the result is the FCC average value over the entire lifetime. The methodology of levelized FCC calculation is based on the concept of taking into account the disparity in the value of money, referring to different moments of time, and thus, the possibility of technical and economic comparison of projects with significant lifecycle. The nuclear fuel life cycle is known to normally cover a period of 50-100 years. The paper describes the basic essential methodological and factual materials for the fuel component calculation for NPPs with fast and thermal reactors. However, these reactors are expected to be in the NE system, together with the nuclear fuel cycle facilities. In such a case, as is well known, plutonium is a link between thermal and fast reactors. The calculations were performed for the fast reactor BN-1200 in version with MOX-fuel, as well as for the WWER-TOI thermal reactor. The calculations have shown that at constant prices for natural uranium the values of levelized FCC with BN and WWER reactors are sufficiently close to each other. With regard to the escalation of prices for natural uranium, the levelized FCC for the entire life cycle of nuclear power plants with a natural uranium fuel thermal reactor significantly increases depending on the MOX fuel fraction in the core inventory, whereas for fast reactor NPPs it remains constant and much lower. The calculations have shown that for the fast reactor the fuel fabrication cost makes the main contribution to FCC, and for the thermal reactor – it is the cost of natural uranium and its enrichment.

Speaker: Dr Viktor Dekusar (Laboratory of Nuclear Fuel Cycle system analysis taking into account National and International tendencies)

Paper Contribution_ID435_Fuel-cost.docx

116 How to take into account the fleet composition in order to evaluate Fast Breeder Competitiveness

Fast reactor competitiveness is usually examined by comparing fast reactors and light water reactors (LWRs) LCOE (levelized cost of electricity). As fast reactors have an investment cost higher than LWRs, their kWh production cost is higher than that of LWRs (with the natural uranium current price) and their competitiveness will thus take place when the increase of the natural uranium cost will be enough to counterbalance their additional investment cost. In fact, the interest of the fast reactors is to allow the implementation of sustainable nuclear power fleet not consuming natural uranium, but only depleted uranium which we have, unlike the natural uranium, considerable resources. The real objective is not to build a reactor not consuming natural uranium, but to have a sustainable nuclear fleet consuming depleted uranium only. For this purpose, fast reactors are necessary, but it doesn’t imply that the whole fleet will be made of fast reactors only. It is a possibility, but not the only one. Indeed, these break-even(isogenerator) reactors can become fast-breeder reactors (FBRs) by using blankets and available plutonium surplus can be used in other reactors consuming plutonium, but not consuming natural uranium. For example, we can build a fleet including fast-breeder reactors but also LWRs like EPR with a 100 % MOX load. The different shares of the two reactor types will be defined by the balance between plutonium produced in FBRs and plutonium consumed in LWRs. By doing this, the additional investment cost of FBR is diluted because it does not concern more than a part of the sustainable fleet (as a matter of fact the cost of the kWh produced by a100 % MOX EPR is not very different from that of the UOX EPR with a current natural uranium cost). Considering the effect of such a fleet including both FBR and 100% MOX LWR this study suggests that increasing the breeding ratio for FBR and increasing the conversion ratio of LWR by considering high conversion ratio LWR could be economically efficient. This study shows that a corrective factor depending on the fleet composition should be taken into account in the FBR overcosts in order to examine its competitiveness.

Speaker: Mr GILLES MATHONNIERE (CEA)

Paper FR17_Vf_Id_296_Vf.docx

117 Equipment cost estimation for pilot demonstration lead-cooled fast-neutron reactor BREST-OD-300

One of the main problems in determining the investment in the construction of new nuclear facilities at the design stage is the cost estimation of non-standard equipment. The novelty and lack of experience are the features of early stage of the project. The uniqueness of the projected facilities makes it impossible to use catalogues and price-lists. It is required different approaches to estimate the cost of such equipment. The methods of assessing the cost of projected non-standard reactor equipment can be divided into two main groups: analogy and resource or engineering build-up methods. Additional cost estimating methods include parametric methods and corrective amendments. Methods and approaches of non-standard equipment cost estimation are differ from stage to stage, allowing to obtain a more accurate result. It is necessary to take into account features of the application of a method according to the stage of design work. The essence of different approaches of equipment cost estimation is disclosed, advantages and disadvantages of different methods are analyzed, guidance on the applicability depending on the specific conditions of evaluation is provided. The comparison of cost indexes is made. Economies of scale and learning curve must be taken into account. Rosatom Production System (RPS) is proposed as one of additional equipment cost estimation method. Expected accuracy can be determined by cost estimate classification matrix according to Association for the Advancement of Cost Engineering International (AACE International) recommended practice. All available methods and tools must be used to estimate cost of non-standard equipment at the design stage. Obtaining close results by means of different methods indicates reliable estimates of equipment cost. Iterative approach to the assessment of the main BREST-OD-300 reactor equipment, based on decomposition and structural analysis, is presented taking into account the economic characteristics of potential manufacturing plants It is important to develop and improve cost estimation methods taking into account best practices to achieve economic competitiveness.

Speaker: Mr Nikolay Molokanov (ROSATOM, JSC “NIKIET”, MEPhI)

Paper FR-17_Paper_Equipment_cost_estimation_BREST-OD-300_(Goltsov_Molokanov_ROSATOM_NIKIET).docx

12:00 PM Lunch Break

1.4 CORE AND DESIGN FEATURES - 1

Conveners: Mr Dmitry Klinov (IPPE), Mr HIROKI HAYAFUNE (Japan Atomic Energy Agency)

118 SELECTION OF A LAYOUT FOR THE BN-800 REACTOR HYBRID CORE

The initial loading of BN-800 is mainly made of the uranium oxide fuel and partially of MOX fuel subassemblies (16% of the total quantity), which were fabricated using both the pellet technology and vibro-packing technology. With account of this specific completing process, such core is called the hybrid core. The core layout was selected to simplify the future transition from the hybrid core to the core fully loaded with the MOX fuel and to maximally adapt BN-600 uranium fuel subassemblies fabrication to BN-800 uranium fuel subassemblies fabrication. The hybrid core uses three types of fuel subassemblies with the different content of fissile material (degrees of enrichment) to retain fuel enrichment limits and fuel column height the same as in the MOX core. To ensure compatibility of uranium fuel subassemblies and MOX subassemblies, the plutonium content of the MOX fuel was defined to retain the same physical efficiency for respective types of fuel subassemblies. To minimize distortion of the power field, the MOX fuel subassemblies are arranged in the periphery of the hybrid core (within the high enrichment zone). Fuel subassemblies with the MOX pellet fuel are arranged in the first row, and fuel subassemblies with the vibro-packed MOX fuel are arranged in the peripheric row, under less severe operation conditions. The report discusses the main prerequisites to develop the hybrid core, describes the core design, and gives information about main operation characteristics.

Speaker: Mr Artem Kuznetsov (JSC “Afrikantov OKBM”)

Paper IAEA-CN-245-406.pdf

Slides Kuznetsov

119 ALLEGRO Core Neutron Physics Studies

Status of neutron-physical analysis of ALLEGRO - demonstrator of the gas cooled reactor is characterised at this article. Benchmarking of existing neutronic codes utilised for PWR analyses mainly, is first task, solved at running projects. As there are available no neutronic experiments with He coolant at fast spectrum, code to code comparison was selected as first stage of validation process. First ALLEGRO oriented neutronic benchmark was split into two phases. Definition, solution and partial conclusions of first phase concentrated on pin calculation - Methodological benchmark with simplified geometry for the group constant generating tools - are described at the article. Definition of second phase oriented on assembly calculations and its first evaluations are treated as well. Evolution of ALLEGRO core evolution is driven by two factors - problems with DHR proportional to power density and by better availability of UOX fuel for first cycles (in comparison with MOX). First round of calculations oriented on fuel and power density optimisation including resulting direction of core modifications is characterised in the paper.

Speaker: Dr Petr Dařílek (VUJE Inc., SK91864 Trnava, Slovak Republic)

Paper Darilek_ALLEGRO_Core_Neutron_Physics_Studies.pdf

120 BN-800 core with MOX fuel

One of the main objectives for BN-800 development is to master the closed nuclear fuel cycle technologies using the mixed uranium-plutonium (MOX) fuel. The initial loading of BN-800 with the hybrid core mainly made of the uranium oxide fuel includes the limited quantity (16%) of MOX fuel subassemblies fabricated at experimental production facilities of Mayak Production Association and JSC SSC RIAR. The core will be completely fueled with the MOX fuel fabricated at the Mining and Chemical Combine by step-by-step replacement (three refuelings) of fuel subassemblies in the hybrid core with MOX fuel subassemblies. To flatten power distribution, the core uses three types of fuel subassemblies with the different plutonium content in the fuel. A technique to adjust plutonium enrichment in fuel depending on the fuel isotope composition makes it possible to fabricate plutonium-based MOX fuel with a wide range of isotope compositions and retain core operation parameters within the design limits. To reduce the sodium void reactivity effect, the fuel subassembly design has an upper sodium cavity and absorbing shield made of natural boron carbide. In the BN-800 core, the ChS-68 steel is used for fuel rod claddings the same as for standard fuel rods in BN-600. Fuel rods with such cladding continue to operate up to the damaging dose of  90 dpa, which as applied to BN-800 corresponds to the average burnup of 66 MW day/kg for the fuel unloaded. Prospects to increase BN-800 fuel burnup are connected with a transition to the more radiation-resistant steel EK-164 for fuel rod cladding and later on, to ferritic-martensitic steels and oxide-dispersion-strengthened ferritic steels.

Speaker: Mr Artem Kuznetsov (JSC “Afrikantov OKBM”)

Paper IAEA-CN-245-405.pdf

Slides Kuznetsov

121 Physics Investigation of a Supercritical CO2–cooled Micro-Modular Reactor (MMR) for Autonomous Load-Follow Operation

This paper presents a physics study for a passive autonomous load-follow operation in a supercritical CO2-cooled micro-modular reactor (MMR). The proposed long-life 36.2 MWth MMR is a super-compact, fully-integrated, and truck-transportable fast reactor module in which all components are integrated in a single pressure vessel. UC fuel is considered to maximize the fuel inventory and to enhance neutron economy. The core lifetime is designed to be over 20 years without any refueling. To minimize the excess reactivity, a replaceable fixed absorber (RFA) is used and the resulting excess reactivity is found to be less than 1 $ during the whole lifetime of the core. For demonstration of a passive autonomous operation of the MMR, analyses for the reactor startup from initial CZP to HFP and daily load-follow operation are performed. In a passive autonomous load-follow operation, the reactor power is automatically controlled by the feedback reactivity only, which mainly depends on the fuel and coolant temperatures. For favorable passive autonomous load-follower operations, the core expansion feedback reactivity is also taken into account in this paper. All neutronics calculations are performed using the continuous energy Monte-Carlo Serpent code with the ENDF/B-VII.1 library.

Speaker: Mr Hwan Yeal YU (KAIST)

Paper FR17-mmr_final.pdf

3.4 Sodium leak/fire and other safety issues

Conveners: Mr IURII SHVETSOV (Private institution «Innovation and technology center for the «PRORYV» project»), Dr Yasushi OKANO (Japan Atomic Energy Agency)

122 Numerical – experimental research in justification of fire (sodium) safety of sodium cooled fast reactors

On the basis of the normative documents requirements on fire safety of nuclear power facilities the build concept and the composition of fire protective system in fast reactors premises with sodium equipment are presented. The main purpose of sodium fire safety system is to protect the technological areas of nuclear power plants with sodium cooled fast reactors from hazards of sodium fire. The hazards of sodium fire are: increasing the pressure and temperature of gas environment in emergency rooms, raise of building structures temperatures upon burning of sodium. Another hazard of sodium fire is spreading of sodium aerosols in the premises of the plant which are harmful to human health. The numerical justification of the sodium fire extinguishing system effectiveness in case of possible accidents with sodium burning in certain areas of fast reactor is performed. During the formation of the fire safety conception on sodium cooled fast reactors the special attention is focused on the nature of the outflow and sodium burning and on the sodium leakage limitation. The numerical and experimental researches aimed at the performance possibility and efficiency of system for early detection of leaks and sodium burning based on automatic smoke fire detectors VESDA are performed. The issues of jet outflow and spray burning of the sodium coolant and related problem with increasing of the gas pressure and temperature in the room are considered. In the framework of these issues the results of experimental works for sodium spray burning made by French experts are considered. The results of the analysis and processing of experimental data are presented. The method is developed for the gas pressure raising calculation in the room based on processing of experimental data. The main experimental results with sodium flow through the defects in the pipeline under the insulation are presented. A possibility is shown for safe localization of the jet outflow and sodium spray burning in the presence of pipelines and equipment insulation and cladding based on this experimental data.

Speaker: Mr Olga Myazdrikova (Leypunsky Institute for Physics and Power Engineering (IPPE))

Material Paper_IAEA-CN-245-102_correction.docx

123 Identification of important phenomena under sodium fire accidents based on PIRT process

Since sodium has high chemical reactivity with oxygen and moisture, sodium fire accident is one of key issues in sodium-cooled fast reactor (SFR) plants when the sodium leaks out of a coolant circuit. In order to evaluate the consequence of the sodium fire event numerically, JAEA has developed sodium fire analysis codes such as SPHINCS and AQUA-SF. This paper describes a PIRT (Phenomena Identification and Ranking Table) process for a sodium fire event. The present PIRT is aimed to utilize for validation and improvement of the sodium fire analysis codes. Because a sodium fire accident in an SFR plant involves complex phenomena, various figures of merit (FOMs) for importance ranking could exist in the PIRT process. Therefore, the FOMs are specified through factor analysis. Associated phenomena in a sodium fire event are identified through the element- and sequence-based phenomena analyses. Then importance of each associated phenomenon is evaluated by considering the sequence-based analysis of associated phenomena related to the FOMs. Finally, we have established the ranking table through the factor and phenomenon analyses.

Speaker: Dr Mitsuhiro Aoyagi (Japan Atomic Energy Agency)

Material FR17_paper_draft_aoy.pdf

124 Learning from 1970 and 1980-Era Sodium Fire Experiments

The original sodium fast reactor concepts date back to the 1950s and 1960s, and there were large programs in France, Japan, Russia, and the United States in the 1970s and 1980s to design and build commercial-scale SFRs. Many of these programs were abandoned in the 1990s, but a considerable amount of work was done prior to that in order to demonstrate the concepts, and to support the safety cases for the commercial prototypes. There is renewed interest in the sodium fast reactor technology with the advent of Gen-IV concepts developed though the GIF initiative, and safety standards for these designs are higher than they were for the original Gen-III designs. Evaluating the effects of sodium fires in the containment vessel would be an essential part of any modern safety evaluation because beyond the risk of thermal load of the structures and containment overpressure, and the fact that they would be a source of airborne fission products. Validated computational tools able to simulate the in-containment phenomenology are then necessary for a reliable estimation of the source term to the environment in the case of an accident. Many of the fundamental safety questions have already been explored in the past, however, and there is a huge amount of value in re-visiting the experimental and theoretical work that has already been done. This paper will discuss an effort to retrieve and re-examine experimental work that was carried out in the 1970s and 1980s at the Cadarache research center in France. Different sodium fire programs will be outlined, and will be linked to new theoretical analyses and modeling efforts. As examples, studies on theoretical developments for sodium spray fire combustion, pool fire combustion, combustion product aerosolization, and fission product emission phenomena have been enabled from access to this data. The experimental results have also been used in studies to validate sodium fire modules in the severe accident code ASTEC-Na. This paper will underline how important it is to preserve the knowledge that was generated in the past, and will outline some of the ways that it can still be applied today.

Speaker: Dr Luke Lebel (Institut de Radioprotection et de Sûreté Nucléaire)

Material PaperHistoricalExpLebel.docx

125 SEISMIC SLOSHING EFFECTS IN LEAD-COOLED FAST REACTORS

Pool-type configuration of LFR (lead-cooled fast reactor) primary systems allows for simple and economic reactor design solutions. However, partially-filled heavy liquid pools pose seismic safety concerns related to sloshing. Violent sloshing can lead to structural failures, gas entrapment and potential core voiding. Seismic isolation systems are used to reduce the mechanical stresses in structures, but its effect on sloshing must be clarified. This paper describes a numerical CFD (computational fluid dynamics) study of lead sloshing in ELSY reactor. The motion of free surface is modeled using a VOF (volume of fluid) method. A fixed base reactor and seismically isolated reactor cases are modeled using synthetic earthquake data produced in SILER project. Verification and validation of the numerical model is presented. The adverse effects of seismic isolation system in terms of sloshing-induced hydrodynamic loads and gas entrapment are demonstrated. Furthermore, influence of geometry on sloshing behavior has been discussed. A mitigation solution using flow restrictions is proposed and analyzed.

Speaker: Mr Marti Jeltsov (KTH Royal Institute of Technology)

Material FR_17_355_Seismic_sloshing_effects_in_LFRs_Rev1.pdf

126 Evaluation of multiple primary coolant leakages accidents in Monju with consideration of passive safety features

To maintain sodium level inside reactor vessels above cores is essential to keep core-cooling for sodium leakage accidents in loop-type sodium-cooled fast reactors. In the loop-type prototype fast reactor Monju which has three primary heat transport systems (PHTSs), a single coolant leakage accident in a PHTS has been taken into account as a design basis accident (DBA). On the other hand, it is important to investigate that another primary coolant leakage would occur after the first coolant leakage accident as a design extension condition (DEC). In this presentation, we evaluate multiple primary coolant leakages accidents in Monju with consideration of passive safety features. Concretely speaking, the flow rate and the amount of leakages can be reduced by the effect of the decrease of the cover gas pressure due to lowering reactor coolant level (negative pressure effect). The sodium coolant level necessary for the decay heat removal can be maintained, taking account of the negative pressure effect and other measures.

Speaker: Mr Kazuo YOSHIMURA (Japan Atomic Energy Agency)

Paper IAEA-CN-245-290_20170420.pdf

5.10 Fuel Modeling and Simulation

Conveners: Mr Marc LAINET (French Alternative Energies and Atomic Energy Commission (CEA)), Dr Mikhail Veshchunov (IAEA)

127 PROBLEMS OF CALCULATION MODELLING OF NITRIDE FUEL PERFORMANCE: DRAKON CODE

The main life limiting factor of nitride fuel pins at high burn-up is fuel cladding mechanical interaction (FCMI) leading to strong deformation or even cladding destruction. The consequences of FCMI depend on fuel and cladding swelling rates, cladding creep rate, cladding long-term stress rupture etc. The calculation modelling problem arise from not enough data on nitride out-of-pile properties and in-pile behavior in dependence on plutonium content, fuel density, irradiation temperature, as well as lack of reliable data on irradiation steel cladding properties. Within the framework of the PRORYV project a comprehensive program for calculation and experimental studies of mixed nitride fuel for BN-1200 and BREST-OD-300 reactors has been designed to provide the required data. The DRAKON code is designed for numerical simulation of temperature and stress-strain state of fast reactors nitride fuel pins. The code verification was based on PIE data obtained after irradiation of standard FA of BR-10 reactor (uranium mononitride) and experiment BORA-BORA in BOR-60 reactor (mixed nitride fuel). The calculation results are in good agreement with the experimental data. It is planned to continue the code verification using the result of PIE of experimental FAs with nitride fuel, which are being irradiated now in BOR-60 and BN-600 reactors. Currently DRAKON code is used to study performance of the experimental nitride fuel pins of BN-600 reactor and to analyze the PIE results for such pins. As an example, the calculation results of experimental fuel pins with low-swelling FM steel cladding are given in the presentation. It is shown that there are two major limiting factors: 1) FCMI stress in the lower cladding sections with the "cold" fuel where cladding creep rate is negligible; 2) cladding damageability due to FCMI and fission gas pressure in the upper "hot" sections, where high-temperature strength of FM cladding is low. The fuel rod performance is limited by one of these factors depending on actual irradiation parameters.

Speaker: Mr Evgenii Marinenko (Institute for Physics and Power Engineering)

Material IAEA-CN-245-63.docx

128 3D SIMULATION IN THE PLEIADES SOFTWARE ENVIRONMENT FOR SODIUM FAST REACTOR FUEL PIN BEHAVIOR UNDER IRRADIATION

In the framework of the basic design of ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration) the GERMINAL fuel performance code is developed in the PLEIADES software environment. In order to improve one dimensional modelling of GERMINAL, a 3D simulation for the SFR fuel pin behavior under irradiation has been proposed. The 3D model represents a single pellet fragment and its associated piece of cladding. The scale transfer between this single fragment model and the fuel pin scale is achieved through appropriate boundary conditions given by GERMINAL results. The 3D thermo-mechanical computation scheme is implemented in the LICOS code of the PLEIADES platform. In this approach, chemo-physical aspects are still computed by the GERMINAL code and are introduced in the 3D computation scheme as some input data in a two-step procedure. First studies have been achieved in order to analyze pellet-to-cladding gap closure mechanisms at the beginning of irradiation. Two mechanisms of fuel relocation are simulated through the 3D simulation. The first one is linked to the hourglass shape of the fragmented fuel pellet under thermal gradient, and the second one is induced by the mass transfer due the central hole formation and fuel restructuration. According our results, the gap closure rate given by the GERMINAL empirical model can be understood. The 3D coupling formulation has now to be extended to the mass transfer equations in order to improve the results.

Speaker: Dr bruno MICHEL (CEA/DEN/DEC/SESC/LSC)

Material FR17_IAEA-CN-245-223-3D_bruno_MICHEL.pdf

129 Current status and progression of GERMINAL fuel performance code for SFR oxide fuel pins

A fuel performance code for SFR oxide fuel pins, GERMINAL, is developed by CEA within the PLEIADES simulation framework. The present main goal of GERMINAL is to meet the needs of the design studies of ASTRID, the future Advanced Sodium Technological Reactor for Industrial Demonstration in France. Recent works have been conducted to improve the modelling of different physical mechanisms having a strong influence on the design criteria evaluation. Thus, the formulation of the fuel pellet fragments relocation model has been revisited, by introducing a dependence to the thermal gradient inside the pellet. The description of this mechanism represents a key point to evaluate the pellet-to-cladding gap closure and the margin to melting at beginning of life. Another evolution concerns the pellet-clad mechanical interaction. The ability to simulate a stronger interaction for fuel pins with a higher filling fraction has been acquired with a focused work on fuel mechanical behavior. A stronger mechanical interaction may also happen with lower power operating conditions and a cladding material remaining stable under irradiation. Moreover, the description of the thermochemistry of oxide fuel is currently being improved by coupling GERMINAL with the OpenCalphad thermodynamic calculation software. In doing this, the goal is to obtain a better prediction of the amount of volatile fission products being transported outside the fuel pellet, and then contributing to the “Joint Oxyde-Gaine” formation. With refined estimations of JOG volume and composition, we expect further to improve the evaluation of heat transfer through pellet-to-cladding gap at high burn-up, and also a more mechanistic description of cladding corrosion due to released fission products. These works are based on a systematic comparison of calculation results to post-irradiation measures, by integrating progressively additional objects to our validation base. This process leads to a wider validity range targeting ASTRID design, and brings out new working perspectives.

