Dr
Štefan Čerba
(Slovak University of Technology in Bratislava)
28/06/2017, 17:50
Track 6. Test Reactors, Experiments and Modeling and Simulations
POSTER
Slovakia is involved in the development of the ALLEGRO reactor, the demonstrator of the unique GFR technology. Since the Gas-cooled Fast Reactor lacks any applicable experimental data, the design and optimization of its core must rely on data from similar reactor concepts and on calculations using Monte Carlo and deterministic methods. Although these two methods differ in their nature, both...
Mr
JEAN-MARIE HAMY
(AREVA NP)
28/06/2017, 17:50
Track 6. Test Reactors, Experiments and Modeling and Simulations
POSTER
The purpose of this paper is to describe in a first part a coupling methodology between two codes in order to describe global thermalhydraulic behavior inside a sodium-cooled fast reactor. A CFD code (STAR-CCM+) is used for the modelling of primary circuit, while a system code (CATHARE) is used for the modeling of specifics area in primary circuit (core structures and primary pumps) and the...
Mr
Bálint Batki
(Centre for Energy Research, Hungarian Academy of Sciences)
28/06/2017, 17:50
Track 6. Test Reactors, Experiments and Modeling and Simulations
POSTER
In this study, the fuel, the coolant, the cladding and the wrapper temperature reactivity coefficients were calculated with Serpent Monte Carlo code for the ALLEGRO demonstrational GFR core and for an SFR core with 3600 MWth power. The results were compared with each other and with thermal reactor reactivity coefficients, and it was found that the thermal expansion of the core structural...
Mr
Xiaoliang Chen
(China Institute of Atomic Energy)
28/06/2017, 17:50
Track 6. Test Reactors, Experiments and Modeling and Simulations
POSTER
China Experimental Fast Reactor (CEFR) has completed physics start-up tests in 2010 and connected the grid on FP in 2014. Characteristic of neutron field for irradiation in CEFR has been researched by calculation and experiments. In future, CEFR will been operated as an irradiation test facility for fuel, material and other application, and some irradiation projects, such as irradiation of...
Dr
Willem Frederik Geert Van Rooijen
(Research Institute of Nuclear Engineering, University of Fukui)
28/06/2017, 17:50
Track 6. Test Reactors, Experiments and Modeling and Simulations
POSTER
A Coordinated Research Project (CRP), initiated by the International Atomic Energy Agency (IAEA), consisted of benchmark calculations with the goal to analyze two of the Shutdown Heat Removal Tests performed in EBR-II, namely SHRT-17 and SHRT-45R. Test SHRT-45R concerned an Unprotected Loss Of Flow (UOLF) scenario. In this case, only the inherent feedback mechanisms due to the change of the...
Mrs
Irina Chernova
(Nuclear Safety Institute of Russian Academy of Sciences)
28/06/2017, 17:50
Track 6. Test Reactors, Experiments and Modeling and Simulations
POSTER
A module of transient neutron transport problem solution is an essential part of complex codes designed for the analysis of nuclear safety in different modes of operation of a nuclear reactor. The neutron transport problem can be solved with a variety of approximate schemes