Speaker: Mr Marc LAINET (French Alternative Energies and Atomic Energy Commission (CEA))

Material IAEA-CN-245-222.docx

130 BERKUT – Best Estimate Code for Modelling of Fast Reactor Fuel Rod Behavior under Normal and Accidental Conditions

The advanced version of code BERKUT designed for mechanistic modelling of oxide and nitride single fuel rod behavior under normal and accidental conditions of liquid metal cooled fast reactor operation, is under development at IBRAE RAN during the last five years in frame of “Codes of new generation” project included into the “BREAKTHROUGH” (or “PRORYV”) project. The code models are grounded on the contemporary understanding of mechanisms governing the most important processes in fuel rods under irradiation, which substantially enhances the predictive ability of the code in comparison with the engineering analogs. The code is the multi-scale one simulating the processes characterized by the range varied from 1nm to 1 m. At the micro-level the code describes evolution of fuel micro-structure in the fuel grain scale: • vacancy/interstitial field, nucleation and development of dislocation network and gas filled porosity, • fission product generation, their radioactive transformations, transport and release out of fuel grains, • formation of chemical compounds, the fuel phase composition. At the meso-level the code simulates the processes in the fuel pellet scale: • mass transfer of fission products, oxygen or nitrogen within pellets, • evolution of as-fabricated porosity and formation of columnar grains, • fission product release by recoil and knockout mechanisms. At the macro-level the code describes thermomechanical behavior of the fuel rod as a whole: • heat transfer within the rods and heat exchange with the coolant, • temperature distribution in fuel, fuel-cladding gap and cladding, • evolution of the stress-strain state of fuel and cladding, • fission gas composition and gas pressure within the cladding. The code models describing the processes in oxide fuel, which are common for thermal and fast neutron reactors, have been validated against extensive experimental data set found in the literature. Some particular microscopic parameters have been defined through the theoretical estimates. The calculations have been performed simulating oxide and nitride fuel rod behavior in BN-600 and BOR-60 reactors. Analysis of the calculation results and their comparison with the data of the post-reactor fuel rod examination has demonstrated that BERKUT describes satisfactorily the fuel and cladding geometry changes, fission gas release as well as porosity profiles and fission product concentration profiles within fuel pallets. The calculation results obtained allow to make a conclusion that mechanistic fuel rod codes can be used both for safety justification and to predict ways of achieving the specified fuel rod characteristic.

Speaker: Mr Andrey Boldyrev

Material tarasov tarasov

131 EXPERIENCE AND APPLICABILITY OF HIGH DENSE METAL URANIUM IN ADVANCED BN-REACTORS

To guarantee an inherent safety in advanced BN-reactors the breeding ratio of its active core (BRC) must have the meaning of BRC ≥ 1,0. It could be reached in heterogeneous oxide-metal cores of various types in use of metal uranium as the fertile components in the proportion of MOX:U ≈ 2:1. We obtained the experience of manufacturing the fertile columns of various types from the metal uranium, the experience of manufacturing and irradiation in fast reactors BOR-60 and BN-350 of full-size elements (FE) and fuel assemblies (FA) that have such columns (4010 elements in part of 108 fuel assemblies). Besides the obtaining of the inherent safety in advanced BN-reactors with the heterogeneous oxide-metal cores of various types (by FA-heterogenization of the core, IFAH – by intra FA-heterogenization, IFEH – by intra fuel elements heterogenization) we could achieve considerable additional economic and ecological preferences. Among them there are the increase of the admissible average burnup of MOX-fuel by ≈20%, the decrease of the mass of manufactured and consumable Pu-containing MOX-fuel by ≈30%, the decrease of consumable Pu-containing FE or Pu-containing FA by ≈30%, etc.

Speaker: Prof. Iulian Golovchenko (Russian Federation)

Paper FR17-396.doc

5.4 Advanced Fast Reactor Cladding Development II

Conveners: Dr Tai Asayama (Japan Atomic Energy Agency), Mr V. Chuyev

132 Modeling of Processes in Austenitic Steel Produced Under Irradiation in Fast Reactors and Possibilities of Model Practical Application

Nowadays austenitic stainless steels are used as a cladding material for BN-600 and BN-800 reactors. Examinations carried out at the end of fuel element service life give the information about cladding state. On the basis of the examination it is necessary: 1 – to determine residual life and 2 – to find the way for its extension. Extrapolation methods are generally used for the first aspect. The results obtained for fuel elements of different fuel assemblies at attaining different damage doses and fuel burn-ups are used. As a rule the results are limited by linear extrapolation. The prediction accuracy is quite low for several reasons. The initial state of claddings from different lots (and casts) is not the same, therefore there is some error even when processes of properties changes linearly depend on irradiation parameters (for example, damage dose). Moreover, the dependence of some processes, radiation-induced swelling in particular, on dose and temperature is quite nonlinear. Therefore linear extrapolation is unacceptable. Extrapolation used for the second aspect almost gives no results as the characteristics to be correlated are not defined. At JSC "INM" a technical description of the processes occurring in metal materials under irradiation has been developed for a long time. Description of point defect formation (vacancies and interstitials) is the key concept. All further microstructural changes are determined by the formation intensity, migration and interaction with other microstructural elements (impurity atoms, dislocations, grain boundaries and other sinks). A machine for quantitative description of point defect migration and concentration has been developed and is used for austenitic stainless steels. Based on the developed theoretical concepts different stages of structural changes, radiation-induced swelling in particular, as well as the effect of structural changes on physical and mechanical properties have been modeled. These models were used to predict changes in material structure and properties of the claddings operated in BN-600 reactor core. The paper aims to show the developed at JSC "INM" models of changes in austenitic steel structure and properties under irradiation in fast reactors and to demonstrate their application for BN-600 reactor claddings.

Speaker: Mr Alexander Kozlov (Joint Stock Company "Institute of Nuclear Materials")

Material IAEA-CN-245-95_Rev.pdf

133 Preliminary Inspection of Spent Fast Reactor Fuel Claddings

Electrical potential testing which is a primary nondestructive testing method is used at JSC "INM" hot cells as an incoming inspection of spent BN-600 and BN-800 reactor fuel elements. Electrical resistivity curves demonstrate the level of the fuel element defectiveness and help to work out a cladding dismantling plan for further post irradiation materials examination of the cladding problem areas. Electrical potential testing through the electrical resistivity distribution profile enables immediate evaluation of cladding structural changes resulting from material swelling under irradiation. Depending on the damage dose it is possible to evaluate and compare values of cladding swelling by resistograph images. Theoretical dependence and experimental results showing the correlation between material radiation-induced swelling as well as cladding corrosion thinning and the change of electrical resistivity are shown. The effect of radiation and technological defects on electrical resistivity change of the claddings made of ChS-68, EK-164, and EP-450 steels are discussed.

Speaker: Mr Vyacheslav Shikhalev (Joint Stock Company "Institute of Nuclear Materials")

Material IAEA-CN-245-107_Rev.pdf

134 OPERABILITY VALIDATION OF FUEL PINS WITH CLADDINGS MADE OF EK164-ID STEEL IN THE BN-600 REACTOR

To ensure the increased fuel burn-up in the BN-600 and BN-800 reactors, EKA64-ID steel is going to be used as a fuel rod cladding material because it has bigger index of radiation resistance (swelling and creeping) in comparison with the used ChS68-ID steel. To introduce this steel, an irradiation examination of experimental FSAs is needed to be performed. Owing to the irradiation examination, experimental data will be obtained to validate FSA operability and a database on properties of the steel will be updated thanks to which computational codes will be verified. Tests are performed as per appropriate procedure in cooperation with operating organization and Rostechnadzor experts. By now, reactor examination of 14 experimental FSAs has been successfully performed in the BN-600 reactor. The maximum achieved irradiation parameters are as follows: the fuel burn-up is ~ 14 % h.а., the damaging dose is ~ 100 dpa. The examination is planned to be continued for using the steel as fuel rod cladding with higher parameters: the fuel burn-up should be 14.8% h.а., the damaging dose should be ~ 112 dpa. Activities aimed at improving the quality of cladding tubes both in the stage of fuel rod cladding manufacture and in the metallurgic stage of tubing stock manufacture are performed simultaneously with manufacture and irradiation of the experimental FSAs. Results of these experimental activities will be used to validate operability of fuel rods made of this steel in the initial stage of the BN-1200 reactor operation.

Speaker: Mr Sergey Belov (JSC “Afrikantov OKBM”)

Paper IAEA-CN-245-409.pdf

Slides IAEA-CN-245-409.ppt

135 Creep resistance and fracture toughness of recently-developed optimized Grade 92 and its weldments for advanced fast reactors

Optimized Grade 92 has been developed at Oak Ridge National Laboratory in support of the USA Sodium-cooled Fast Reactor program. Composition modification and processing optimization successfully achieved the development of optimized Grade 92 with desired microstructures for superior properties. A variety of properties have been assessed for optimized Grade 92, which include tensile, creep, fatigue, creep-fatigue, impact and fracture toughness, weldability, thermal aging resistance, and sodium compatibility. This paper focuses on presenting the results of creep and fracture toughness tests of optimized Grade 92 and its weldments. Compared to the literature data of Grade 92 and similar 9Cr ferritic-martensitic steels, optimized Grade 92 exhibited significantly enhanced creep resistance, together with superior or comparable fracture toughness. Creep rupture ductility of the ruptured samples is discussed by comparing to the reference steels. Samples extracted from tungsten-inert-gas fabricated weldments showed slight reductions in creep life and creep strength compared to the base metal of optimized Grade 92. The reductions, however, are noticeably smaller than that of the reference steels. Satisfactory fracture toughness was observed for the weldments of optimized Grade 92. Hardness measurements and microstructural characterization following the tests shed light on the superior properties of optimized Grade 92 and its weldments. The enhanced properties are expected to favor the application of optimized Grade 92 for advanced fast reactors.

Speaker: Dr Lizhen Tan (Oak Ridge National Laboratory)

Material FR17-Opt_G92_Tan_rev.docx

6.4 Neutronics – 2

Conveners: Mr Evgeny Seleznev (NUCLEAR SAFETY INSTITUTE OF RUSSIAN ACADEMY OF SCIENCES), Mr Robert Jacqmin (CEA- Cadarache)

136 Stability Analysis of a Liquid Metal Cooled Fast Reactor

Under specific transients, fast reactor cores often show significant deviation in their power distribution which leads to spatial instability. As a quantitative indication of these decoupling characteristics, the λ-mode eigenvalue separation has been frequently employed. The physical interpretation of eigenvalue separation provides a measure of the spatial neutronic coupling among various parts of a reactor and, hence is indicative of the space-time dynamic behaviour. In this paper the core-wide and regional stability of a Korean Prototype GEN-IV Sodium-cooled Fast Reactor (PGSFR) design is investigated using deterministic approaches. To calculate higher mode eigenvalues and associated eigenvectors the methodology of flux higher eigen-modes calculation was implemented into DIF3D 10.0 code and is thoroughly described in the paper. This specific DIF3D modification is denoted as DIFHH where the decontamination (or in some literature known as deflation) method was adopted as the simplest solution. In order to validate and demonstrate the performance of DIFHH code modification, the simple benchmark problem based on paper prepared by Mr. Obaidurrahman was chosen and investigated. The comparison of achieved trends and absolute values confirmed a favourable consistency between the reference and calculated results. The D/H ratio of the reactor core was identified as an indicator of the extent of core stability, therefore the present analyses include the investigation of eigenvalue separation and flux distribution of various core D/H ratios. The findings and the results are deeply discussed in the paper.

Speaker: Dr Branislav Vrban (B&J NUCLEAR ltd.)

Material Vrban_paper_FR17.docx

137 Analysis of the BFS-115-1 experiments

As part of a bilateral agreement on the study of large axially-heterogeneous oxide-fueled SFR cores, CEA and IPPE have recently performed neutron physics experiments in the BFS facility. The configurations of interest are pancake-shape cores with a split fissile column and a sodium plenum, designed to favor a high inner plutonium conversion ratio and a low sodium void worth. Separate effect tests, including local and global sodium void situations as well as various rodded cases, have been done. The measurements included reactivity effects, spectral indices, detailed reaction rate traverses, neutron importance, etc. The analysis of the experiments with Monte Carlo codes and recent nuclear data files shows the following trends: Core reactivity is predicted within 1.5 $, depending on the nuclear data file used Sodium voiding in the 91 central tubes is predicted within 0.25 $ The calculated axial reaction rate traverses match the experimental ones The weight of the simulated control rod is predicted within 10%

Speaker: Dr Robert Jacqmin (CEA, Cadarache)

Paper According to the reviewers comment, size (and, when possible, definition) of figs 5-8 has been increased

138 Physical start-up test of China Experimental Fast Reactor

China Experimental Fast Reactor (abbr. CEFR) is a pool-type sodium-cooled fast reactor in China Institute of Atomic Energy (abbr. CIAE), with a thermal power of 65MW and an electric power of 20MW. The construction started in 2000 and the first criticality was reached in July 2010. On December 15th 2014, CEFR reached full power for the first time and was successfully operated for 72 hours. During the physical start-up of CEFR, a series of tests were carried out in four aspects, i.e., fuel loading and first criticality, control rod worth measurements, reactivity coefficient measurements, and foil activation measurements. A large amount of experiment data was obtained in the process. In order to compile and reserve the experimental data in a standard and refined form, and to benefit the worldwide fast reactor society on the validation of codes and nuclear data, China Institute of Atomic Energy proposed an IAEA Coordinated Research Project, and got approved preliminarily. The specific objectives of the project lie in 4 aspects: firstly, to collect and evaluate experiment data obtained from CEFR physics start-up experiments mentioned above; secondly, to establish a simplified model of the CEFR core and give the correction factors and descriptions of associated methods; thirdly, to share the experiment data and the simplified core model with CRP participants for joint calculations and analysis; fourthly, to gather and analyze the calculation results, and to publish a benchmark analysis report. China Institute of Atomic Energy would like to take this great opportunity to express their welcome to all organizations to participate in this project!

Speaker: Mr Xingkai Huo (China Institute of Atomic Energy)

Material hxkciae_revised.docx

139 Neutronic evaluation of a GFR of 100 MWt with reprocessed fuel and thorium using SCALE 6.0 and MCNPX

A GFR core model with 100 MWt was evaluated using three different fuel compositions: conventional (U, Pu)C and two reprocessed fuels with transuranic (TRU) (Pu, Am, Np, Cm). One reprocessed by UREX+ technique and spiked with depleted uranium, (U,TRU)C, and the other reprocessed by the same technique but spiked with thorium, (Th,TRU)C. The reprocessed fuel came from a PWR standard fuel (33,000 MWd/T burned) with 3.1% of initial enrichment and left in the pool by 5 years. Some important nuclides were followed for burns and neutron absorption and kinf was evaluated 1400 days burning. Tests were also made for B4C absorber insertion and the temperature coefficient. The study concludes with an evaluation of power distribution in the core. The simulations were made comparing results of MCNPX and SCALE 6.0 programs. The goal is to validate the simulated model and evaluate the possibility to use TRU spiked with Th in a GFR conception.

Speaker: Prof. Claubia Pereira (Universidade Federal de Minas Gerais)

Material artigo-russia-enviado.docx

140 Solution of the OECD/NEA SFR Benchmark with the Mexican neutron diffusion code AZNHEX

The AZTLAN Platform project is a Mexican national initiative led by the National Institute for Nuclear Research of Mexico, which brings together nuclear institutions of higher education in Mexico: the National Polytechnic Institute, the National Autonomous University of Mexico and the Autonomous Metropolitan University, in an effort to take a significant step towards positioning Mexico, in the medium term, in a competitive international level on nuclear reactors analysis and modeling software. The project is funded by the Sectorial Fund for Energy Sustainability CONACYT-SENER and one of its main goals is to build up as well as strengthen the national development of specialized nuclear knowledge and human resources. The AZTLAN platform consists of several neutronics and thermal-hydraulics modules. Among the neutronics tools, the AZNHEX code has been developed. AZNHEX is a 3D diffusion code that solves numerically the time dependent neutron diffusion equations in hexagonal-z geometry. The diffusion solver is based on the RTN0 (Raviart-Thomas-Nédélec of index 0) nodal finite element method together with the Gordon-Hall transfinite interpolation which is used to convert, in the radial plane, each one of the four trapezoids in a hexagon to squares. In order to support and provide reliability to the platform, a stringent verification and validation (V&V) process in which the use of international Benchmarks and Monte Carlo reference solutions has been started. As a part of this V&V activities, results obtained with AZNHEX for the full-core simulations of the two nuclear cores of the OECD/NEA SFR Benchmark (a 1000 MW metallic-fueled and a 3600 MW MOX-fueled) are shown and compared with the ones obtained with the reference Monte Carlo code Serpent. The cross sections sets used in AZNHEX were also generated in a previous step with the Serpent code to maintain consistency between calculations. The obtained Results for keff, sodium void worth and control rods worth are within reasonable agreement; in the order of tens of pcms. The results presented are not only useful for the verification of AZNHEX, but also these ones help to define a well-tested methodology in order to generate cross section sets for future dynamic calculations with AZNHEX. Based on the results, the strengths and limitations of the AZNHEX code are discussed in the conclusions and a series of improvements have been identified and planned to be implemented.

Speaker: Prof. EDMUNDO DEL VALLE GALLEGOS (Instituto Nacional de Investigaciones Nucleares (On Sabbatical Leave from IPN-Mexico))

Paper AZNHEXvsSERPENT_final.pdf

8.1 Professional Development and Knowledge Management - I

Conveners: Dr Tatiana Ivanova (OECD Nuclear Energy Agency), Prof. Vladimir Artisiuk (Rosatom Central Institute for Continuing Education&Training (ROSATOM CICE&T))

141 Development and Deployment of Knowledge Management Portal for Fast Breeder Reactors

Knowledge Management is the process of creating value from an organization’s tangible and intangible assets and regarded as a significant contributing tool to enhance the performance of organization. Knowledge accumulated over decades of nuclear research, development & operation (organizational memory) have to be preserved and used for the future design, innovations and continued safe operation of nuclear plants. IGCAR's IT-enabled nuclear knowledge management system is designed as a generic, customizable framework and developed in-house fully using open-source platform and APIs. This paper describes the development and implementation of web-enabled, taxonomy based, advanced knowledge management system for effective management and utilization of the Prototype Fast Breeder Reactor(PFBR) records available in the form of Control Notes, Design Notes, Operation Notes, Experiments Notes, Specifications, Project Reports, Commissioning Documents, Test Procedures & Reports, Manuals, Drawings etc. The portal deployed acts as a gateway to FR Knowledge repository and enables collection, retrieval, preservation and presentation of knowledge assets in different forms. It also highlights the capabilities with which the system has been designed like controlled-vocabulary based organization of documents, multi-format document upload facility with meta-data, enhanced authentication and multi-level access control, advanced search and retrieval mechanism, online viewing and print requests management, dynamic reports generation facility.

Speaker: Mr Venkatesan Arasappan (Indira Gandhi Centre for Atomic Research)

Material FR17-FRKM-382_Revised.docx

142 Topical issues of training of specialists for fast nuclear power engineering and the closed nuclear fuel cycle

Nuclear education and training for innovative projects are regarded. National Research Nuclear University (MEPI) history is shown as an example of the university, which has great experience in nuclear education and training, including innovative fast reactor projects. The activity classification and steps of technical specialist training for fast nuclear power engineering are presented. The efficiency of various educational technologies, implemented by MEPhI, is discussed. The report emphasizes the role of student teaching and research work and practice in the formation of specialists. The use of professional databases and international projects in the field of fast reactor technology is discussed separately. Stages of formation of the department of "Technology of closed nuclear fuel cycle", organized in MEPhI to carry out targeted training for the "Breakthrough" project, are described.

Speakers: Mr Georgy Tikhomirov (NRNU MEPHI), Mr Vyacheslav Pershukov (ROSATOM)

Paper FR-17_paper_533.pdf

143 'EURATOM SUCCESS STORIES’ IN FACILITATING PAN-EUROPEAN E&T COLLABORATIVE EFFORTS

The European Atomic Energy Community (Euratom) Research and Training framework programmes are benefitting from a consistent success in pursuing excellence in research and facilitating Pan European collaborative efforts across a broad range of nuclear science and technologies, nuclear fission and radiation protection. To fulfil Euratom R&D programmes keys objectives of maintaining high levels of nuclear knowledge and building a more dynamic and competitive European industry, promotion of Pan-European mobility of researchers are implemented by co-financing transnational access to research infrastructures and joint research activities through to Research and Innovation and Coordination and Support Actions funding schemes. Establishment by the research community of European technology platforms are being capitalised. Mapping of research infrastructures and E&T capabilities is allowing a closer cooperation within the European Union and beyond, benefiting from multilateral international agreements and from closer cooperation between Euratom, OECD/NEA and IAEA and international fora. 'Euratom success stories' in facilitating Pan-European E&T collaborative efforts through Research and Training framework programs show the benefits of research efforts in key fields, of building an effective ‘critical mass’, of promoting the creation of ‘centres of excellence’ with an increased support for ‘open access to key research infrastructures’, exploitation of research results, management of knowledge, dissemination and sharing of learning outcomes.

Speaker: Mr Roger GARBIL (European Commission Euratom Nuclear Fission)

Material IAEA-CN-245-351_-_FR17_EURATOM_SUCCESS_STORIES_RG_V-2_3.01.2017.pdf

144 GEN IV Education and Training Initiative via Public Webinars

An increasing number of countries are opting for new nuclear energy as an important step towards economic development and environmental protection. According to the IAEA, electricity from nuclear energy may triple by 2050 as evidenced in the report IAEA-RDS-1/33; therefore, the projected use of this carbon-free technology will require many new nuclear engineers and scientists. In addition, countries such as France, the United States, which are the world’s largest producers of nuclear energy, are experiencing a decline in the nuclear energy workforce both in their national laboratories and in the private sector. The future vigor and prosperity of nuclear energy and associated nuclear science, clearly depend on continued use of available nuclear reactors as well as the development of advanced nuclear reactor technologies. To maintain the know-how in this field, to increase the knowledge of new advanced concepts, and to avoid the loss of the knowledge and competences that could seriously and adversely affect the future of nuclear energy, the Generation IV International Forum (GIF) established the GIF Education and Training Task Force. The task force serves as a platform to enhance open education and training as well as communication and networking in support to GIF. Indeed, its first initiative is the organization of a webinar series on the next generation of nuclear energy systems (Sodium Fast Reactor, Supercritical Water Reactor, Molten Salt Reactor, Lead Fast Reactor, Very High Temperature Reactor, and Gas-cooled Fast Reactor) and other cross-cutting subjects such as the basics on nuclear reactor systems, thorium fuel cycle, and nuclear fuel and materials. By exploiting modern internet technologies, the GIF Education and Training Task Force is reaching out to a broad audience and is raising the interest and strengthening the knowledge of participants in topics related to advanced reactor systems and advanced nuclear fuel cycles. This achievement is the direct result of partnering with internationally recognized subject matter experts and leading scientists in the nuclear energy arena who conduct live webinars on a monthly basis (for more details on the webinar series please see https://www.gen-4.org/). Besides opening the classroom to everyone in the world, the webinars offer earlier opportunities for interdisciplinary networking and educational and research collaboration. The details and examples of the GIF webinar modules will be presented in our paper.

Speaker: Dr Konstantin Mikityuk (Paul Scherrer Institute)

Material PAVIET-_Paper_ID#9_-_FR'17_Conference.docx

145 A proposal for a pan-European E&T programme supporting the development and deployment of ALFRED

The entire process of implementation of a nuclear program relies on the availability of qualified expertise and of national infrastructures providing the general framework for the smooth execution of regulated activities. Building an innovative reactor, besides the challenges related to the advanced nuclear technology and the important aspects of costs and financing, implies also the availability of: reactor design theoretical and experimental tools, communication methods and tools, adequate regulatory approaches, building techniques and, of course, connection to the past and current reactors operational experience. The EU ARCADIA project was conceived so as to promote the further development of nuclear research programs in Europe, including providing support for the ALFRED project towards its realization in Romania. Consequently, crucial focus was put both on the identification of a comprehensive list of primary needs for the ALFRED project, mainly for what concerns E&T, supporting Infrastructures and regulatory aspects, and on the investigation of the existing national and regional supporting structures – with a particular attention to the ones in Romania and in all the participating New Member States – for defining a map of competences potentially eligible to satisfy the previously identified needs. According to the output of this analysis and to the definition of Competence as “a holistic notion, consisting of cognitive, technical and behavioural aspects, each of them necessary for the complete definition of the job requirements”, which is found in the nuclear field Job Taxonomy formulated by the EHRO-N working group, an approach was proposed to fill the gaps in competences and infrastructures required for a country to develop and pursue a Gen-IV nuclear programme, which were identified based on the overall picture of the competences required for the implementation of ALFRED demonstrator as innovative reactor in Romania. In particular, two building blocks were identified and discussed as essential for developing and implementing such an E&T progamme, the latter consisting in applying an outcome-based pedagogical approach to lifelong learning, and harmonizing with the European Credit System for Vocational Education and Training (ECVET) principles.

Speaker: Ms Sara Bortot (KTH)

Material FR17_ARCADIA_FINAL_01.pdf

3:10 PM Coffee Break

1.5 LFR DESIGN & DEVELOPMENT

Conveners: Dr JAEWOON YOO (Korea Atomic Energy Research Institute), Mr Vadim Lemehov

146 Status of Generation-IV Lead Fast Reactor Activities

Since 2012 the Lead Fast Reactor provisional System Steering committee (LFR-pSSC) of Generation IV International Forum (GIF) has developed a number of top level activities with the aim to assist and support member countries developments of Lead Fast Reactor technology. The current full members (MoU signatories) of the GIF-LFR-pSSC are: EURATOM, JAPAN, the RUSSIAN FEDERATION and the REPUBLIC OF KOREA. The pSSC benefits also from the active participation of its observers: the UNITED STATES and the PEOPLE’S REPUBLIC OF CHINA. The paper highlights some of the main achievements of LFR-pSSC starting from the development of LFR System Research Plan, LFR white paper on safety and LFR Safety Assessment as well as Safety Design Criteria development. After the presentation of LFR-pSSC top level activities the status of the development of LFR in GIF countries is presented. The collaboration among partners of GIF-LFR-pSSC has proved its effectiveness to help the development of LFR through an open and friendly environment, developing important synergies and exchange of both technical and strategic information.

Speaker: Dr Alessandro Alemberti (Ansaldo Nucleare SpA)

Material Status_of_GIF_LFR_activities_065.docx

147 Simplification, the atout of LFR-AS-200

LFR-AS-200 is under development by Hydromine in cooperation with ENEA. LFR stands for Lead-cooled Fast Reactor, AS stands for Amphora-Shaped, referring to the shape of the inner vessel and 200 is the electrical power in MW. The project has been carried out by a team of engineers who had participated to the construction of SPX1. The strengths of the LFR-AS-200 are safety, simplicity, cost-competitiveness and operational simplicity. Safety relies on lead properties and is enhanced by innovative solutions including passively actuated and operated decay heat removal systems and a Steam Generator (GV) featuring a spiral-tube bundle, partially raised above the cold collector free level. The SG features a lower inlet window and an upper outlet window in correspondence of the lead free level, in order to drastically reduce the mass of displaced lead in case of SGTR. LFR-AS-200 dispenses with several components, hitherto typical of fast reactors, up to achieving a volume of the primary system per unit power of less than 1 m3/MWe, i.e. about 4 times lower than that of the SPX1 Sodium-cooled Fast Reactor (SFR), and also several times less than other international LFR projects, a key-factor for cost competition. The smaller size has been achieved through design simplification, that did mainly consist in the elimination, besides of the intermediate circuits (feature common to any other LFR project), of several components typical of SFRs and also of previous LFRs, namely (i) the in-vessel refueling machine, (ii) the above-core structure, (iii) the diagrid, (iv) the strongback, (v) the shielding elements, (vi) in-lead bearings of the pumps, (vii) the “LIPOSO” or equivalent tubular hydraulic connection between the pumps and the core and (viii) the “Deversoir” or equivalent system aimed at keeping the reactor vessel at the temperature of the cold collector. Several operational benefits pertaining to the proposed LFR-AS-200 technology are the result of insightful choices, typically the adoption of Fuel Assemblies (FAs) with a stem that extends above the lead free surface, and hung by a support system which is integral part of the FA’s head, i.e. located in the gas plenum and therefore visible by the operators. This keeps the support system free from neutron damage and thermal loads and strongly reduces the burden of in-service inspection of the primary system.

Speaker: Dr Luciano Cinotti (Hydromine Nuclear Energy S.a.r.l.)

Material IAEA-CN-245-140_paper.docx

148 Strategy and R&D status of China Lead-based Reactor

Lead-based reactor is one of the most promising nuclear energy systems for Generation-IV Accelerator Driven subcritical System (ADS) and reactors. Chinese Academy of Sciences (CAS) had launched a project to develop ADS and lead-based fast reactor technology since 2011. China LEAd-based Reactor (CLEAR) was selected as the reference reactor, which was performed by Institute of Nuclear Energy Safety Technology (INEST/FDS Team), CAS. The program consists of three stages with the goal of developing 10MWth lead-based research reactor (CLEAR-I), 100MWth lead-based engineering demonstration reactor (CLEAR-II) and 1000MWth lead-based commercial prototype reactor (CLEAR-III) on each stage. To promote the CLEAR project successfully, INEST places more emphases on reactor design, reactor safety assessment, design and analysis software development, lead alloy experiment loop, key technologies and components R&D activities. Detailed conceptual design of CLEAR-I has been completed and the engineering design is underway, which has subcritical and critical dual-mode operation capability for validation of lead cooled fast reactor (LFR) and ADS transmutation system and technology. KYLIN series Lead-Bismuth Eutectic (LBE) experimental loops have been constructed which is a large multi-functional lead-bismuth experiment loop platform. It has three independent loops, including material test loop, thermal hydraulic loop and safety loop. The objective of KYLIN is to perform structural material corrosion experiments, thermal-hydraulics tests and safety experiments. The key components including the control rod drive mechanism, refueling system, fuel assembly, and simulator for principle verification have been fabricated and tested. In order to integrated test the technologies of lead-based reactor, three integrated test facilities have been built, including the lead alloy cooled engineering validation reactor CLEAR-S, the lead-based zero power critical/subcritica reactor CLEAR-0, the lead-based virtual reactor CLEAR-V. In addition, series of innovative concepts for different purpose are being developed to enlarge the application perspective of lead-based reactors, which are not only for ADS and fast reactors but also for other innovative applications, such as CLEAR-SFB for spent fuel burning, CLEAR-Th for thorium utilization, CLEAR-H for hydrogen production, etc..

Speaker: Prof. Qunying Huang (CnINEST)

Material R&D_Progress_of_China_Lead-Based_Reactor(2017-04-07).docx

149 BREST-OD-300 REACTOR FACILITY. DEVELOPMENT STAGES AND JUSTIFICATION

BREST-OD-300, an innovative natural-safe fast reactor, is being developed as a pilot and demonstration prototype for the base commercial reactor facilities of future nuclear power with a closed nuclear cycle. Coolant in the reactor facility is lead, the layout of the primary circuit is integral, the reactor vessel material is multilayer metal concrete. The reactor core design uses mixed uranium-plutonium nitride as the fuel, and the fuel elements are contained in shroudless fuel assemblies (FA). Small reactivity margin, excluding prompt-neutron runaway is provided in the core. Decisions are based on a computational and experimental justification. To confirm the fuel serviceability, radiation tests of fuel elements are conducted in fast reactors. Full-scale fuel-free mockups of FA are tested. Tests have been conducted of the vessel elements. Experiments have confirmed the absence of a dependent break of steam generator tubes. Neutronic codes have been verified, including with the use of BFS critical assemblies. Loop facilities have been built on which studies are conducted to determine the radionuclide release from the coolant. It has been shown based on the calculation results that the probability of the core damage (without core melting) for nuclear power plants with the BREST-OD-300 reactor facility does not exceed 8.65•10-9 1/year.

Speaker: Mr Vadim Lemekhov (JSC NIKIET)

Paper ID-539_-_BREST-OD-300_reactor_facility_-_development_stages_and_justification_.doc

5.5 Large Component Technology I

Conveners: Mr Boris MARGOLIN, Dr Christian LATGE (CEA Cadarache 13108 Saint Paul lez Durance)

150 Experimental qualification of rotatable plug seals for Sodium Fast Reactor on a large scale test stand

In the framework of the ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration) project, the CEA Sealing Laboratory with its partner TECHNETICS (TGF) was involved to propose a new concept of rotating plug seals to replace the commonly used liquid-metal seals. An innovative combination of static, dynamic and inflatable seals in silicone rubber ensuring double tightness-barriers for the cover gas was developed. Following the design phase and materials studies, a dedicated test stand was built to qualify the technical performances of these seals. The large size of the test stand composed of a 2.5 m diameter rotating plates was chosen to provide a small profile height on seal diameter ratio, and a volume of enclosed gas large enough to allow representative qualification of tightness test methods. After a description of the test stand, the paper presents the mains outcomes of the technical qualifications (mechanical behavior, sealing performance, endurance test) led on several seals design.

Speaker: Mr karl vulliez (CEA, DEN, SDTC, Laboratoire d’Etanchéité, 30207, Bagnols Sur Cèze France.)

Paper Paper_ID24_FR17_KV_v1.docx

151 Heat Transfer Performance Test for a Sodium-to-Air Heat Exchanger with an Inclined Finned-Tube Banks

A separate effect test facility called SELFA (Sodium thermal-hydraulic Experiment Loop for Finned-tube sodium-to-Air heat exchanger) using liquid sodium and air as operating fluids has been developed. SELFA is one of the requisite sodium thermal-hydraulic test facilities within the framework of STELLA (Sodium Test Loop for Safety Simulation and Assessment) program, which is indispensible for the support of PGSFR (Prototype Gen IV Sodium-cooled Fast Reactor) development. The model heat exchanger (M-FHX) of SELFA was designed for performance demonstration of FHX (Forced-draft sodium-to-air Heat eXchanger) in PGSFR, which has three-row inclined finned-tube banks with staggered arrangement. Using this dedicated sodium heat exchanger test facility, several sets of heat transfer performance test have been conducted for validation of computational codes such as the heat exchanger thermal-sizing code (FHXSA) and the safety analysis code (MARS-LMR). In this study, we carried out performance tests for the M-FHX at the design point (i.e., thermal duty of 320 kWt). The test results obtained from this test have been used for its heat transfer performance evaluation through comparisons with the computational analyses results obtained from both a commersial CFD analyses as well as in-house thermal design and analysis computational codes developed by KAERI. Finally, it was confirmed that we have got reasonable experimenal datasets through this work.

Speaker: Dr Hyungmo Kim (KAERI)

Material FR17_IAEA-CN-245-183(Hyungmo_Kim)_Full_Paper_F.pdf

152 Development of the built-in primary sodium purification system for the

To purify primary sodium in the advanced BN-1200 reactor plant, a purification system with cold traps has been used that are located in the reactor vessel (in-built purification system). Such decision has excluded external communications of the auxiliary system with radioactive sodium and respectively a possibility that sodium will outflow to compartments outside the reactor. The sizes of cold traps located in the reactor are small that has limited the sodium flowrate through them, impurity storage capacity, and has made it necessary to replace traps in the course of reactor plant operation. Cold traps include such main components of the conventional external purification system as sodium communications, a portion of the cooling circuit, flow meter devices, and electromagnetic devices (a pump and throttle pump) to ensure sodium circulation and to control the sodium flowrate. In the course of development, options have been considered to cool traps with argon at the pressure of 1.5 MPa, liquid sodium, and gallium. To validate operation of electromagnetic devices for the cold trap, a package of research activities and R&D activities has been done: - Thermal irradiation studies have been done of sample electrotechnical materials intended for the electromagnetic pump and throttle pump. - Mockups of the electromagnetic pump and throttle pump have been manufactured and tested.

Speaker: Mr Sergey Rukhlin (JSC “Afrikantov OKBM”)

Paper IAEA-CN-245-404.pdf

Slides Rukhlin

153 ASTRID French SFR: Progress in Sodium Gas Heat Exchanger development

Within the framework of the French 600MWe Advanced Sodium Technological Reactor for Industrial Demonstration project (ASTRID), a Gas Power Conversion System (PCS) based on a Brayton cycle is studied. This innovative option has never been implemented in any Sodium Fast Reactor and is mainly justified by safety and acceptance considerations in inherently eliminating the sodium-water and sodium-water-air reaction risk existing in Steam Generators with a Rankine cycle. The present work describes the current status of the design of an innovative compact Sodium Gas Heat Exchanger (SGHE) and highlights the industrial challenges this technology raises. This paper presents the details of the design of the SGHE which allows a high thermal compactness. The main studies supporting the development are described whether on the external pressure vessel or on the compact internal heat exchanger modules; the thermal hydraulic program demonstrates the potential of the technology used whereas the thermo mechanical analyses show the good behavior of this exchanger under the ASTRID operating conditions. The manufacturing welding process optimization for the heat exchanger modules is ongoing in order to produce a component with nuclear specifications. Specific sensors and control techniques are also being developed in order to assess the manufacturing process quality and to allow future in-service inspections. At last, the qualification program is presented and the results obtained on an operating small scale SGHE mock up (DIADEMO) working under ASTRID conditions are described.

Speaker: Mr david plancq (CEA)

Material Paper_IAEA-CN-245-286.docx

6.5 Uncertainty Analysis and Tools

Conveners: Mr Arkadii Kiselev, Prof. Gerald Rimpault (CEA)

154 Recent and Potential Advances of the HGPT methodology

**Recent and potential advances of the Heuristically based Generalized Perturbation Theory (HGPT) methodology are discussed:** - *The subcriticality monitoring method* Basing on the HGPT methodology applied to subcritical systems, a procedure is described for the online monitoring of the subcriticality level of ADS reactors with minimal interaction with the plant normal operation. The proposed method consists in compensating slow, small movements of a control rod with likewise slow, small alterations of the external source strength, so that the overall power is maintained constant. The estimation of the subcriticality level requires the knowledge of a bias factor. This implies the standard precalibration of a control rod and the precalculation of the importance function associated with the reactor power control (in this case, the external neutron source strength). - *The hot spot identification by sensitivity analysis and probabilistic inference* A method is described by which the information obtained on-line through a system of neutron measuring devices such as self-powered neutron detectors (SPNDs, called also collectrons) inserted in the core of a nuclear power reactor allows the on-line detection of a possible hot spot during plant operation. The method is based on the HGPT techniques, for the calculation of the sensitivity coefficients of the integral quantities measured with the collectrons with respect to parameters representative of the hot spot, and on the use of statistical inference techniques, taking into account the errors associated with the measurements. The methodology allows to assess the effect on the quality of the hot point detection system following possible failures of the measuring devices during the core life cycle. Such an assessment may be useful for defining an adequate protection strategy in terms of quality, number and distribution of the collectrons. This method has been initially aimed to be adopted in thermal power systems, but with advanced detection techniques underway it might be adopted also in fast ones. - *Use of the GPT methodology for the analysis of reactivity worths with Monte Carlo* Perturbation methods are part of the reactor physics foundation for the study of fundamental quantities considered in design and safety analysis of nuclear reactors. In deterministic codes standard perturbation theory (SPT) and generalized perturbation theory (GPT) methods have been historically developed and used. Monte Carlo codes, such as MCNP 6.1, can also perform, via adjoint weighted tally, SPT calculations of reactivity worths. A method is proposed to enable Monte Carlo codes to implement GPT.

Speaker: Prof. Augusto Gandini (Sapienza University of Rome)

Paper IAEA-CN-245-567rev3.docx

155 Evaluation of βeff measurements from BERENICE programme with TRIPOLI4® and uncertainties quantification

The use of the Iterated Fission Probability method in the Monte Carlo code Tripoli4® gives credit to deterministic codes such as ERANOS for calculating βeff. The asset of Tripoli4® is the possibility to get a better representation of experimental cores, especially the R2 experimental core which exhibit more experimental canals for hosting large fission chambers. The BERENICE measurements campaign took place in the experimental facility MASURCA at CEA Cadarache with the two cores R2 reference and R2 experimental using enriched uranium fuel and one core ZONA2 using MOX fuel. For JEFF3.2, the revised C/E ratios are of 1.2% ± 2.0% for the ZONA2 core and -1.2% ± 2.9 % for the R2 experimental core when using the Noise measurement technique. The nuclear data uncertainty propagation has been leading to a 2.6% uncertainty for U-Pu core and 2.8% for enriched uranium cores with main contributors being the delayed neutron fission yield and the fission cross section of U238.

Speaker: Prof. Gerald Rimpault (CEA)

Material BERENICE_FR17_218.pdf

156 System of Codes and Nuclear Data for Neutronics Calculations of Fast Reactors and Uncertainty Estimation

Designing of neutronics characteristics of fast reactor cores and fuel cycle requires to use certified and qualified sets of computer codes and nuclear data. The calculation codes should be related to the modern state of computational techniques. The used nuclear constants should be adequate to the most reliable evaluations, adopted in modern libraries of evaluated nuclear data. The paper consider a modern state of Russian neutronics computer codes and nuclear data used in fast reactor applications for calculation of core and nuclear cycle parameters. The ROSFOND evaluated nuclear data files and the ABBN group data set are used as the basis of nuclear input data. The ROSFOND library now contains about 650 files of data for most important and not so important reactor materials. The selection of files was made based on BROND-3, ENDF/B-VI.8 and VII.0, JEF-2.2 and JEFF-3.1, JENDL-3.3 evaluations by comprehensive study of their quality. For treating the ABBN data the special code system CONSYST/ABBN was developed. Three directions in developing of codes for fast reactor neutronics calculations can be stated: (1) discrete codes, (2) based on Monte-Carlo, (3) used synthesis methods. Codes, which are used in the design calculations, mostly solve the Boltzman transport equation in diffusion approximation, they are: TRIGEX, JARFR, GEFEST, FACT-BR, SYNTES. Codes, which are based on Monte-Carlo method, were developed during many years. Nowadays they have additional impulse in interest due to fast developing of the computational technique. Recently a code MMKK was developed. It now used in planning and analyzing of reactor-physics experiments as well as for precise calculations of fast reactors BN. For the shielding calculations as well as for determining diffusion-transport corrections codes TWODANT and DORT-TORT are used. For the depletion and kinetic calculations CARE and ORIGEN codes are used. The main feature of the all mentioned codes is that they use one, same and unique constants data base ABBN with the code CONSYST for generation of effective cross-sections. The system INDECS of codes and archives is now used for uncertainty estimations which is based on usage of perturbation theory and covariance matrices of nuclear constants. The TRIUM code based on GRS method is now developed. It is a synthesis of TRIGEX, MMKK and INDECS codes.

Speaker: Dr Gennady Manturov (PhD)

Paper SystemOfCodes_475_1.docx SystemOfCodes_475_1.pdf

Slides SystemOfCodes_475_1.pptx

157 Sensitivity and Uncertainty Analysis in Best-Estimate modeling for PGSFR Under ULOF Transient

In this research, uncertainty analyses for multiple safety parameters were performed for Unprotected Loss of Flow (ULOF) for the Prototype Gen-IV Sodium-cooled Fast Reactor (PGSFR) by using the PArallel Computing Platform IntegRated for Uncertainty and Sensitivity analysis (PAPIRUS). The objective of the global uncertainty analysis is to evaluate all safety parameters of the system in the combined phase space formed by the parameters and dependent variables. The uncertainty propagation was performed by mapping the uncertainty bands of the model parameters through the MARS-LMR to determine the distributions for the fuel centerline, cladding, and coolant temperatures. The Best Estimate Plus Uncertainty (BEPU) analysis adopted for uncertainty quantification of the code predictions has been performed through a statistical approach where the Figure of Merit (FOM) is evaluated multiple times by using several combinations of parameters that are randomly generated according to their distributions. The statistical approach of uncertainty quantification is known to be very powerful for estimating response distributions, but sometimes inapplicable owing to demanding calculation requirements. In this research, Wilks’ formula was used to estimate the 95% probability value of the FOM from a limited number of code calculations. This paper also introduces the application of data assimilation in best-estimate modeling to improve the prediction of the reactor system performance by refining various sources of uncertainties through model calibration technique. An inverse problem was formulated based upon Bayes theorem and solved to estimate the posteriori distributions of parameters.

Speaker: Jaeseok Heo (Korea Atomic Energy Research Institute)

Paper FR17_255_Jaeseok_Heo_rev1.pdf

158 Objectives and Status of the OECD/NEA sub-group on Uncertainty Analysis in Modelling (UAM) for Design, Operation and Safety Analysis of SFRs (SFR-UAM)

An OECD/NEA sub-group on Uncertainty Analysis in Modelling (UAM) for Design, Operation and Safety Analysis of Sodium-cooled Fast Reactors (SFR-UAM) has been formed under the NSC/WPRS/EGUAM and is currently undertaking preliminary studies after having specified a series of benchmarks. The incentive for launching the SFR-UAM task force comes from the desire to utilize current understanding of important phenomena to define and quantify the main core characteristics affecting safety and performance of SFRs. Best-estimate codes and data together with an evaluation of the uncertainties are required for that purpose, which challenges existing calculation methods. The group benefits from the results of a previous Sodium Fast Reactor core Feed-back and Transient response (SFR-FT) Task Force work under the NSC/WPRS/EGRPANS. Two SFR cores have been selected for the SFR-UAM benchmark, a 3600MWth oxide core and a 1000MWth metallic core. Their neutronic feedback coefficients are being calculated for transient analyses. The SFR-UAM sub-group is currently defining the grace period or the margin to melting available in the different accident scenarios and this within uncertainty margins. Recently, the work of the sub-group has been updated to incorporate new exercises, namely, the depletion benchmark, the control rod withdrawal benchmark, and the SUPER-PHENIX start up transient. Experimental evidence in support of the studies is also being developed.

Speaker: Prof. Gerald Rimpault (CEA)

Paper UAM-SFR_FR17_220rev.pdf

5:10 PM Coffee Break

Poster Session 1

159 "Peculiarities of behavior of Coated Particle fuel in the core of Fast Gas Reactor BGR-1000"

Fast Gas Reactor BGR-1000 with thermal power of 2 GW is cooled with high-pressure helium (16 MPa), heated in the core from 350 to 750oC. In a steam generator of the power conversion system the thermal power is transferred to the SCW-coolant of secondary circuit, which goes to the turbine with pressure of 30 MPa and temperature of 650oC. Reactor core contains Fuel Assemblies (FA) having perforated shrouds. FA's inside cavity among shroud, control rod guide tubes and central perforated collector is filled with pebble-bed of micro-fuel coated particles (CP). Helium coolant goes into FA through the perforated shroud, passes over CPs removing heat from them and goes then to the FA outlet collector through its perforated wall. The mix-carbide fuel UPuC with mean plutonium content of 16.5% is dispersed in the core in the form of CPs kernels. While loading of heavy atoms is 3640 kg, reactor average burnup amounts 9.7% h.a. Having a breeding ratio of 1.025 reactor can operates in the regime of self-provision of the secondary fuel in the closed fuel cycle. Calculational optimization of CP design has given the following performance of the CP kernel and coatings: CP outer diameter of 2000 um, kernel diameter of 1640 um, nondense pyrocarbon buffer coating of 125 um, dense pyrocarbon inner layer (IPyC) of 10 um and outer protective SiC layer of 50 um. In the paper the basic positions of the model of the thermo-mechanics of BGR-1000 coated particles are presented and calculational results revealing the effect of CP design on their behavior during irradiation are demonstrated. It is shown, that in the result of the viscous deforming the summarized volume of the kernel and buffer, limited by the elastic SiC, keeps practically invariable. In an equilibrium state volume changes of the fuel (due to its swelling) and of the pyrocarbon layers (due to radiation-induced size changes) are compensated by changing of the volume fraction of porosity in the fuel and buffer owing to their viscous deformations.

Speaker: Mr Aleksei Sedov (NRC "Kurchatov Institute")

Material alexsedov

160 (U,Pu)O2-x MOX pellet for Astrid reactor project

Abstract -------- Since 2015, **AREVA and CEA** teams decided to launch yearly industrial tests of MOX pellets with an adapted GEN III design in the **MELOX** plant, to prepare the future manufacturing of MOX fuels bundle for Astrid reactor. First campaign (2015) of tests was dedicated to demonstrate the feasibility of this manufacturing at half industrial scale; Main modifications involved the pelletizing station of LCT workshop (small scale line for MOX manufacturing) and one of the industrial furnaces, in order to define the range of main parameters (powder preparation, pelletizing and sintering steps and MOX pellet analyses procedures). Specified analyses results were performed in MELOX plant laboratory, completed with EPMA analyses on MOX pellet sent to CADARACHE CEA laboratory : First results show that required properties of these MOX pellet, meet the specified criteria defined by CEA teams, the most important one’s are related to pellet design (dimensions and density), Pu distribution and stoichiometry. Second campaign (2016) of tests, included a powder preparation step at industrial scale on one of the blender of the MELOX plant, in order to prepare the industrial manufacturing of MOX pellet for one fuel bundle, designed for a prototypical irradiation. Main results show again that the specified criteria are respected increasing the confidence in the process route. Keywords: MOX, annular pellet, Astrid, EPMA.

Speaker: Dr Thierry GERVAIS (AREVA)

Material IAEA_FR17_CN582_v01.pdf

161 3-D Core Design of the TRU-Incinerating Thorium RBWR Using Accident Tolerant Cladding

This project investigates the safety of the optimal core design for the RBWR-TR – a reduced moderation BWR with a high transuranic (TRU) consumption rate. This design is a variant of the Hitachi RBWR-TB2, which arranges its fuel in a hexagonal lattice, axially segregated seed and blanket regions, and fits within a standard ABWR pressure vessel. The RBWR-TR eliminates the internal axial blanket and absorbers from the upper reflector, and uses thorium instead of depleted uranium for the axial blankets. Its coolant flow rate is higher than of the TB2 design. Both designs are initially presented with Zircaloy-2 cladding. The softer neutron spectrum of the RBWR-TR core, along with its lower peak linear heat generation rate, results in a lower cladding fast neutron fluence than of the RBWR-TB2 core. However, the peak fluence of fast neutrons (E >= 0.1 MeV) the cladding is exposed to exceeds the limits for Zircaloy-2, making this material not acceptable even for the RBWR-TR design. At such high fast fluence levels, zirconium-based cladding experiences faster rates of corrosion and hydrogen pickup, which embrittles the cladding and eliminates any margin for accident scenarios. Alternative cladding materials to Zr-based alloys are being investigated for accident tolerant fuels such as stainless steel based materials that are not limited by hydrogen pickup phenomena. However, the steel cladding penalizes the neutron economy and limits the discharge burnup. The design variables of the parametric studies include the cladding material type, cladding thickness, gap between fuel and cladding, fuel smear density and fuel-to-moderator volume ratio. The changes of the void feedback, cycle length, burnup, shutdown margin, and critical power ratio to variation in each of the design variables are calculated to determine the optimal design. A design that meets all the design constraints will be presented.

Speaker: Ms Sandra Bogetic (University of California at Berkeley)

Material IAEA-CN-245-_371.pdf

162 A Conceptual design of engineering-scale plant applied the simplified MA-bearing fuel fabrication process

Researchers at Japan Atomic Energy Agency (JAEA) have proposed the transmutation of minor actinides (MAs) by both fast reactors (FRs) and accelerator driven system (ADSs) as a way to contribute significantly to the reduction of the volume and the potential radiotoxicity of radioactive wastes. In order to achieve this goal, it is important to introduce a fully automated and remote operation fuel fabrication plant with shielded hot cells and manipulators to deal with extremely strong radiation dose and heat generation from MAs. JAEA’s facilities including Plutonium Fuel Production Facility (PFPF) have fabricated MOX fuel. On the basis of the operational and technical experience obtained in above facilities, the conceptual design of engineering-scale plant applied the simplified MA-bearing fuel fabrication process with shielded hot cells and manipulator was done. It will be able to fabricate high MA-bearing fuel and to perform the maintenance and repairing of each equipment with manipulators. This plant will be constructed based on this concept and development plan.

Speaker: Mr YOSHIKAZU YAMADA (Japan Atomic Energy Agency)

Paper IAEA-CN-245-241.docx

163 A Demand Driven Way of Thinking Nuclear Development – Neutron Physical Feasibility of a Reactor Directly Operating SNF from LWR

Invention and innovation with regards to nuclear reactor development can be described with the concept of developments in s-curves. This view is taken to identify the problems of nuclear development. Following this, the request are formed for the definition of demand driven objectives. The future objectives should follow the UN sustainable development goals. The key words to form the vision of electric energy production are: very limited request for resources and production of waste, affordable economics and safe, secure and reliable operation which can be assembled to the dream solution – the perpetuum mobile. It is concluded that a reactor operating in closed fuel cycle using spent nuclear fuel from Light Water reactors would come close to this vision. Additionally, the technology fully fulfils the UN sustainability request of using technologies which provide future generation with solutions to increase the amount of available resources. Following this demand driven theoretical discussion of objectives, a new innovative proposal is presented. A proposed reactor which is operated directly on SNF from LWRs as main fuel resource. The simulation tools and the limitation of the simulation are discussed. A proof of feasibility is given from neutron physical point of view. The major challenge is to establish a breeding process which provides enough new fissile material from the inserted SNF. For the start-up of the system a support of fissile material in the initial core and the transition phase is required. The feasibility of sufficient breeding is demonstrated, a first estimation of the resources in a possible fuel cycle is given, and the consequences on the back end of the fuel cycle are discussed. Finally, the challenges of the proposed technology are highlighted to stimulated future R&D to make a sustainable innovative nuclear reactor possible. This could form attractive major innovation challenge for a wide variety of engineers to form the basis for the long term success of nuclear reactors as a major carbon free, sustainable, and applied highly reliable energy source.

Speaker: Prof. Bruno Merk (University of Liverpool)

Paper Merk_FR17_full_paper_I.docx

164 A High Density Uranium Zirconium Carbonitride LEU Fuel for Application in Fast Reactors

For many years, Russian researchers have developed and tested a high density, high temperature U-Zr-C-N fuel for potential application in different types of reactors, including fast reactors. As part of this effort, reactor tests have been performed to low burnup. However, reactor-testing data is still needed at high burnup to confirm the optimal performance of the fuel. The SM-3 reactor, which is a high-flux reactor located in Dmitrovgrad, Russia, will be used to test a U-Zr-C-N (U0.9Zr0.1C0.5N0.5) fuel to ~40% burnup. The fuel will then be examined to determine its performance during irradiation. The fuel that will be tested has a density of 11.9 g/cm3 and an enrichment of 19.75% (uranium-235), and the uranium density of this fuel material is 10.8 g/cm3. About 1000 effective days of irradiation will be required to achieve the targeted burnup. This presentation will discuss the details of the planned irradiation, along with results of out-of-pile research that has been performed on the as-fabricated fuel. The positive characteristics of the U-Zr-C-N fuel will be discussed, and comparisons will be made to other fuel types.

Speaker: Mr S Sikorin (The Joint Institute for Power and Nuclear Research SOSNY of the National Academy of Sciences of Belarus)

Material Keiser_Fast_Reactor_Conference_2017_Paper_v3.docx

165 A Mechanistic Source Term Calculation for a Metal Fuel Sodium Fast Reactor

A mechanistic source term (MST) calculation attempts to realistically assess the transport and release of radionuclides from a reactor system to the environment during a specific accident sequence. The U.S. Nuclear Regulatory Commission (NRC) has repeated stated its expectation that advanced reactor vendors will utilize an MST during the U.S. reactor licensing process. As part of a project to examine possible impediments to sodium fast reactor (SFR) licensing in the U.S., an analysis was conducted regarding the current capabilities to perform an MST for a metal fuel SFR. The purpose of the project was to identify and prioritize any gaps in current computational tools, and the associated database, for the accurate assessment of an MST. The results of the study demonstrate that an SFR MST is possible with current tools and data, but several gaps exist that may lead to possibly unacceptable levels of uncertainty.

Speaker: Dr David Grabaskas (Argonne National Laboratory)

Material 25_Rev.pdf

166 A Preliminary Study of P&T Scenario on a Sustainable Energy System in China

1.radioactive harmfulness of the spent fuel 2.capability of MA burning in 600 MWe FBR 3.a study of nuclear energy system scenario 4.summary

Speaker: Mr Yong YANG (China Institute of Atomic Energy)

Material A_Preliminary_Study_of_P&T_Scenario_on_a_Sustainable_Energy_System_in_China.docx

167 Advanced Energy Conversion for Sodium-Cooled Fast Reactors

Advanced energy conversion using the supercritical carbon dioxide (sCO2) Brayton cycle has been under development at Argonne National Laboratory (ANL) for over twelve years. It has been shown to enable the SFR capital cost per unit output electrical power ($/kWe) or Levelized Cost of Electricity (LCOE) to be significantly reduced improving SFR economics (a U.S. DOE SFR goal) and eliminating sodium-water reactions, although there still remains a need to understand potential sodium-CO2 interactions that is being addressed through ongoing sodium-CO2 interaction tests. It has been shown that the cycle enables the use of dry air cooling whereby heat is rejected directly to the air atmosphere through the use of finned tube air coolers. A Plant Dynamics Code for system level dynamic analysis of sCO2 cycles has been developed, coupled to the SAS4A/SASSYS-1 SFR transient analysis code, and is being validated through comparison with data from sCO2 integrated cycle test loops.

Speaker: Dr James Sienicki (Argonne National Laboratory)

Material IAEA-CN-245-22_Draft_Full_Paper.docx

168 Advanced Reactor PSA Methodologies for System Reliability Analysis and Source Term Assessment

Beginning in 2015, a project was initiated to update and modernize the probabilistic safety assessment (PSA) of the GE-Hitachi PRISM sodium fast reactor. This project is a collaboration between GE-Hitachi and Argonne National Laboratory (Argonne), and funded in part by the U.S. Department of Energy. Specifically, the role of Argonne is to assess the reliability of passive safety systems, complete a mechanistic source term calculation, and provide component reliability estimates. The assessment of passive system reliability focused on the performance of the Reactor Vessel Auxiliary Cooling System (RVACS) and the inherent reactivity feedbacks mechanisms of the metal fuel core. The mechanistic source term assessment attempted to provide a sequence-specific source term evaluation to quantify offsite consequences. Lastly, the reliability assessment focused on components specific to the sodium fast reactor, including electromagnetic pumps, intermediate heat exchangers, the steam generator, and sodium valves and piping.

Speaker: Dr David Grabaskas (Argonne National Laboratory)

Material 26_Rev.pdf

169 Americium Retention During Metallic Fuel Fabrication

Under the US Fuel Cycle Technologies program Advanced Fuels Campaign metallic fuel has been chosen as a leading candidate for fast reactor transmutation fuels. Significant losses were seen in an earlier attempt to incorporate americium into a metallic fuel, although the majority of the losses were likely not caused by the volatility of americium, these concerns have persisted. A furnace has been installed in a transuranic qualified glovebox in order to verify that americium losses can be controlled during the fuel casting process through atmospheric pressure. A charge of 81U-7.5Pu-1.5Am-10Zr was melted under an argon atmosphere a total of three times. Each time the charge was held at 1450°C for approximately 10 minutes under flowing argon. After each melting cycle, the resulting fuel ingot was sampled for chemical analysis to verify americium content. Resulting analysis showed americium content remained stable throughout the heating cycles.

Speaker: Mr Randall Fielding (Idaho National Laboratory)

Material Am_Retention-Full_Paper_Fielding_FINAL.docx

170 An Assessment of Fission Product Scrubbing in Sodium Pools Following a Core Damage Event in a Sodium Cooled Fast Reactor

The U.S. Nuclear Regulatory Commission has stated that mechanistic source term (MST) calculations are expected to be required as part of the advanced reactor licensing process. A recent study by Argonne National Laboratory has concluded that fission product scrubbing in sodium pools is an important aspect of an MST calculation for a sodium cooled fast reactor (SFR). To model the phenomena associated with sodium pool scrubbing, a computational tool has been developed. This tool was developed by applying classical theories of aerosol scrubbing, developed for the case of isolated bubbles rising through water, to the decontamination of gases produced as a result of a postulated core damage event in an SFR. The model currently considers aerosol capture by Brownian diffusion, inertial deposition, and gravitational sedimentation. The effects of sodium vapor condensation on aerosol scrubbing are also treated. This paper provides details of a parametric study performed to determine key modeling uncertainties and sensitivities.

Speaker: Dr Matthew Bucknor (Argonne National Laboratory)

Material 55-Final.pdf

171 APPLICATION OF PHYSICAL MODELING WHEN CALIBRATING HIGH RANGE ELECTROMAGNETIC FLOWMETERS

At high flowrates the «drift» of magnetic field characterized by the criterion Rem = μ0σvD can be revealed in the readings of the electromagnetic flowmeters if it’s magnetic field is insufficiently extensive. The modeling of high range flowmeters presented involves using a small on scale transducer sample as compared to full scale flowmeter maintaining similarity of measuring section and magnetic field distribution. The hydrodynamic and MHD criteria (Re,Rem), corresponding to full scale flow conditions can be provided at much lower flowrate values. The experiments were carried out at sodium calibration test facility IRS-M, using its main loop and two parallel auxiliary loops supplied by calibrated electromagnetic flowmeters. On the model of a measuring section with the pipe DN150 at a total flowrate G=360 m3/h the value Rem=7,5 has been achieved, that corresponds to parameter of the flowmeter installed in accident heat removal system of BN-800 reactor (Gmax = 720 m3/h, DN300). Calibration characteristics have been determined for different electrode pairs the longitudinal extension of magnetic field being Lm=0,7DN. An estimate of the nonlinearity introduced by the quadratic term in the dimensionless representation is obtained, E=k0(1 - αRem)Rem. The coefficient α can be used further to adjust the characteristic of the full scale flowmeter.

Speaker: Mr Vadim Alexandrovich Shurupov (JSC “SSC RF – IPPE”)

Paper FR17-166.pdf

172 Application of Heterogeneous Fuel Assemblies in the Core of Modular Fast Sodium Reactor

The work contains results of calculation studies of neutron physics characteristics of the fast modular sodium reactor core, in which fuel assemblies without casing with heterogeneity inside fuel assemblies are used. Metal fuel (U-Pu-Zr) is the most advantageous fuel of all known challenging fuel types for a fast sodium reactor regarding neutron characteristics. It enables obtaining the maximal mass of heavy nuclei in the core and a harder neutron spectrum due to absence of light nuclei in comparison with other fuel types. However, experience of metal fuel application is extremely little, and this fuel has not been in commercial operation yet. Application of a heterogeneous fuel assembly consisting of fuel elements with highly enriched (<30%) mixed oxide fuel combined with fuel elements of metallic uranium (or alloy) enables increasing concentration of fissionable and fertile nuclides in comparison with homogeneous fuel assemblies with MOX fuel and obtain similar indices to ones of homogeneous fuel assemblies with metal U-Pu-Zr fuel. A heterogeneous fuel assembly consisting of fuel elements with MOX fuel and fuel elements with metallic uranium of natural composition or U-Zr alloy and a homogeneous fuel assemblies were compared in the course of research. Use of U-Zr alloy without plutonium at the beginning of the campaign and its relatively low average burnup reduces requirements to metal fuel and enables using it already in the nearest future. A heterogeneous fuel assemblies can become an intermediate variant during conversion to the metal fuel core or a final variant if it has better indices than fuel assemblies with metal U-Pu-Zr fuel.

Speaker: Mr Yaroslav Kotov (National Recearch Centre "Kurchatov Institute")

Material FR17-KOTOV-IAEA-CN-245-302.pdf

173 Assessment of accuracy from the use of point kinetics when analyzing transition processes in high power fast reactor

“Point kinetics” approximation is widely used in reactor justification for calculation of transient and emergency modes in the first place. The point kinetic model is used as the base model for Russian DINROS, GRIF, SOKRAT-BN software programs used for safety justification of fast reactors. Its popularity is explained by its relative simplicity and physical transparency (possibility to interpret results on the language of reactivity effects and easily demonstrative verification). Computational study of errors caused by the use of point kinetic model is performed with the use of UNICO multi-physical software (3D neutronics in diffusional approximation + 3D thermohydraulics) for three non-static test example problems for BN-1200 reactor: • The problem of sudden change of coolant temperature at the inlet of pressure header of reactor core (in one of 4 first circuit loops). • Emergency protection rods drop at nominal power (example of fast running process). • Self-act of one of the control rods. It is shown that fuel rod temperature estimation error during self-act of one of the control rods can reach 100ºС.

Speaker: Dr Igor Suslov (ITC "PRORYV" Project)

Material 204_Suslov-Shvetsov_paper_revised.doc

174 Assessment of the reactivity effects of Gas cooled Fast Reactor

Presented paper assess standard reactivity effects, as coolant void effect and Doppler effect, of the power scale Gas cooled Fast Reactor (GFR 2400) in a comprehensive manner by application of the perturbation theory. To achieve high validity of the results the conventional SCALE 6 system and adapted computational scheme (ACS) are utilized. The ACS is based on standard computational package incorporating codes like TRANSX, PARTISN, DIF3D, PORK and STUUP with cross section data library optimized for fast reactor applications. The reactivity effects of the GFR 2400 core were calculated in a range of the pre-defined temperatures and coolant pressures. Coolant void effect for nominal and lowest operational pressure and Doppler effects for highest temperature increase and decrease were identified as most important reactivity effects. Spatial distribution and reactivity components decomposition of the selected reactivity effects is analyzed and presented in this paper for further evaluation. In the next step, sensitivity and uncertainty analysis is performed for these reactivity effects where the sensitivity coefficients are validated via direct perturbation calculation and energy profile comparison. In case of possible optimization of selected reactivity effects the most sensitive isotopes and contributors to the overall uncertainty are identified. The final part of the paper is dedicated to the first optimization studies and preliminary results are presented. Two possible options of optimizations are proposed; homogeneous and heterogeneous. In the heterogeneous case the rod follower volume is used for application of materials which can possibly influence the reactivity effects. In the second case the core design modifications are homogeneously distributed over the entire core volume. Finally in the conclusion recommendations and some drawbacks are collected for further analyses.

Speaker: Mr Jakub Lüley (Slovak University of Technology)

Material Contribution_Luley_v5.docx

175 ASTRID hot cells

Authors: René-Paul BENARD, Bernard GUILLOU (SEIV) Christoph DÖDERLEIN (CEA) The ASTRID reactor is the French demonstrator for Generation IV sodium cooled fast reactors and needs as such to respond to challenges in the qualification of innovative components and materials. Considering its role as R&D platform for the fast reactor line to come, ASTRID will be endowed with a set of hot cells. The French company SEIV, subsidiary of the ALCEN group, has been in charge since 2013 of the full preliminary design of this facility. The main purpose of the ASTRID hot cells is to perform non-destructive examinations (NDE, such as visual inspection, 3D X-ray tomography, dimensional inspection, eddy current testing) on the spent core sub-assemblies and fuel pins. To extract the latter from the sub-assemblies, a dismantling unit is foreseen in the facility. The proposed paper gives a description of the components and capabilities of the ASTRID hot cell facility. The facility consists of a main cell, where the NDE equipment are installed, the lower cells with the dismantling machine and 3D X-ray scanner device and finally the upper cell which serves as an airlock for handling functions. The ASTRID hot cells will feature remote operations of NDE equipment, eg. with new generation manipulator with electrical master arm using haptic technology. This design aims to minimize of the use of expensive lead windows, increase handling capabilities and improve operator ergonomics.

Speaker: Dr Bernard GUILLOU (SEIV)

Material 20170113Article-conference_juin_russie-seiv-01_format_iaea_1.pdf

176 ASTRID reactor: design overview and main innovative options for Basic Design

Abstract FR 17 Track 1. Innovative Fast Reactor Designs ASTRID reactor: design overview and main innovative options for Basic Design F. Chanteclair(a) in association with all the ASTRID industrial partners (a) French Alternative Energies and Atomic Energy Commission (CEA), F- 13 108 Saint Paul Lez Durance, France Sodium-cooled Fast Reactors (SFR) is one of the Generation IV reactor concepts selected to secure the nuclear fuel resources and to manage radioactive wastes. In the frame of the June 2006 French act on sustainable management of radioactive materials and wastes, French Government entrusted CEA (French Commission for Atomic Energy and Alternative Energy) to conduct design studies of ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration) prototype in collaboration with industrial partners. The ambitious objectives of ASTRID reactor are to fulfil the GEN IV requirements. It has led to the implementation of innovative technological solutions which go beyond the current feedbacks. Necessarily, these innovations will have to be consolidated within the framework of Research & Development actions and qualification programs. In its Basic Design stage, ASTRID has built a coherent conceptual design configuration with innovative techniques and systems across all domains: core, fuel assembly technology, nuclear island, civil engineering, energy conversion system, plant layout, ISI&R, fabricability, … and even in the project management. The object of this document is to provide an overview of the significant innovations under consideration on ASTRID. It will also allow to present the partners contribution to this seek for innovations for better performances and/or enhanced safety.

Speaker: Mr FREDERIC CHANTECLAIR (French Alternative Energies and Atomic Energy Commission (CEA))

Paper CN245-400_V4.pdf

177 Benchmark Between EDF And IPPE On The Behavior Of Low Sodium Void Reactivity Effect Sodium Fast Reactor During An Unprotected Loss Of Flow Accident

The validation of severe accident analysis codes for Sodium Fast Reactors (SFR) is a difficult task as it is not possible to carry out full scale integral experiments. Therefore, in addition to the validation of specific models with dedicated experiments, it is of the utmost importance to increase the confidence we have in these codes by performing benchmarking exercises with independent codes and by independent teams. As EDF R&D and IPPE are both interested in the analysis of the behavior of low Sodium Void Reactivity Effect (SVRE) cores during severe accidents, whether to support R&D on the ASTRID project (conducted by CEA) or to support R&D on the BN family reactors, a benchmarking exercise has been launched in this purpose. As a first step, a low SVRE core design has been developed especially for this benchmark. Its main neutronics properties related to severe accident behavior - sodium density and void effect and fuel Doppler effect - have been evaluated with the CEA code ERANOS for EDF and with TRIGEX for IPPE and are compared in this article. Finally, the primary phase of an Unprotected Loss Of Flow (ULOF) accident has been simulated by each partner. On EDF side, the SIMMER code has been used whereas IPPE performed its calculations with its code COREMELT. Main results concerning power evolution and sodium boiling are compared.

Speaker: Mr David Lemasson (EDF)

Paper Final version of the paper, including review corrections.

178 BISON for Metallic Fuels Modeling

The fuel performance code BISON has recently been extended to simulate U-Zr and U-Pu-Zr metallic fuel rods irradiated in the US sodium cooled fast reactor EBR-II. By introducing fuel and clad specific material models, the backbone provided by the MOOSE/BISON software architecture has allowed rapid development of metallic fuel capabilities. Zirconium based fuels present unique challenges due to the different phases that exist at irradiation temperatures. Each phase possess differing thermo-mechanical properties, necessitating explicit tracking of the relative concentration of phases throughout the fuel rod in order to capture the integral behavior. In addition, the transition temperatures between phases change over the course of irradiation due to zirconium diffusion in the fuel, necessitating coupling of the thermo-mechanical simulation to the Fickian and Soret diffusion of Zr. Along with a robust U-Zr and U-Pu-Zr zirconium redistribution model, newly formulated thermodynamic properties such as thermal conductivity, and phase-dependent mechanical properties such as swelling, allow BISON to capture the behavior of zirconium based metallic fuel at a variety of operating temperatures and irradiation histories. These models have been integrated into BISON, verified using standard practices, and validated against full 2D-RZ simulations of fuel irradiated in EBR-II. The incorporation of phase-dependent models allows BISON to be extended to test other novel fuel designs, some of which show promising characteristics depending on fabrication feasibility.

Speaker: Christopher Matthews (Los Alamos National Laboratory)

Material FR17_BISON_rev2.docx

179 CFD Simulation of Corium / Materials Interaction for Severe Accidents

In case of a severe accident inside a sodium-cooled fast reactor, corium interacts with many materials along its way towards the core catcher. Once deposited, corium can also interact with protective and / or sacrificial material. Among those materials, refractory ceramics like zirconia are credible candidates due to their high fusion temperature. Through CFD methodology, calculations have been made in order to reproduce the fusion mechanism and kinetics of the UO2-ZrO2 system, in order to reproduce ablation phenomena of ZrO2 by UO2. This process is not only thermal but also chemical, as eutectic material formation is expected, around 2500°C. That is why the eutectic diagram of the UO2-ZrO2 system has been linearized and put directly inside the CFD software in order to take into account the formation of this eutectic material. Comparisons have been made with experimental data: a layer of UO2 is deposited inside a cooled zirconia crucible. Results show good correspondence between calculated and experimental data: the onset and effective melting of the zirconia is modelled, but also chemical saturation processes are identified, explaining the inhibition of the melting after a certain time. A practical application of this development has been made in the frame of the AREVA research program on sodium reactor, demonstrating that in the case of a jet of corium flowing down on an internal core catcher, the shape of the molten sacrificial material enables the apparition of a so-called “pool effect” being very favorable with respect to the local ablation of the core catcher. Finally, another application of this methodology could be in the frame of In-Vessel Retention study. A CFD simulation is made modelling the progression of the melting front inside the thickness of the bottom of a steel vessel due to the presence of molten UO2 emitting residual thermal power. Results show the formation of a floating steel layer at the surface of the UO2 molten pool, and the consecutive focusing effect occurring on the solid remaining parts of the steel vessel. All these calculations show that some of the complex thermo-chemical phenomena occurring during a severe accident can be modelled and used in order to give better understanding of the main phenomena.

Speaker: Mr Stéphane BEILS (AREVA NP)

Material Paper_FARGES_530.pdf

180 CHALLENGES DURING CONSTRUCTION OF SODIUM PIPING SYSTEMS FOR 500MWe PROTOTYPE FAST BREEDER REACTOR

Prototype Fast Breeder Reactor (PFBR) consists of Primary Sodium Circuit (PSC), Secondary Sodium Circuits (SSC), Safety Grade Heat Removal Circuits (SGDHRC) and Steam-Water circuit. The principal material of construction for sodium piping circuits is austenitic SS316LN/SS304LN. Manufacturing of thin and big bore piping with tight tolerances along with the high distortion in stainless steels due to high thermal expansion and low thermal conductivity makes fabrication extremely challenging. With strict rules of sloping to be given to the piping to make conducive for full draining of the sodium loops, the fabrication challenges become multifold. All sodium pipelines inside Reactor Containment Building (RCB) are provided with hot guard pipe and are inerted with nitrogen. The guard piping and the containment penetrations require sequential welding. Limited space at site for the erection of sodium piping along with welding at inaccessible areas with confined space makes the work all the more challenging. Terminal joints hook-up to tanks having frozen sodium inventory needs to be done meticulously adhering to highest level of industrial safety standards. The welding standards and acceptance criteria of PFBR sodium piping system is very stringent compared to conventional piping systems. Due to pyrophoric nature of sodium, the boundaries of various sodium piping systems must possess a high degree of reliability against failure. The welding of sodium piping systems are carried out by combination of Shielded Metal Arc Welding (SMAW) and Gas Tungsten Arc Welding (GTAW) process. Due to complex constructional features of the sodium piping systems, the argon gas purging, welding and non-destructive examinations are extremely difficult and challenging task. Apart from deployment of innovative purging methodologies, various special tools and fixtures were designed, developed and used for welding & fabrication. All the sodium pipe lines and components are provided with surface heaters, thermocouples, wire type leak detectors and insulation. Measurement of deflections of the sodium pipe lines during preheating and comparing with the analysis results is a vital step during the commissioning of sodium systems. This paper highlights on welding and fabrication aspects, challenges faced and innovations during construction of sodium piping circuits for 500MWe Prototype Fast Breeder Reactor. Key Words: Sodium piping, welding, fabrication

Speaker: Mr Rajan Babu Vinayagamoorthy (Director (Technical))

Material IAEA-CN-245-516.pdf

181 CHALLENGES IN THE FABRICATION AND RECYCLING OF MIXED CARBIDE FUEL

Sudhir Mishra1, Amrit Prakash1, K. B. Khan1 , Vivek Bhasin2 Radiometallurgy Division, 2Nuclear Fuels Group, Bhabha Atomic Research Centre, Mumbai- 400 085, India. E-mail: sudhir@barc.gov.in Abstract Mixed (U0.3, Pu0.7)C fuel is the driver fuel for Fast Breeder Test Reactor (FBTR ) at Kalpakkam, India. This fuel is being fabricated at Radio metallurgy Division, Bhabha Atomic Research Centre (BARC). The reactor was made critical with Mark-I fuel having composition (U0.3, Pu0.7)C in year 1985. The fuel has seen a maximum burn up of 165Gwd/t. The carbide fuel is phyrophoric in nature and very much susceptible to hydrolysis. Hence the handling of fuel is done inside alpha leak tight glove-boxes having N2 as cover gas. The fuel is fabricated by classical powder pellet route. In the recent past a new fuel fabrication facility has been commissioned and improvement over the existing equipments and process steps have been carried out to make the fuel fabrication process more efficient resulting in higher productivity and lesser contact between personnel and radioactive powder. The use of liquid binder and lubricant has eliminated dewaxing step from the process flow sheet for UC pellet fabrication. Dry recycling of the fuel is carried out on regular basis by oxidizing the mixed carbide powder. Chemically accepted pellets having physical defects are directly recycled by crushing and milling the pellets to powder form and subsequently following other regular process steps to produce sintered pellets.

Speaker: Mr SUDHIR MISHRA (BHABHA ATOMIC RESEARCH CENTRE, INDIA)

Material Challanges_in_Mixed_carbide_.docx

182 Change in Mechanical Properties of Spent Fast Reactor Claddings

During fuel element operation changes in structure and physical and mechanical properties of the claddings are induced by irradiation and other factors. In particular, there are such changes as swelling, embrittlement, softening and corrosion damages. To predict cladding limit state it is necessary to know the changes occurred. In particular, it is important that cladding mechanical properties after operation are properly determined. In this respect there are different mechanical tests. Tensile test of annular cladding specimens is a conventional method. Test stress strain state essentially differs from that occurring in claddings during operation when they are subjected to gas pressure and deformation from the swelling fuel. Mechanical properties determined with annular sample tensile test are too conservative and cannot be used for a proper description of the cladding behaviour during operation. At JSC "INM" a technique for mechanical testing with tough plastic aggregate internal pressure has been developed. Aggregate compression leads to its plastic deformation exerting internal pressure on the cladding tubular sample. Mechanical properties of the cladding material are calculated according to the recorded 'movement of aggregate compressing plungers - compression force' curve. During the test a loading pattern and a stress strain state of the cladding simulate its loading under irradiation in the reactor. Characteristics of tubular samples tested with internal pressure clearly demonstrate operated cladding behaviour. The paper shows the results of the short-term mechanical properties change after irradiation in fast reactors obtained for the mentioned techniques. Mechanical characteristics after testing in different loading patterns are compared. Advantages and disadvantages of each technique and possibility of their integration to predict cladding behaviour during operation are pointed out.

Speaker: Ms Svetlana Barsanova (Joint Stock Company «Institute of Nuclear Materials»)

Material IAEA-CN-245-265.pdf

183 Chugging boiling in low-void SFR core: new phenomenology of unprotected loss of flow

Calculational analysis of the unprotected loss of flow (ULOF) accident in a Generation-IV SFR, featuring a low-void core design, shows that the chugging sodium boiling regime in the core could last for several hundred seconds during the accident. While in the case of the traditional positive-void SFR core the sodium boiling onset is almost immediately followed by the power run-away, fuel bundle overheating, melting and relocation (i.e. severe accident), the chugging boiling regime in the low-void SFR core could allow avoiding the power runaway and avoiding or at least significantly postponing the cladding overheating and melting caused by the permanent dryout. The low-void core design therefore could be classified as a new safety measure acting as a level of defence preventing the severe accidents. The state-of-the-art in the area of the chugging regime of the sodium boiling is very limited and very few corresponding experiments were performed. The paper will present the detailed transient analysis of the low-void core behaviour in unprotected loss of flow accident performed with the TRACE code (modified for the sodium boiling modeling) and discuss the physics of the predicted phenomena as well as the future research needed, including new experiments.

Speaker: Dr Konstantin Mikityuk (Paul Scherrer Institut)

Material FR17_Mambelli_v03.pdf

184 Code Qualification Plan for an Advanced Austenitic Stainless Steel, Alloy 709, for Sodium Fast Reactor Structural Applications

Sodium Fast Reactor (SFR) is one of the leading advanced reactor concepts that would provide a low-carbon energy option to a diverse U.S. power sources. Nuclear energy releases zero carbon emissions during electricity production, and thus is essential in reducing CO2 emissions from the U.S. power sector. SFR also supports other possible missions, including recycling of used fuel for closing the fuel cycle. Improved structural material performance is one way to improve the economics of SFRs; by increasing thermal efficiency, power output, and design lifetimes of the reactor system. Improved performance and reliability of structural materials could also enable greater safety margins and more stable performance over longer times, and reduce down time of the reactor plant. Advanced materials could also spur improvements in high temperature design methodologies and thereby allowing design simplifications and more flexibility in plant operations. Thus, they could have a significant, positive impact on levelized electricity production cost even if the commodity costs for the advanced materials are higher. Capital cost reduction and improvement in economic return are important incentives for commercial deployments of SFRs. Alloy 709 is an advanced austenitic stainless steel with enhanced creep strength relative to Code-approved reference construction materials (Type 304 and 316 stainless steels) and that makes it an attractive candidate material for SFR structural applications. In this paper, some preliminary data for Alloy 709 will be presented and a qualification plan for developing an ASME nuclear code case will be reviewed.

Speaker: Mr Krishnamurti Natesan (Argonne National Laboratory)

Material FR17_Paper-Alloy_709.docx

185 Complex discussion of inherent safety fast reactors start-up with enriched uranium concept (strategical, economical aspects, problems of neutron physics etc.). R&D program proposal

Due to the growing population of Earth, the development of a full-scale nuclear power industry is becoming an increasingly challenging task in the 21st century and onwards. The Breakthrough («Proryv») Project is focused on inherently safe fast reactors which are expected to resolve, for a first time, the economic competitiveness problems of the nuclear power sector. In order to develop a full-fledged nuclear power industry based on such reactors within acceptable timeframe, these reactors must first be put into operation with enriched uranium. The article provides the results of systemic calculations confirming this thesis. Moreover, it supports the economic benefits (in the nearest future) of the uranium-based start of fast reactors versus the uranium-plutonium start. For the first time, it demonstrates the possibility of a noticeably simpler transition from uranium fuel-based start to uranium-plutonium regime in the closed fuel cycle compared to the previous alternatives (reduced number of structural changes in the core during the transient mode, less restrictive requirements to the start load, etc.). An R&D program is proposed in order to justify the start of inherently safe fast reactors on enriched uranium.

Speaker: Mr Michael Orlov (Private institution «Innovation and technology center for the «PRORYV» project»)

Paper Complex_discussion_of_FR_uranium_start-up_(Orlov_M.).docx

186 Computational investigation of nuclear waste incineration efficiency in a subcritical molten salt driven by 50-100 MeV protons

Molten salt reactors were designed and operated at 1960s. The subcritical accelerator driven MSRs are being considered recently. In the present work, accelerator driven homogeneous subcritical core configuration was Modelled using MCNPX code. The composition of NaF-BeF2-ThF4-TRUF4 and NaF-233UF4-ThF4-TRUF4 was selected as the fuel loaded inside a 58×60 cm cylindrical core respectively. NaBF was selected as coolant salt of the fuel salt circuit. Accelerated proton particles were used to induce fission in the transuranic nuclei. The projectiles energy was changed from 50 MeV up 100 MeV in five steps. TRU fission rate, deposited heat distribution and neutron flux distribution were determined inside the subcritical core. Neutron and proton flux distribution inside the subcritical molten salt core was compared with each other. Energy gain, source multiplication factor and proton importance parameters were calculated for any different projectile energy. Optimized proton energy was suggested to be applied for nuclear waste incineration using such system. Burn-up calculations were carried out for the cores with different fuel loading.

Speakers: Dr Seyed Mohammad Mirvakili (AEOI), Dr Zohreh Gholamzadeh (AEOI)

Paper Contribution_template_893.784304819_c126_(1).docx

187 Computational modelling of inter-wrapper flow and primary system temperature evolution in FBTR under extended Station Blackout

To handle a station black-out (SBO) event, sodium cooled fast reactors are equipped with passive systems to remove decay heat, which do not require external power source. The duration of SBO can extend up to a week. Decay heat removal in FBTR, depends on natural convection, driving flow through the core subassemblies and inter-wrapper spaces. The relatively low decay heat and the extended duration of SBO presents a risk of coolant freezing in the primary circuit inlet pipes. This impacts coolability of core, as flow through subassemblies would be impeded and major heat removal path would become ineffective. It becomes essential to understand heat transfer to reactor vault, temperature distribution in the primary sodium pool and evolution of inter-wrapper flow and clad temperature. The present study aims to develop mathematical models for this scenario and investigates thermal hydraulics of the reactor. The study assumes that internal flow through subassemblies is impeded and only inter wrapper space is available for heat removal from core. Heat sink is provided by primary vessel sodium plenum and surrounding structural components. Ultimate heat sink is atmosphere. A two dimensional axisymmetric Computational Fluid Dynamics model of primary plenum is developed. Influence of reactor subassemblies on inter-wrapper flow, which act as primary heat removal path is modeled with the aid of appropriate momentum and heat sinks. Decay heat generation in core is obtained from available reactor physics calculations. Thermal effects of structures surrounding the reactor vessel (thermal capacity and resistance) are modeled using a one dimensional lumped parameter model, which supplies boundary conditions to the reactor vessel model. Flow in primary reactor plenum is seen to be controlled by natural convection in the inter-wrapper space, carrying heat from reactor core to main plenum (above core region), which dissipates heat to the surrounding structures through reactor vessel. Flow and temperature in reactor plenum during duration of the transient is predicted and compared against available safety limits. These predicted temperatures are extended using a two dimensional model, to predict fuel and clad temperatures inside fuel subassemblies. Results reveal that fuel clad temperatures reach design safety limit(s) after two days. It is concluded that adequate time is available for deploying Emergency Diesel Generator(s) and initiation of double envelope cooling. Full paper would present the sequentially coupled thermal hydraulic model, evolution of inter-wrapper flow and finally clad temperature as a function of time.

Speaker: Dr Velusamy Karuppanna Gounder (Indira Gandhi Center for Atomic Research)

Material All the suggestions of the reviewers have been included in this revised draft

188 Controlling FCCI with Pd in metallic fuel

A major factor limiting the lifetime of U-Zr based fuel is fuel-cladding chemical interactions (FCCI). As the fuel is burned, fission product lanthanides (Ln) interact with the Fe-based cladding, causing thinning of the cladding wall and eventual breach of the cladding. In order to extend the lifetime of the fuel in reactor, FCCI must be controlled. Palladium has been shown to be a promising metallic fuel additive to control FCCI due to the stable Pd-Ln intermetallics formed. The current investigation is focused on the characterization of U-Zr-Pd fuel, with and without added lanthanides. Characterization includes as-cast fuel as well as annealed fuel, and comparison to recent postirradiation examination results from U-10Zr fuel. Preliminary diffusion couple results between the fuel (with and without Ln) and iron will also be presented.

Speaker: Dr Michael Benson (Idaho National Laboratory)

Material FR-17_UZrPd_anneal_manuscript.doc

189 Corrosion behavior of tube steel for BREST-OD-300 steam generator

Abstract Once-through type steam generator (SG) is designed for BREST-OD-300 reactor unit. There is the steam overheating zone in the upper part of long-length heat exchange tubes. Consequently, under service conditions SG tubes to be exposed to three corrosion media – liquid lead (primary coolant), pressurized water, and superheated steel at temperatures in the range 350-505 °C, pressure of 17 MPa. To confirm appropriate corrosion resistance SG tubes material in NIKIET, CRIFM, CRISM Prometey research work has been conducting. Required corrosion properties of SG tube material are obtained due to usage of specially alloyed austenitic stainless steel (SS). This paper presents the results of corrosion and corrosion-mechanical behavior studies, and material investigation results. Type 321 SS is used for comparison. Use of modified SS as the material of SG tubes allows obtaining required corrosion properties for established SG service-life. Keywords: austenitic stainless steel, corrosion, corrosion-mechanical properties, liquid lead, water coolant.

Speaker: Mr Kirill Shutko (N.A. Dollezhal Research and Development Institute of Power Engineering)

Paper 542_Corrosion_behavior_of_tube_steel_for_BREST-OD-300_steam_generator_(revised).docx

190 CORROSION OF 12X18H10T STEEL IN Ce-, Nd- AND U-CONTAINING MOLTEN LiCl-KCl EUTECTIC

The present work is aimed at the study of the 12Х18Н10Т steel corrosion in the molten LiCl-KCl eutectic, which contains different proportions of CeCl3, NdCl3 and UCl3. The CeCl3 and NdCl3 concentrations varied within the interval of 0.2-5.0 mol.%, and the UCl3concentration varied within the interval of 1.0-2.5 mol.%. The temperature of the experiments was 500 0C. The composition of the melts under study was close to the composition of real electrolytes, which appear at the nitride SNF processing. The basic method of study is the gravimetric method with the exposure time from 24 to 100 hours. Atomic-adsorption, micro X-ray spectral and X-ray structural methods were used for samples analysis. The first component of selective dissolution in the steel under study was Fe. Chromium and manganese dissolution degrees were the smallest. The presence of UCl3 in the melt was found to have the largest impact on corrosion. The corrosion rate is rather small. For example, in the melt containing 1 mol.% of NdCl3 the corrosion is equal to 1.93 g/(m2∙h), and in the (LiCl-KCl)eut. + 1%CeCl3 +1%NdCl3 + 1%UCl3 melt it is 2.61 g/(m2∙h). The corrosion mechanism was found to be electrochemical.

Speaker: Dr Evgeniya Nikitina (Institute of High Temperature Electrochemistry)

Material Corrosion_of_12Cr18Ni10Ti_steel_...__Nikitina_E.,_Russia.doc

191 DECAY HEAT REMOVAL SYSTEM IN THE SECONDARY CIRCUIT OF THE SODIUM-COOLED FAST REACTOR AND EVALUATION OF ITS CAPACITY

Decay heat removal system (DHRS) option for the secondary circuit of the sodium-cooled fast reactor (SFR) by means of air cooling the outer surface of piping and equipment of heat removal loops of the SFR secondary circuit is proposed. The DHRS option under consideration implies case mounted around main piping and equipment of heat removal loops of the SFR secondary circuit and divided into a number of sections connected in parallel to each other with an exhaust chimney. This case performs certain containment function under normal operation condition, and it is arranged a natural circulation of air through the gap between piping and equipment of the secondary circuit and this case under emergency cooling modes by opening the air dampers. Effectiveness of this decay heat removal system is evaluated by using specially developed computational code that allows modeling transient emergency cooling modes and optimization of the DHRS characteristics to reduce the maximum value of coolant temperature in these transients. Results of effectiveness evaluation for the proposed decay heat removal system applied to fast reactor with sodium coolant are presented.

Speaker: Iurii Ashurko (IPPE)

Material IAEA-CN-245-161(Ashurko).pdf

192 Decay-heat removal in accidents in fast reactors with liquid metal coolant

The problem of decay-heat removal from a shutdown reactor is still pending and Fucushima accident proved it. The complexity of this problem grows with the increase reactor power. High power reactors with sodium and lead coolants were analyzed and compared in terms of decay -heat removal using 3D thermo-hydraulic calculations of reactor cooldown. Two DHRS designs are compared that differ by the location of emergency heat-exchanger. In the first design emergency heat-exchanger is located in the upper reactor chamber and heat is removed from reactor core due to following circulation path: “emergency heat-exchanger – upper plenum – inter-wrapper space of reactor core – upper plenum”. In the second design emergency heat-exchanger is located in the downtake slit of reactor and design includes backflow valve that in cooldown mode allows “hot” coolant from the upper plenum to enter emergency heat-exchanger and blocking this flow while the reactor operates in power mode. DHRS of a sodium reactor results to be more effective for both DHRS designs. As for the lead cooled reactor the second DHRS design also allows to remove after-heat without exceeding the allowed temperature limits. With the 1st DHRS design fuel rods overheat for a short period of time.

Speaker: Mr IURII SHVETSOV (Private institution «Innovation and technology center for the «PRORYV» project»)

Material FR-17_CN-245-_192(Shvetsov)Rev.docx

193 Design and Development of Stroke Limiting Device for Control & Safety Rod Drive Mechanisms (CSRDMs) of future FBRs

The core degradation due to Anticipated Transients Without Scram (ATWS) has to be practically eliminated as the fast reactor core is not in its most reactive configuration during normal operation. The ATWS events can lead to early and large release of radioactivity. Deterministic demonstration of dispersal of fuel to avoid further core compaction after initiation of large scale core damage is almost impractical. Hence the following are the important decisions related to shutdown systems for next generation SFRs to facilitate practical elimination of core degradation due to anticipated Transients Without Scram. i) Strengthen the first two shutdown systems by addition of passive/active features. ii) Introduce an additional shutdown system, which is completely diverse, independent, passive & confined within core sub-assembly. This shall come into action on failure of first two systems. In this paper design augmentation of first Active Shutdown system by addition of a safety device is discussed. The first shutdown mechanism of future FBR is Control and Safety Rod Drive Mechanism (CSRDM). A Stroke Limiting Device (SLD) is provided in CSRDM of future FBR to prevent inadvertent withdrawal of Control & Safety Rods beyond a pre-set level and thereby limit the consequences of inadvertent control rod withdrawal event well within limits even with the failure of other safety actions. SLD limits the consequences of inadvertent withdrawal of CSR by physically limiting the withdrawal stroke length of CSR to 20 mm. Two different concepts of SLD have been conceived, manufactured and standalone endurance tested. One of them has also been integrated with CSRDM and tested. Based on the above, Mark-II design of SLD with some additional design improvements is being developed. The final design of SLD (Mark - II) will be adopted in CSRDM for future Fast Breeder Reactors. The details of two concepts developed, testing carried out as part of design validation and design improvements being incorporated in Mark-II design are presented in this paper.

Speaker: Mr RAGHUPATHY S. (Indira Gandhi Centre for Atomic Research, Kalpakkam)

Material FR17_327.docx

194 Development of Fast Reactors in the USSR and the Russian Federation; Malfunctions and Incidents in the Course of their Operation and Solution of Problems.

The initial idea of potential nuclear fuel breeding originated in the USA and the first success in development of fast reactors designed for implementation of this idea was achieved there. With a very small delay, similar studies started in the USSR; however, that was the place where fast reactor development reached its peak. The chief scientific supervisor of these research studies in the USSR was A.I. Leypunsky. Great achievements in this area were made by scientists and engineers from France and the UK. After completion of nuclear weapon tests, A.I. Leypunsky sent a Position Paper to the First Chief Directorate, where he stated the principal physical ideas and the high-priority tasks on fast reactors. These proposals were approved in the Government Decree of 1950. It was followed by development and construction of the critical facility BR- 1, research reactor BR-2 , research reactor BR-5 (BR-10), critical facility BFS-1, research reactor BOR-60, critical facility BFS-2, the pilot and demonstration power reactor BN-350, power reactor BN-600 and power reactor the BN-800. Over the past 60 years operation of research and power reactors in our country has accumulated vast experience, including abnormal and emergency situations, their causes and ways of overcoming them, ensuring reliable and safe operation of fast reactors with sodium coolant. In the future – assimilation of the closed fuel cycle, NPPs with the BN-1200 reactors, ensuring competitive economy.

Speaker: Mr Lev Kochetkov (Alexeevich)

Paper Development_of_fast_reactors_ID579_(002).docx

195 Development of the U.S. Sodium Component Reliability Database

With the advent of the use of Probabilistic Risk Assessments (PRAs) for safety analysis of Light Water Reactors (LWRs) in the 1970s, the SFR community used PRA as a tool which can demonstrate the safety of SFR designs while avoiding the pitfalls associated with an over-reliance on highly conservative safety requirements. Throughout the 1970s, 80s, and 90s, the US compiled sodium reactor specific PRA information into the Centralized Reliability Database Organization (CREDO) database, maintained by Oak Ridge National Laboratories in collaboration with the Japanese Atomic Energy Agency (JAEA). Unfortunately, the funding for the CREDO database was cut in the 1990s and the database was lost and was regained in August of 2016. This paper will describe three databases being developed at Sandia National Laboratories(SNL): 1. CREDO-I – A summary of the state of the original CREDO database; 2. CREDO-II – Early attempts by Argonne National Laboratory (ANL) and SNL to recreate the CREDO database from operational documents; 3. The future combination of the CREDO-I and CREDO-II databases into a unified database.

Speaker: Dr Matthew Denman (Sandia National Laboratories)

Material FR17_CREDO.pdf

196 Development of Ultra Sub-size Tensile Specimen for Evaluation of Tensile Properties of Irradiated Materials

The idea of using small specimens for mechanical testing had actually originated in the nuclear industry to cater to the irradiation material testing and reactor surveillance programs. Small or miniaturized specimens ensure efficient use of available irradiation volume in nuclear reactors, reduce uncertainty in irradiation parameters due to flux and temperature variations and also reduce radiological hazard during testing. A procedure for tensile testing employing a miniature tensile specimen called ultra sub-size (USS) tensile specimen carved out of a 10.0 mm diameter and 0.5mm thick disc sample has been developed and standardised. The geometry of the specimen was optimized using Finite Element Analysis (FEA) with the purpose of maintaining stress concentration in the fillet radius and gripping area equivalent to that in standard ASTM and sub-size tensile specimen. FEA was also employed to evaluate the allowable fabrication tolerances for gage width and thickness of USS by examining its effects on the stress strain curves obtained and comparing with that for standard ASTM and sub-size tensile specimens. USS specimens along with ASTM standard and sub-size specimens were tested on a range of fast reactor structural materials for comparison of mechanical properties. Due to difficulty of employing extensometer on a small gage length of 3.0mm, digital image correlation (DIC) was employed for strain measurement. The strain obtained through DIC was co-related with that obtained from the cross-head displacement of the UTM. Online strain distribution was extracted from DIC images to study nature of strain distribution over the USS specimen during tensile test and was compared with that obtained from standard specimens. The results obtained from tensile testing of ultra sub-size specimen at ambient and elevated temperatures were found to be consistent and comparable with those obtained from standard specimens for a wide range of alloys examined. The scatter obtained in the UTS, yield strength and uniform strain values evaluated from USS specimen is comparable to that obtained from standard and sub-size samples. The uncertainty in the YS and UTS values from USS specimen were evaluated and compared with that of the ASTM standard and sub-size specimens. The results of this study show that USS tensile specimen geometry can be reliably employed for mechanical property evaluation of irradiated structural materials. Further efforts are required to formulate standards through round robin testing, before the technique can be deployed in the field.

Speaker: Mr Puthiyavinayagam Pillai (Indira Gandhi Centre for Atomic Research)

Material FR17-Paper-PIED-03-USS_tensile_testing-170131-AVK.docx FR17-Paper-PIED-03-USS_tensile_testing-170131-AVK.pdf

197 ECOLOGICAL ASPECTS OF THE USE OF FAST REACTORS IN A CLOSED NUCLEAR FUEL CYCLE UNDER THE “PRORYV” PROJECT

The development of nuclear power engineering with closing of a fuel cycle and the use of fast reactors must ensure a higher level of ecological safety of the population and the environment. The highest ecological effect is achieved by recycling of spent fuel and isolation of long-lived radionuclides (90Sr, 137Cs and 99Тс) and transmutation of 99% of americium. In normal operation of CNFC facilities exposure doses to the population are formed via different critical pathways: for a rector plant – due to inhalation intake of 3Н, for fabrication and refabrication module – due to inhalation of Pu aerosols, for SNF recycling module – due to external radiation from the soil and ingestion in case of surface contamination of plants. The use of nitride fuel generates large amounts of 14С (270 g per 1 ton fuel). Insolubilizing most of 14С ensures compliance with the project standards for the population exposure to the gas phase of release.

Speaker: Mr RUDOLF ALEXAKHIN (INSTITUTION «ITC «PRORYV» PROJECT»)

Material RADIOLOGICAL ASPECTS OF THE USE OF FAST REACTORS IN A CLOSED NUCLEAR FUEL CYCLE UNDER THE “PRORYV” PROJECT

198 ELECTRICAL CONDUCTIVITY OF MOLTEN LiCl-KCl EUTECTIC WITH COMPONENTS OF SPENT NUCLEAR FUEL

Pyroelectrochemistry is one of the most prospective approaches for spent nitride nuclear fuel (nitride SNF) reprocessing. It includes the anodic dissolution of the SNF pellets into the molten LiCl-KCl eutectic with the subsequent electrochemical separation of U, Pu, Np, Am from other constituents. Physical-chemical properties of such complex melts are still insufficiently studied. The electrical conductivity of a number of quasi binary melts (LiCl-KCl)eut., containing CeCl3, NdCl3, UCl3, as well as CsCl and CdCl2, are studied in detail in the present work. The majority of measurements were performed in the whole concentration range and in the wide temperature range from the liquidus point to 900 - 920 0C. The electrical conductivity of a number of 3-4 component (LiCl-KCl)eut. - CeCl3 - NdCl3 - UCl3 mixtures was measured. The density of the melts under study was evaluated and their molar conductivity was calculated. The liquidus line of these salt systems was built using the polytherm breakpoints. These data are required to create a general model for the electrolyzer operation and to develop complied technological equipment. This research was partially supported by the Russian Ministry of Education and Science through Targeted Federal Program (project number: 14.607.21.0084).

Speaker: Dr Alexander Salyulev (Institute of High Temperature Electrochemistry)

Material Electrical_Conductivity..._Salyulev,_Russia.doc

199 ESFR-SMART: new Horizon-2020 project on SFR safety

To improve the public acceptance of the future nuclear power in Europe we have to demonstrate that the new reactors have significantly higher safety level compared to traditional reactors. The ESFR-SMART project (European Sodium Fast Reactor Safety Measures Assessment and Research Tools) aims at enhancing further the safety of Generation-IV SFRs and in particular of the commercial-size European Sodium Fast Reactor (ESFR) in accordance with the ESNII roadmap and in close cooperation with the ASTRID program. The project aims at 5 specific objectives: 1) Produce new experimental data in order to support calibration and validation of the computational tools for each defence-in-depth level. 2) Test and qualify new instrumentations in order to support their utilization in the reactor protection system. 3) Perform further calibration and validation of the computational tools for each defence-in-depth level in order to support safety assessments of Generation-IV SFRs, using the data produced in the project as well as selected legacy data. 4) Select, implement and assess new safety measures for the commercial-size ESFR, using the GIF methodologies, the FP7 CP-ESFR project legacy, the calibrated and validated codes and being in accordance with the update of the European and international safety frameworks taking into account the Fukushima accident. 5) Strengthen and link together new networks, in particular, the network of the European sodium facilities and the network of the European students working on the SFR technology. Close interactions with the main European and international SFR stakeholders (GIF, ARDECo, ESNII and IAEA) via the Advisory Review Panel will enable reviews and recommendations on the project’s progress as well as dissemination of the new knowledge created by the project. By addressing the industry, policy makers and general public, the project is expected to make a meaningful impact on economics, environment, EU policy and society. The full paper will present the project in details

Speaker: Dr Konstantin Mikityuk (Paul Scherrer Institut)

Material IAEA-CN-245-_450_v02.pdf

200 Evaluation of Anticipated Transient without Scram for SM-SFR using SAS4A/SASSYS-1

Small Modular Sodium-cooled Fast Reactor (SM-SFR) was developed in UNIST as a breeder reactor with the target of ultra-long cycle operation. The depletion analysis and quasi-static reactivity balance analysis were performed to see its inherent safety in the neutronics point of view. In this study, the inherent safety evaluation is performed in the thermal-hydraulic point of view by using transient analysis for LMR code SAS4A/SASSYS-1 which was developed in Argonne National Laboratory. Three major events of Anticipated Transient without Scram (ATWS) were tested for this research; Unprotected Loss of Flow (ULOF), Unprotected Loss of Heat Sink (ULOHS), Unprotected Transient Over Power (UTOP). Every perturbation for each transient event occurs at 10 second and each simulation time is 100 minute. The power to flow change, the reactivity profiles, and the temperature changes were investigated to trace each transient trend. It has been confirmed that SM-SFR has inherent safety from the fact that any of the events doesn’t have a clad failure or a coolant boiling.

Speaker: Prof. Deokjung Lee (Ulsan National Institute of Science and Technology)

Paper Fullpaper_FR17_TaewooTak_review.pdf

201 EVALUATION OF COBALT FREE COATINGS AS HARDFACING MATERIAL CANDIDATES IN SODIUM FAST REACTOR

The need for materials having good tribological properties in Sodium Fast Reactors has been identified (SFR) from the first reactors operation. Where galling or adhesive wear cannot be tolerated, hardfacing alloys or galling-resistant coatings are usually applied on rubbing surfaces. The most used coating is the cobalt base alloy named Stellite because of its outstanding friction and wear behavior. Nevertheless, cobalt is an element which activates in the reactor leading to complex management of safety during reactor maintenance and decommissioning. As a consequence, a collaborative work between CEA, EDF, AREVA and French academic laboratories has been launched for selecting promising cobalt free hardfacing alloys for SFR applications. Several nickel base alloys and aluminides have been selected from literature review then manufactured on two candidate steel grades: 9Cr ferritic-martensitic steel EM10 and 18Cr austenitic steel AISI 316L(N). Nickel base alloy coatings were deposited through Plasma Transferred Arc or Laser Cladding, and the aluminides coatings, through pack cementation or slurry. Among the numerous properties required for qualifying their use as hardfacing alloys in SFR, good corrosion behaviour and good friction and wear behaviour in sodium are essential. The results obtained on these properties are shown in this presentation. First, the corrosion behaviour of all coatings was evaluated through exposure tests in purified sodium for 5000 h at 200 °C and 550 °C. The degradation of the surface was carefully measured thanks to several complementary analysis techniques (GD-OES, FESEM, XRD, …). Finally, the friction and wear properties of all candidates were evaluated in sodium in a newly designed facility. The influences of temperature and of oxygen content in sodium on these properties are detailed.

Speaker: Dr FABIEN ROUILLARD (CEA)

Paper Proceeding_FR17_final_paper_Rouillard_IAEA-CN-245-126_rev.pdf

202 Evaluation of irradiation-induced point defects migration during neutron irradiation in modified 316 stainless steel

For the development of nuclear core materials, especially fuel cladding tube, in sodium-cooling type fast breeder reactor, void swelling suppression is one of the most important issues to keep the dimensional stability in reactor. A large number of theoretical and experimental investigations on void swelling behavior have been carried out, and void swelling directly depends on the diffusion of point defects induced by neutron irradiation as well as the strength of point defect sinks such as dislocations and precipitates. Evaluations of the point defects diffusion in metal during neutron irradiation have been qualitatively done through various researches, however the quantitative estimation is hardly performed due to the difficulty of in-situ experiments during neutron irradiation. Instead of that, the indirect estimations from the temperature dependence measurements of dislocation loop densities and growth rates using electron in-situ observation are often carried out, but the irradiation correlation between electron and neutron irradiations, such as the differences of irradiation dose rate and damage morphology, should be discussed with accuracy. Therefore, the evaluation of point defects diffusion, especially vacancy migration, during neutron irradiation by the other method was tried in this study. In detail, from already neutron-irradiated microstructures, vacancy migration energy was estimated using the knowledge that void denuded zone (VDZ) widths formed near random grain boundaries depend on temperatures. The test material was PNC316 steel, which is the modified 316 stainless steel with cold-working and additives to improve the void swelling resistance. The fuels assemblies composed of PNC316 steel were irradiated in the experimental fast reactor JOYO. For the PNC316 specimens cut from these assemblies, which were irradiated at temperatures from 722 K to 821 K and doses of 74.5–87.5 dpa, VDZ widths were analyzed from the transmission electron microscope observations and the temperature dependence was investigated. As the result, VDZ widths increased with increasing temperature. From the Arrhenius plots of VDZ widths and the reciprocal temperatures, the vacancy migration energy during neutron irradiation in PNC316 steel was quantitatively estimated to be about 1.4-1.5 eV. As vacancy migration energy in Fe-Cr-Ni model alloy is about 1.05 eV, the value of PNC316 steel implies that the vacancy mobility is low as a result of interaction of vacancies with minor alloying elements.

Speaker: Mr Yoshihiro Sekio (Japan Atomic Energy Agency)

Material Contribution_template_(JAEA_Sekio_IAEA-CN-245-_254)_170113.pdf

203 Examination of ChS-68 Steel Used as a BN-600 Reactor Cladding Material

Austenitic stainless steel has been used as a standard material for BN-600 fast reactor claddings for many years. High-temperature strength is one of the advantages of austenitic steels over ferritic-martensitic ones. Nevertheless, corrosion damages, radiation-induced swelling, creep, embrittlement, and strength reduction are topical problems for claddings made of austenitic steels. In this respect at high burn-up levels swelling is the main problem limiting operation of the material. Originally thoroughly studied EI-847 steel was used for developing new austenitic steels. After boron modification EP-172 steel was obtained. ChS-68 steel doped with elements reducing radiation-induced swelling, such as boron, silicon, and titanium, was based on EI-847 steel. At the initial stage significant radiation-induced deformation of fuel assembly elements was one of the factors limiting fuel burnup to 7.2 % FIMA and damage dose to 44 dpa. Transition to new steels and reactor core modifications made it possible by 2000 to attain burnup level of 9.2 % FIMA and damage dose level of 73 dpa per cladding. ChS-68 properties optimization was carried out by High-technology Research Institute of Inorganic Materials (VNIINM) using the results of post irradiation examination made by Beloyarsk NPP and ROSATOM materials testing enterprises, including Institute of Nuclear Materials (INM). Irradiated in BN-600 claddings of standard and test fuel assemblies, as well as materials test assembly samples (ChS-68 and other austenitic steels, similar in composition) were examined at INM hot cells. Characteristics and properties of claddings made with variation of steel chemical composition within specifications, at different cold work levels, according to different metal manufacturing and tube production technologies, were determined. Results of post irradiation examination of physical, mechanical and corrosion properties, and structural characteristics were used to analyze processes leading to structure and properties changes and to predict residual life. On the basis of the examination INM researchers have published a number of articles and presented a number of papers at the conferences of different levels. Using the obtained results VNIINM in collaboration with Machine-Building Plant (MSZ) has improved cladding manufacturing technology. It resulted in a stepwise (during 15 years) extension of standard fuel assembly service life up to 73, 78, 82, 87 dpa with burnup increase from 9.2 to 13.2 % FIMA. It is expected to increase damage dose at least to 92 dpa and fuel burnup to 14-15% FIMA.

Speaker: Mrs Natalia Glushkova (Joint Stock Company «Institute of Nuclear Materials»)

Material IAEA-CN-245-80.pdf

204 Experience on MOX fuel fabrication for fast reactor at PFPF

Japan Atomic Energy Agency has developed mixed plutonium-uranium oxide (MOX) fuel fabrication technologies in large-scale and fabricated MOX fuel assemblies for experimental fast reactor “JOYO” and prototype fast reactor “MONJU” at Plutonium Fuel Production Facility (PFPF) since 1988. Low density pellet is adopted as MONJU fuel. For the low density pellet fabrication in large-scale, various challenges were encountered. Typical examples of the challenges are as shown below; 1. Thermal degradation of organic compound used as pore former 2. Large standard deviation of pellet density due to inhomogeneous dispersion of pore former in granulated MOX powder In order to resolve these challenges, countermeasures such as new pore former with high softening temperature and improved granulation method for MOX powder were considered. In this presentation, accumulated MOX fuel fabrication technologies as mentioned above and recent R&D activity for low-decontaminated TRU fuel fabrication such as new pelletizing method, or die wall lubrication pelletizing, will be discussed.

Speaker: Mr Kiichi Suzuki (Japan Atomic Energy Agency)

Material suzuki.kiichi

205 Extending the grid plate life - Incorporation of lower axial shield for FBTR

Operational life of Fast Breeder Test Reactor (FBTR) is limited by the grid plate life. An irradiation experiment was carried out in FBTR to determine changes in the mechanical properties of specimens of grid plate material at the desired low fluence irradiation conditions. Based on the analyses of these experiments and flux measurements at the grid plate location, the residual life of FBTR was estimated to be 6.52 EFPY at the end of 18th campaign. Possibility of reducing the neutron damage by including lower axial shields has been considered. Neutronics studies on the effectiveness of materials such as tungsten, tungsten carbide, boron carbide and ferro-boron have been conducted. A suitable arrangement of enriched boron carbide and stainless steel has been analyzed too. Based on these studies, tungsten carbide emerges as the best option. Chemical and metallurgical studies indicate that the material is compatible with sodium, has good thermo-physical properties and, hence suitable for introducing in the FBTR core as the lower axial shield. On implementation, it is expected that the life of FBTR would be increased by 35% of its remnant life.

Speaker: Mr Sureshkumar KV (RFG, IGCAR)

Material FR17_JUNE_PAPER_1D_169.pdf

206 Fabrication process of NpO2 pellets

In order to increase dissolution ratio of the irradiated NpO2 targets, it’s necessary to add a little diluent into NpO2 pellet. In this paper, pressureless sintering processes and microstructures of NpO2-10% CaO, NpO2-10%SrO, NpO2-10%MgO and NpO2-5%MgO pellets were studied, sintered at 1730℃ for 2 hours in Ar-5%H2 gases. Only NpO2 solid solution phase structure was found in all the pellets. NpO2-10%CaO pellet melts at the sintering process. NpO2-10%SrO pellet has a sintered density of 60.0% TD with cracking and porous microstructures. NpO2-10%MgO pellet has a sintered density of 83.1%TD with irregular grains. NpO2-5%MgO pellet can be sintered to 90.0%TD with cobble grains. Density of NpO2-5%MgO pellet will increase to 92.5%TD using UO2 powder embedded sintering process.

Speaker: Mr Bangyue YIN (CIAE)

Material Fabrication_process_of_NpO2_pellets__IAEA_YINBANGYUE.docx

207 FACILITY FOR ADVANCED FUELS THROUGH THE SOL-GEL METHOD

Advanced fuels forms need to be developed for use with the evolving nuclear fuel cycles. The fabrication of sphere-pac fuel pins that employ microspheres of U,Pu mixed oxide (MOX) or microspheres containing the oxides of Minor Actinides (MA) prepared through the internal gelation process are particularly relevant in this context. Since the sol-gel process used in the fabrication of these microspheres offers significant advantages over the conventional powder metallurgical processes used in the fabrication of pellets, efforts are underway at IGCAR in collaboration with Bhabha Atomic Research Centre (BARC)), Mumbai in order to establish a remote handling facility suitable for fabricating a sphere-pac fuel pin containing MOX microspheres. This paper describes the details of a facility that is being created at IGCAR to accomplish the above and the recent experience on the test fabrication of sphere-pac fuel pins for test irradiation. This facility comprises a jet-entrainment facility for the preparation of fine fraction UO2 microspheres (115±10 μm ), a column gelation facility for the preparation of the coarse fraction U, Pu MOX (53% Pu) microspheres ( 775±75μm ) through column gelation and equipment for characterization of the microspheres prepared in our laboratory. In addition a glove box train comprising facilites handling the microspheres, vibro-packing, fuel-pin welding and decontaminating the fuel pin was commissioned and qualified for handling MOX microspheres. As many as 75 production runs were carried out for the production of UO2 microspheres with the jet entrainment set-up, earlier, in order to optimize the process parameters viz., size of the nozzle, temperature of gelation, composition of the broth, washing routine as well as the conditions for calcination and sintering. Both the MOX as well as the UO2 microspheres were checked for their physical integrity, dimensions and were found to conform to the desired chemical composition stipulated for the test irradiation in FBTR. Conditions for vibrocompaction were optimized. Quality control checks were performed on the fuel pins after the fabrication. Trials were carried out for the preparation of Am containing UO2 fuel microspheres. The viscosity of a broth containing both U and Pu was measured in order to establish the optimum conditions for gelation of the sol. This is a “first of its kind” measurement, made in a custom-made facility created for this purpose.

Speaker: Mr Balija Sreenivasulu (IGCAR)

Material IAEA-CN-245-570-3.docx

208 Fast Reactors - The Belgian Regulatory Approach

In Belgium, the research center SCK•CEN is planning to build a lead alloy-cooled fast reactor as part of an ADS facility. And Whilst the Belgian Regulator, i.e. the Federal Agency for Nuclear Control - FANC, has experience with the licensing and control of PWRs, there was little to no experience with these less-common innovative reactor types. For this reason a prelicensing project was launched. The scope, the method and the goal of this prelicensing project is presented from the point of view of the regulator.

Speaker: Dr Matthias Vanderhaegen (Federal Agency for Nuclear Control)

Material Paper_FR17_v2.docx

209 Feasibility of MA Transmutation by (MA, Zr)Hx in Radial Blanket Region of Fast Reactor and Plan of Technology Development

This paper shows a feasibility study of transmuting minor-actinide (MA) by MA-zirconium hydride, (MA, Zr)Hx in radial blanket region of a fast reactor and a plan of technology development for the MA target. The feature of this concept is that it has a great potential of transmutation and can be used in proven fast reactors, but naturally requires research and development. The proposed (MA, Zr)Hx subassembly concept can be realized that the ratio of hydrogen to MA is enough for neutron energy spectrum shift and the loaded weight of MA is also enough for enhancing the transmutation because it densely contains both of MA and hydrogen. Preliminarily the MA transmutation rates were compared about four types of MA target: MA-Zr alloy pin; (MA, Zr)Hx one; lightly and heavily moderated combinations of (MA, Zr)O2 and ZrH1.6 ones. It was assumed that they are loaded around an active core in a 280 MWe sodium-cooled reactor; 54 MA target assemblies are respectively arranged in the radial blanket zone. It was followed that the MA transmutation of (MA, Zr)Hx doubles or triples, compared with the other types. One of the other issues is optimizing the irradiation condition and specification. Shorter terms from irradiation to acceptable decay heat for spent fuel storage, smaller power distortion of neighbor subassemblies, higher ratio of transmutation, and greater mass transmutation are preferable, but they are near incompatible. Therefore, the feasibility study is optimizing the irradiation condition and specification of the (MA, Zr)Hx target so as to harmonize the requirements. We started to research and develop key technologies of this concept toward an innovative actinide fuel cycle, conducted as the nuclear system research and development program under the contract with MEXT and supported by Nuclear Safety Research Association in Japan. The items of R&D contains measurements of the physical properties of (MA, Zr)Hx, fabrication testing, and laboratory-scale reprocessing test of (MA, Zr)Hx samples. Pellets of MA hydride target will be fabricated by hydrogenating of MA-Zr alloys in a Sieverts system to produce homogenous mixed pellet without a crack. In the next phase, sample of (MA, Zr)H pellet will be irradiated in a fast reactor and their irradiation behavior are measured by post-irradiation tests. These R&Ds would create a practicable and effective strategy of MA transmutation.

Speaker: Dr Kazumi IKEDA (Mitsubishi FBR Systems, Inc.)

Paper 564_MA_hydride_FR17_final.pdf

210 FEATURES OF THE NUCLEAR FUEL CYCLE SYSTEMS BASED ON JOINT OPERATION OF FAST AND THERMAL REACTORS

In a time of the existence of the national nuclear program in the framework of the weapons and civil complex it has been accumulated and continues to accumulate a significant amount of plutonium. The isotopic vector of plutonium produced in reactors varies strongly depending on the type of reactor, fuel burnup and the time elapsed since the moment of unloading it from the reactor before loading it as a fuel component in another reactor. It is fundamental fact that in a two component nuclear power system based on thermal and fast reactors, there is plutonium exchange between these types of reactors in a joint closed nuclear fuel cycle (NFC). Plutonium vectors coming into fast and thermal reactors can vary within a wide range because they will not only depend on the reactor features, but also on NFC management. The neutronic properties of plutonium isotopes differ also dramatically also. This leads to the fact that the physical characteristics (including safety features) of the reactor, in which the plutonium is used as fuel, will depend on the isotopic vector. The aim of the paper is to determine the characteristics of stationary fuel cycles of nuclear power system based on VVER-TOI and BN-1200 loaded with oxide fuel of various compositions. Characteristics of reactor systems with a partial or complete recycling of spent nuclear fuel and plutonium recycled are compared with those of the reference system which consist of the VVER-TOI reactors with uranium fuel, operating in an open NFC. The results of the computational researches of the transition of the two-component system of into the equilibrium mode in the closed NFC are presented. A feature of the system which is balanced by plutonium is that both types of reactors spent fuel is completely reprocessed and the separated plutonium is used totally to make MOX - fuel. The MOX fuel is used not only in the BN-1200, but also as a partial load in reactors VVER-TOI. The optimization of fuel the reactor fuel performances is needed for its effective cooperation. Complete closure by plutonium in the NFC consisting only of the VVER-TOI reactors using MOX - fuel is impossible.

Speaker: Dr Victor Dekusar (Dekusar)

Paper IAEA-CN-245-_480.docx

211 Full-fledged affination extractive-crystallizng platform for technology validation of the fast reactor spent fuel reprocessing on fast neutrons – the results of first experiments

Further effective development of fast energy engineering could not be realized without strategy of nuclear fuel cycle closure. Throughout the realization of this strategy combined PH-technology of mixed nitride uranium-plutonium fast reactor spent fuel is proposed. Full-fledged affination extractive-crystallizng platform was created for hydrometallurgical pro-cesses adjustment and above mentioned technology functional test. The platform guarantees the compliance with the radiation and nuclear safety requirements for working processes with hot spent fuel simulator, which main component is U-Pu-Np mixture, including Am, Тс, stable elements and radioactive tracers. Scientific and engineering solutions provide the conduction of spent fuel reprocessing technology adjustment research on low capacity, which appears to be the boundary between laboratory test equipment and industrial grade equipment, also provide the performance of equipment rerouting for different alternate layouts verification, the performance of computer-assisted control and operating procedure monitoring. The platform provides the research results verification along with mathematical model prediction results, the testing of different reagents, the performance procedure adjustment of analytical meas-urements; the platform can be also used as training complex. Platform equipment functional test operations on uranium simulator with stable fission products were accomplished, the results of first experiments were obtained. Keywords: affination platform, spent fuel reprocessing, spent fuel simulator

Speaker: Mr Sergei Terentev (JOIN STOCK COMPANY «SIBERIAN GROUP OF CHEMICAL ENTERPRISES»)

Paper Paper_376_corrected_Terentyev_ENG.pdf

212 High temperature design and evaluation of forced draft sodium-to-air heat exchanger in PGSFR

In PGSFR (prototype-gen IV sodium-cooled fast reactor), two kinds of DHRSs (decay heat removal systems) are employed for emergency decay heat removal during a loss of normal heat sink accident, which are ADHRS (active decay heat removal system) and PDHRS (the passive decay heat removal system). The ADHRS is a safety-grade active system, which is comprised of two independent loops with a single sodium-to-sodium decay heat exchanger (DHX) immersed in cold pool region and a single forced-draft sodium-to-air heat exchanger (FHX) located in upper region of the reactor building. The total heat removal capacity of the DHRS is 10 MWt which amounts to about 2.5% of the rated core thermal power. The DHRS is capable of cooling the plant from an initial temperature corresponding to any power operation condition to the safe shutdown condition within 72 hours after reactor shutdown with a single failure. The FHX employed in the ADHRS is a shell-and-tube type counter-current flow heat exchanger with M-shaped finned-tube arrangement. Liquid sodium flows inside the heat transfer tubes and atmospheric air flows over the finned tubes. During normal plant operation, small amount of heat loss through the FHX is permitted to prevent potential flow reversal or stagnation in each decay heat removal sodium loop. After the reactor shutdown, heat removal rate increases by opening dampers located in air flow paths of FHXs, and then the heat transferred to the decay heat removal system is finally dissipated into the atmosphere. In this study, high temperature design and creep-fatigue damage evaluation for a FHX were conducted. A creep-fatigue damage evaluation was performed according to the elevated temperature design codes of ASME B&PV Section III Division 5 based on a full 3D finite element analysis. The integrity of the heat exchangers under creep-fatigue loading was confirmed.

Speaker: Dr Nak Hyun Kim (Korea Atomic Energy Reasearch Institute)

Material FR17-High_temperature_design_and_evaluation_of_forced_draft_sodium-to-air_heat_exchanger_in_PGSFR.pdf

213 Impact of the irradiation of an ASTRID-type core during an ULOF with SIMMER-III

Innovative Sodium-cooled Fast Reactors (SFRs) are currently investigated in the ESNII+ European project. The goal of the WP6 “Core safety” of this project is to support the development of the ESNII roadmap as well as the implementation of the ESNII deployment strategy and licensing of the ESNII systems by identifying the experimental and theoretical R&D activities which are necessary for improving the present designs, as well as the existing methods, tools and databases for static and transient safety analysis of the ESNII critical reactor cores. One of the main issues of the WP6 “Core safety” of the ESNII+ project is to assess the behavior of the ESNII core (ASTRID-like core) in severe accidents at a representative stage, ie. the end of equivalent cycle (EOC), as the sodium voiding effect is less favorable at this moment. Consequently, the SIMMER-III code system (coupled thermohydraulics, pin mechanics and neutronics) is used as it can represent the accident up to an advanced core degradation. However, it has been developed to perform neutronics calculations at the beginning of life (BOL, without irradiation), and a new methodology needs to be implemented to perform neutronics calculations at the EOC. The aim of this paper is to present the difference of behaviors of the ESNII+ core at BOL and EOC, so as to highlight the importance of the irradiation in the accident scenario. Thus, a new methodology developed in the framework of the ESNII+ project to perform neutronics calculations at EOC is presented. Then, Unprotected Loss Of Flow (ULOF) calculations, with a 30s primary flow-rate halving time, are performed at BOL and EOC. The sodium boiling and the pin degradation happen earlier at EOC, but the core degradation is slow in both calculations and there are no power excursions. Despite less favorable feedback coefficients at EOC, and thanks to its heterogeneous geometry, the ESNII+ core in ULOF with a 30s halving time, does not lead, with the given hypotheses, to a power excursion.

Speaker: Dr Sandra POUMEROULY (EDF R&amp;D)

Material FR17_final.docx

214 IMPLEMENTATION STATUS OF CONTAIN-LMR SODIUM CHEMISTRY MODELS INTO MELCOR 2.1

This paper describes the progress of the CONTAIN-LMR sodium physics and chemistry models to be implemented into MELCOR 2.1. It also describes the progress to implement these models into CONTAIN2. In the past three years, the implementation included the addition of sodium equations of state and sodium properties from two different sources. The first source is based on the previous work done by Idaho National Laboratory by modifying MELCOR to include liquid lithium equation of state as a working fluid to model the nuclear fusion safety research. The second source uses properties generated for the SIMMER code. Testing and results from this implementation of sodium properties are given. Many of CONTAIN2’s physical models were developed since CONTAIN-LMR. Therefore, CONTAIN 2 is being updated with the sodium models in CONTAIN-LMR in order to facilitate verification of these models with the MELCOR code.

Speaker: Dr David Louie (Sandia National Laboratories)

Paper IAEA-CN-245-_367_v1a.docx

215 Improving inherent safety BN-800 by the use of fuel assembly with (U, Pu)C microfuel.

The task undertaken in the report is to increase inherent safety of the fast reactor with a sodium coolant of type BN-800 due to considering the possibility of using an innovation fuel assemblies with mixed uranium-plutonium carbide fuel in form of coated particles. Fuel assemblies with pellet MOX fuel and fuel rods are directly replaced by microspherical mixed (U,Pu)C-fuel. Calculation evaluations of characteristics of fuel assemblies with microspherical fuel are realized. A calculation comparison of neutron physics and thermal hydraulics characteristics of the innovation fuel assemblies with microspherical mixed (U,Pu)C-fuel and the traditional fuel assemblies with pellet MOX fuel and fuel rods was conducted. The chosen calculation model was BN-800 reactor core with MOX fuel, where a three-zone radial power density field flattening due to plutonium content change in fuel was used. Thanks to microspherical carbide fuel, inherent safety of the reactor increases in accidents with loss of coolant flow and introduction of positive reactivity because the coated particles have developed heat-exchange surface and their coats are able to keep fission products at higher temperatures than the steel cladding of traditional fuel rods.

Speaker: Mr Nikolay Maslov (National Research Centre "Kurchatov Institute" National Research Nuclear University MEPhI (Moscow Engineering Physics Institute))

Paper maslov_FR_17english1.pdf

216 INSERTION RELIABILITY STUDIES FOR THE RBC-TYPE CONTROL RODS IN ASTRID

This paper reports on preliminary studies performed regarding the insertion reliability of the RBC-type Control Rods designed for the ASTRID Sodium-cooled Fast Reactor. At this stage, the primary aim of the analysis is to evaluate the mechanical behaviour of RBC Control Rods under Emergency Shutdown conditions, for which reactor core structures are subjected to significant misalignments (including earthquake-imposed displacements). Using a Finite Element Model based on the Cast3M solver and developed specifically for these studies, computations are performed that allow assessing contact reactions (and the associated friction forces and contact pressures), deformations and stresses (mostly due to bending-induced deformations) which are considered for design. Based on these preliminary results, some optimisation of the Control Rod design is proposed that ensures some stable behaviour all along the rod drop, with substantial design margins.

Speaker: Mr Denis Lorenzo (CEA)

Paper FR17_Paper_IAEA-CN-245-_120.pdf

Poster FR17_Poster_IAEA-CN-245-120.pdf

217 Investigation of Radiation-Induced Swelling of EK-164 Steel, an Advanced Material for BN-600 and BN-800 Claddings

One of the main issues of fast reactors is to improve their economic efficiency. Nowadays austenitic stainless steel ChS-68 is used as a state material for BN-600 reactor claddings ensuring damage doses up to 87 dpa. Post irradiation examination results show the residual life of fuel elements with possible maximum damage dose up to 92-95 dpa corresponding to fuel burn-up to 12-13 % FIMA. At JSC "VNIINM" a prospective austenitic stainless steel EK-164 (16Cr-19Ni-2Mo-Ti-Si-V-P-B), more resistant to radiation-induced swelling than ChS-68 steel, has been developed to extend service life of fast reactor fuel assemblies. A trial operation of ChS-68 and EK-164 combined fuel assemblies has been carried out to ensure the operating capacity and assess the capability to improve operational characteristics of fuel elements with EK-164 claddings. At the initial stage maximum damage dose during operation of two test fuel assemblies is 74 and 77 dpa, respectively. Post irradiation examination confirms the advantage of EK-164 steel in terms of radiation-induced swelling resistance. Manufacturing technology of claddings used as a material for fuel elements operated to maximum damage dose in the range between 84 and 96 dpa has been improved considering structure and EK-164 cladding properties investigation results. Post irradiation examination shows that fuel elements retain their operating capacity and have sufficient residual life. Therefore it is possible to predict the operation resource at damage doses above 110 dpa. The paper aims to investigate radiation-induced swelling of EK-164 claddings at different temperatures and neutron irradiation doses, and to distinguish radiation-induced porosity characteristics at different irradiation temperatures.

Speaker: Irina Portnykh (Joint Stock Company "Institute of Nuclear Materials")

Material IAEA-CN-245-49.pdf

218 Investigation of steel corrosion products mass transfer in sodium

The report observes the behavior of the system sodium - oxygen - stainless steel with regard to the sodium cooled circulation loop. Computational and theoretical analysis of mass transport of corrosion products in the channels of non-isothermal circuit in view of chemical interaction of the components of steel with oxygen in sodium, including the reaction of sodium oxide with chromium in sodium is prepared. In the proposed model, we consider the reaction of sodium oxide with chromium in sodium in chromium- nickel steel circuit, taking into account the transfer of the reaction products in sodium and dynamics of sodium flow. The processes of impurities interaction with channel walls, formation and transport of suspended particles in the flow of coolant are considered. Closing relations include the equations describing the mass transfer of impurities between the coolant flow and the channel walls, the deposition of particles on the channel surface, the heat exchange between the coolant flow and channel walls. On the basis of the calculation and the theoretical analysis is refined information about physical and chemical constants, characterizing the mass transport of corrosion products in sodium at presence of increased content of impurities such as oxygen and hydrogen. Experimental study of mass transfer components of steel in sodium at low and high oxygen content in sodium is carried out. At low oxygen content the composition of the deposition is similar to that of steel dissolved. For the case of high oxygen concentration are performed two experiments: the oxygen content in sodium of 80 ppm and 140 ppm. The comparison of the calculation results with the experimental data on distribution of the chromium deposits in the cooling channels is completed, on which basis are defined updated values of the constants that characterize the mass transfer of chromium by dissolving stainless steel in sodium. It was found that an increase in the level of dissolved oxygen in sodium increases the solubility of chromium also.

Speaker: Dr Viktor Alekseev (JSC "SSC RF-IPPE")

Material FR-17_Alekseev_69.docx

219 Investigations in a substantiation of high-temperature nuclear energy technology with fast-neutron reactor cooled by sodium for manufacture of hydrogen and other innovative applications

As results of neutronics and thermal physical investigations of reactor installation BN-HT type with heat rating 600 MW have shown that there is a principal possibility to provide demanded parameters of a high-temperature fast reactor for production of a considerable quantity of hydrogen, for example, on the basis of one of thermochemical cycles or a high-temperature electrolysis with high factor of thermal use of the electric power. Safety requirements will be thus observed. The relative small sizes, the coolant type, the fissionable substance and structural materials allow to create a reactor with immanent to it properties (exclusion of reactor runaway by instantaneous neutrons, passive system of decay heat removal), providing the raised nuclear and radiation safety. By calculations BN-VT for production of electric power and hydrogen on basis of solid oxide electrolysis mass transfer hydrogen and tritium taking into account principal new method of clearing by pumping out through special membranes it is shown, that efficiency such system is ~40%, volume of maded hydrogen is 2.8104 l/s (under normal conditions). Danger from tritium in a finished stock originates after hydrogen combustion in an aerosphere. Therefore at calculation of parameters of the secondary circuit it was accepted, that maximum permissible tritium concentration in maded hydrogen should not exceed 3.26 Bk/l. Maximum concentration of permissible tritium in air is 2.44•103 Bk/l almost in 1000 times above. Clearing of sodium from tritium to the concentration providing in maded hydrogen maximum permissible concentration equal 3.26 Bk/l makes additional demands to system of clearing from hydrogen: the coefficient of permeability of system of clearing of the secondary circuit from tritium should exceed 140 kg/s. Taking into account high temperature experiments in which high efficiency of deduction of suspended matters of products of corrosion on the filters installed in низкотемпературной to a zone is shown, it is offered to use a principle of work of a cold trap: to chill sodium to necessary temperature with simultaneous deduction of products of corrosion on mass transfer surfaces, including filters. Working out of a necessary high temperature material and its studying under radiation demands the further investigations.

Speaker: Mr Aleksandr Sorokin (SSC RF-IPPE)

Paper Paper_436_(IPPE_Sorokin).pdf

220 Isothermal transformation austenite-ferrite in a P92 steel

The Time-Temperature-Transformation (TTT) diagram of an ASTM A335 P92 steel (9CrMoWVNNb) has been established starting from an austenitization temperature of 1050 °C. Isothermal transformation was carried out at temperatures from 625 up to 775 °C taking 25 °C intervals, using a high resolution dilatometer. Only two state fields (i.e., austenite and ferrite + carbides) were observed, in full agreement with previous results on similar steels. A subset of large austenite grains, with sizes significantly exceeding the mean, was observed in all of the tested samples. At temperatures below the nose of the TTT diagram, prior austenite grain boundaries were made visible by decorating them with carbides precipitated at the early stages of the transformation. Carbide decoration allowed to have an accurate picture of the size distribution of austenite grains under the prescribed conditions of thermal cycle. Above the nose, prior austenite grain boundaries are hardly seen due to a drastic change in carbide precipitation mechanisms. At the same time, the ferrite nucleation and growth is markedly different in these two temperature regions; there is a gradual transition between these two extreme behaviors. The dilatometric curves obtained at each temperature were fitted to the Kolmogorov-Johnson-Mehl-Avrami expression in order to extract kinetic information about the austenite-ferrite transformation. Fitting was accomplished so as to take into account the presence of the large austenite grains. At the same time, a thorough examination of the transformed samples was carried out by using optical and electron (FEG-SEM and TEM) microscopy. Carbon replicas were extracted from the surfaces of selected specimens and a detailed study of the carbides present in each case was added to the former information.

Speaker: Dr María Luppo (Argentina Atomic Energy Commission)

Material Paper_Number_443.pdf

221 Key features of design, manufacturing and implementation of laboratory and industrial equipment for Mixed Uranium – Plutonium Oxide (MOX) and Nitride fuel pellets fabrication in Russia

The presentation describes the author’s experience in design, manufacturing and installation of equipment used in Mixed Uranium – Plutonium fuel pellet fabrication in Russia. The key features of mixed uranium-plutonium oxide and nitride powders are described, as well as their influence on main process (furnaces, presses) and auxiliary (gloveboxes) equipment design. Technical solutions for working with low fluidity powders, automatic dimensional and weight control, automatic readjustment of the manufacturing parameters, automatic powder gathering are discussed. Conveyance of boats prone to deformation and gas separation systems, insulation material choice are described, as well as rules and regulations applicable for this kind of equipment.

Speakers: Mr Alexander Denisov (Deputy Director International Business Sosny R&amp;D Company), Mr Vincent Reynaud (CHAMPALLE SAS)

Material Paper_563.docx

222 Main outcomes from the JASMIN project: development of ASTEC-Na for severe accident simulation in Na cooled fast reactors

The JASMIN project was launched in the frame of the 7th Framework Programme (FP) of the European Commission (EC). It was inspired by the Gen-IV target of designing innovative reactors that intrinsically prevent severe accidents from occurring or drastically reduce their consequences. One of the main objectives of the 7th FP was the enhancement of the current capability of analysis of severe accidents in Na-cooled fast reactors notably by developing new simulation tools able to evaluate the consequences of unprotected accidents leading to fuel pin failure, fuel and cladding relocation, primary system loads, fission product and aerosols releases. To do so, the ASTEC platform originally developed for LWRs, was chosen to be adapted and extended to the environment of Na-cooled fast reactors, the result being called ASTEC-Na. The main advantage was to simulate all phenomena of interest that are today generally simulated by separate codes focusing on specific aspects (i.e., SAS-SFR, CATHARE, RELAP, CONTAIN-LMR, etc.) using only a single code. In fact, this integrated approach is not complex to be implemented due to the high modularity of ASTEC-Na which allows developing, validating and maintaining separately each of its modules that represent a macro-phenomenon. In addition, the flexibility in defining the core geometry, materials composition and reactor components makes ASTEC-Na able to study new SFR designs, e.g. with fertile layers in outer radial or inner axial core regions (such as in ASTRID design), subassemblies with an inner duct channel to induce fast fuel axial relocation (such as FAIDUS design), or new safety systems to shut-down the core power. The JASMIN project has addressed four main areas: thermal-hydraulics, pin thermal-mechanical behavior, source term and core neutronics. In each area, model development and assessment have been performed. In addition to the experimental test matrix built within the frame of the project and used as references for the model validation, the adequacy of ASTEC-Na models have been evaluated through the comparison with results of other suitable and referenced codes used for benchmarking purposes. The main outcomes from the assessment and validation work have been summarized in the form of a SWOT analysis (Strengths, Weaknesses, Opportunities and Threats) that clearly allows identifying the main needs for future model developments.

Speaker: Mrs NATHALIE GIRAULT (IRSN)

Material ID324_FR17_main_outcomes_from_the_JASMIN_project_cor.docx

223 Mathematical modeling of the mononitride nuclear fuel production processes

Implementation program of fast reactors in nuclear power engineering provides for the use of new types of nuclear fuel, in particular nitride. RFNC-VNIITF in cooperation with VNIINM develops models and codes for the simulation of physical and chemical processes for the purpose of software development and implementation of nitride mixed uranium-plutonium fuel fabrication technology. The main objective is to create software that allows you to select and optimize the process conditions. To date, developed and implemented the mathematical models of the basic processes of fuel fabrication: - grinding of powders; - carbothermal synthesis; - granulation; - pressing; - sintering tablets. Models are designed to calculate the basic characteristics of the products according to the characteristics of the starting materials and the process conditions. Also designed auxiliary thermodynamic program module that allows calculating the thermodynamic equilibrium of multicomponent and multiphase systems with different initial conditions, the thermodynamic functions of the individual reactions, allows you to work with the database of the thermodynamic properties of substances. In this work module is used for modeling systems, typical for carbothermal synthesis and sintering processes.

Speaker: Mr Igor Peshkichev (RFNC-VNIITF)

Paper Peshkichev_214.pdf

224 Mechanical and Thermal Properties of (U,Pu)O2-x

Designing nuclear fuels and simulating their irradiation behaviors in a reactor require modeling and formulation of a variety of fundamental properties. The property study of uranium-plutonium mixed oxide (MOX) as fast reactor fuels still requires further investigation because of the diverse parameters and the technical obstacles of plutonium operation. Young’s modulus of MOX pellets was evaluated by measuring the sound velocities of longitudinal and transverse waves in the pellets as functions of porosity, oxygen-to-metal ratio (O/M) and plutonium content. The effect of each was fitted to give a single equation, which is important in designing nuclear fuels and simulating their irradiation behaviors in a reactor. The results showed that porosity was the most important factor that 20% of the porosity decreased Young’s modulus by neatly 100GPa while O/M and plutonium content could change the Young’s modulus by ~20GPa. From the measured sound velocities, temperature dependence on Young’s modulus and specific heat capacity were calculated on the Debye model by leveraging the thermal expansion data. The temperature dependence that Young’s modulus decreases with increasing temperature is in good agreement with literature data. The specific heat capacity also agrees with that of calculated value by Kopp’s method, taken the Schottky term and the excited term into account. The relationship between mechanical and thermal properties was well described.

Speaker: Mr Shun Hirooka (Japan Atomic Energy Agency)

Paper FR17-paper_HIROOKA.PDF

225 Mechanical Design Evaluation of Fuel Assembly for PGSFR

The PGSFR core is a heterogeneous, uranium-10% zirconium (U-10Zr) metal alloy fuel design with 112 assemblies: 52 inner core fuel assemblies, 60 outer core fuel assemblies (FA), 6 primary control assemblies (CA), 3 secondary control assemblies, 90 reflector assemblies (RA), and 102 B4C shield assemblies (SA). The core is designed to produce 150 MWe with an average temperature rise of 155 ℃. The inlet temperature is 390 ℃ and the bulk outlet temperature is 545 ℃. The core height is 900 mm and the gas plenum length is 1,250 mm. The fuel assembly is composed of the several structural parts, which are the handling socket, upper/lower reflector, nose piece, hexagonal duct and fuel rods. The face to face dimension and the length are 132.36 mm and 4,550 mm, respectively. In this paper, there are two kinds of analyses for mechanical design and evaluation of FA. One is the dynamic behavior analysis of FA and the other is the structural analysis of FA components as design level. All of these analyses results will be verified by out-pile test of actual size test FA.

Speaker: Dr Kyungho Yoon (Korea Atomic Energy Research Institute)

Material FR17_khyoon_(17).pdf

226 Model validation of the ASTERIA-FBR code related to core expansion phase based on THINA experimental results

Mechanical consequences which might be caused by core disruptive accidents (CDAs) are one of the major concerns in FBR safety. Once a severe re-criticality occurs, core materials are vaporized creating a CDA bubble which consists of fuel vapor, steel vapor and fission gases. Rapid expansion of the CDA bubble drives a sodium slug of the upper plenum and threatens integrity of the shielding plug. Energy conversion from thermal energy to mechanical energy plays an important role for the boundary integrity during core expansion phase. This paper describes model validation study of ASTERIA-FBR* related to the thermal-to-mechanical energy conversion process, focusing on calculation models such as interfacial area model through the THINA** test simulation. As a result, it was found that the energy conversion process and its ratio were in good agreements with the experimental results. Mechanism of CDA bubble expansion and uncertainty brought by calculation models were also discussed. [References] *T.Ishizu, et.al., “Development of Integrated Core Disruptive Accident Analysis Code for FBR –ASTERIA-FBR,” Proceedings of ICAPP’12, Chicago, USA, June 24-28, 2012, Paper 12100. **F. Huber, et.al., “Experiments on the Behaviour of Thermite Melt Injected into Sodium –Final Report on the THINA Tests Results,” Proceedings of IAEA/IWGFR Technical Committee Meeting on Material-Coolant Interactions and Material Movement and Relocation in Liquid Metal Fast Reactors, 167-198, Oarai, Japan, 1994.

Speaker: Dr Tomoko Ishizu (Regulatory Standard and Research Department, Secretariat of Nuclear Regulation Authority (S/NRA/R))

Material IAEA-CN-245-6_(FR17_ishizu).pdf

227 Modeling of hydrodynamic processes at a large leak of water into sodium in the fast reactor coolant circuit

A description is given of a physico-mathematical model of the processes that occur in the sodium circuit with a variable flow cross-section in case of a water leak into sodium. The application area for this technique includes a possibility to analyze consequences of this leak as applied to sodium-water steam generators in fast neutron reactors. Hydrodynamic processes that occur in the sodium circuit in the event of a water leak are described within the framework of a 1-D thermally-nonequilibrium three-component gas-liquid flow model (sodium-hydrogen-sodium hydroxide). In this case hydrogen is assumed to be an ideal gas and its solubility in sodium is taken into account. Consideration is also given to dependence of sodium and sodium hydroxide on pressure and temperature. In the proposed improved approach the sodium circuit is presented in the form of combination of two models: • a 1-D model with distributed parameters, that describes dynamics of the parameters in all the circuit elements (sodium-water reaction region included), with the exception of expansion tank volume; • an expansion tank model built as part of the model with lumped parameters. These two models are cross-linked in the expansion tank inlet and outlet points. The proposed model and calculation technique have been realized in the form of a computer code. A computer code was tested on experimental data obtained from the injection of water vapor into sodium at the Russian sodium loop. Results gained from a comparison of calculations with experimental data, lead us to conclude that the proposed technique adequately reflects the transient behavior of the relevant parameters during the hydrodynamic processes that occur in sodium-water interaction in a sodium circuit.

Speaker: Ms Olga Myazdrikova (Institute for Physics and Power Engineering named after A.I. Leypunsky” (IPPE))

Material Myazdrikova_FR17.docx

228 Modeling of Lanthanide Transport in Metallic Fuels: Recent Progresses

C. Unal1, X. Li2, J. Isler2, S. Abid2, J. Zhang2, C. Matthews1, C. Arnold4, J. Galloway1, N. Carlson1, R. Mariani3 1 Los Alamos National Laboratory 2Ohio State University 3Idaho National Laboratory 4Pajarito Scientific Corporation fission products as well as lanthanide impurities in recycled feedstock are known to migrate to the periphery of metal fuels and initiate FCCI that weakens the cladding material. We will present here the most recent developments regarding efforts to implement reliable lanthanide transport models into the 3D fuel performance analysis code, BISON, and experimental and analytical efforts to determine the solubility of certain lanthanides in liquid metals that may be the mechanism for liquid-like transport. Using ab-initio MD, we found that the solubility of cerium in liquid sodium at 1000K was less than 0.78 at. %, and the diffusion coefficient of cerium in liquid sodium was calculated to be 5.57 10-5 cm2/s. We extended the MD work to two temperatures 723 K and 1000 K for Cerium (Ce), Praseodymium (Pr), and Neodymium (Nd) diffusion in Sodium (Na) and Cesium (Cs) three abundant Ln fission products diffusion coefficients in liquid Na at multiple temperatures. The Ln diffusivities are found to be in the magnitude order of liquid diffusion (10-5cm2/s) and the temperature dependence of diffusivity is developed according to Arrhenius equation. Experiments have been performed to measure the solubility of lanthanides in liquid sodium.. Using ICP-MS to measure the concentration of lanthanide in the sodium sample, the solubility at that testing temperature is calculated. The experimental results indicated that the solubility of cerium, praseodymium, and neodymium varied from 1x10-6 to 3x10-5 at. % between 723 to 823 K with considerable scatter. The time dependence of solubility is also obtained from experiments conducted at different equilibration times. To better describe the Ln transport behavior, a model of Ln transport through porous media is being developed from pore-scale to a continuum description through three main steps: 1) Ln dissolution at an isothermal fuel-pore interface, 2) Ln migration within a single 1D tubular pore along a temperature gradient, 3) Ln transport through porous media with specific porosity. Temperature effects are considered with inclusion of Soret term. Finally, an integrated model with regarding to an effective diffusivity, porosity and Soret effect is obtained. Initial results of modeling are discussed.

Speaker: Dr Cetin Unal (LANL)

Paper FR17_Lanthanide_Final.pdf

229 Modeling technologies of fuel cycles

There exist different variants of organizing the closure of nuclear fuel cycle (CNFC) depending on fast reactor type, fuel types, station or centralized allocation of closed nuclear fuel cycle stages. Many processes and engineering solutions used for implementation of chosen technologies for re-processing spent fuel are little-studied. One of the ways to verify and estimate engineering solution is mathematical modeling of radiochemical technology which in the end will allow to optimize composite technological process in order to increase effectiveness and reduce cost. The mathematical models for key processes of spent fuel reprocessing, fuel refabrication and radio-active waste managing are being developed in the frames of “Proryv” project for these purposes. Also codes VIZART and KOD TP to validate realizability and optimize parameters of CNFC pro-cessing lines are being developed. The codes use integrated library of technology models and allow to calculate material balance, create cyclograms, determine the most loaded parts of processing lines, estimate accumulation of fissile materials in devices and intermediate vessels, estimate the influence of control actions on technology process.

Speaker: Ms Inga Makeyeva (RFNC-VNIITF)

Paper Modellind_TP.docx

230 Multiscale computer modeling of nuclear fuel properties at radiation and thermal impacts

Description and prediction of behavior of nuclear engineering materials under operating conditions is one of the challenging goals in actual materials science. To solve this problem, when the restricted experimental information is only available (e.g. for the new kinds of nuclear fuel), the most perspective method seems to be theoretical description based on multiscale approach. In this case the various subtasks are jointly solved on different various time and spatial scales using theoretical physics and computer modeling. The cooperation of different techniques (such as quantum calculations, atomistic simulation, dislocation dynamics, phase field modeling, kinetic equations and continuum mechanics) allows predicting behavior of the nuclear materials in the absence of experimental data in the analyzed range of temperature, fission rate and other external conditions. In this work, we developed multiscale computer models for various types of nuclear fuel: UN, U-Mo, UO2. The work includes several stages of model development: development of a set of novel interatomic potentials; study of primary irradiation damage (collision cascades and radiation track); calculation of basic properties of matter (diffusion coefficients, dislocation and grain boundaries properties, phase transitions); mesoscale model for evolution of phase-structural composition and change of mechanical/thermodynamics properties under operating conditions.

Speaker: Dr Sergey Starikov (Nuclear Safety Institute of the Russian Academy of Sciences)

Paper FR17-234.pdf

231 Neutronic Self-sustainability of a Breed-and-Burn Fast Reactor Using Super-Simple Fuel Recycling

The breed-and-burn fast reactor (B&BR) is a unique concept of fast reactor, which can breed the fissile fuels and use the bred fuels in situ. Thanks to this characteristic, the fuel utilization in a B&BR can be extremely high and even the spent nuclear fuel of a B&BR can be re-used as a fuel to spawn the next generation B&BR after appropriate reprocessing or reconditioning. In this paper, a super-simplified melt and treatment (SSMT) process, which removes volatile elements only from the spent fuel, is suggested to enhance the proliferation-resistance and economy of the reprocessing. The neutronic feasibility of B&BR self-sustainability with SSMT is studied in terms of the burnup reactivity change, conversion ratio, core lifetime, power profiles and safety parameters. The fuel and core design was optimized to maximize the self-sustainability while preserving the inherent proliferation-resistance of the core.

Speaker: Mr Chihyung Kim (Korea Advanced Institute of Science and Techology)

Material FR17_B&BR_modified2.docx

232 New results on the continuous cooling behavior of an ASTM A335 P92 steel

This work introduces new results on the transformation behavior and microstructural evolution of ASTM A335 P92 steel under continuous cooling conditions (CCT). The first results were already reported and stored in the INIS database under the report number INIS-AR-C--1704. The material was austenized at 1050 ºC and afterwards cooled down at controlled rates (300, 200, 140, 120, 100, 90, 70, 50, 25 and 15 ºC/h). The transformation behavior of the steel samples was followed by dilatometry. The determination of the phases present in the samples after the thermal cycles was performed by optical and field emission scanning electron microscopy for the eleven tested values of cooling rate. Additionally, a full characterization was performed for selected samples by Mössbauer spectroscopy and X-ray diffraction. The phase domains identified according to the cooling rate were completely martensitic, completely ferritic and mixed martensitic-ferritic. Second-phase precipitation has been observed in all of the samples, and indications of the presence of retained austenite after some of the cooling cycles were also detected. The experimental results were collected in the form of a continuous cooling transformation diagram.

Speaker: Dr María Luppo (Argentina Atomic Energy Commission)

Material Paper_Number_445.pdf

233 Numerical and Experimental Investigations of Tube-to-Tube Interaction of Air Heat Exchangers of PFBR under Seismic Excitations

Numerical and experimental investigations of seismic response behavior of the air heat exchanger (AHX) of prototype fast breeder reactor (PFBR) were carried out for operating basis earthquake (OBE), safe shutdown earthquake (SSE) and beyond design basis earthquake conditions. For the numerical study, a finite element model consisting of AHX header and connecting tubes were developed using general purpose finite element code CAST3M and time history analyses were performed for all the three earthquake loading conditions. To perform the analyses, spectrum compatible time histories were generated from the floor response spectrums at the support location of the AHX. Studies predicted the possibility of tube-tube interaction between the middle and outer tubes due to the presence of circumferential fines provided along the tube length. To confirm the analyses findings, shake table experiments were performed using 100 t multi axial shake table. The test set up consists of five AHX tubes along with fines arranged in triangular pitch with tube to tube spacing same as the AHX in the reactor. The tubes were supported simulating the actual supporting conditions in the reactor. To simulate the fluid effects under dynamic conditions, tubes were filled with water and pressurized up to 7 bars. Prior to the seismic studies, free vibration characteristics of the tube bundle were estimated by performing resonance search tests and compared the results with numerical predictions. The responses were captured using accelerometers, strain gauges and no contact type displacement sensors. Tube responses are assessed for OBE, SSE and beyond SSE conditions by performing the tri axial excitations as per IEEE-344 guidelines using spectrum compatible time histories and responses are captured using a 96 channel data acquisition system. Tube to tube interactions at fin locations were observed under SSE and beyond SSE conditions as evidenced from the response spikes in accelerometers and strain gauge readings and from the relative displacements measured using non contact type displacement sensors. However, the structural integrity of tube bundle is demo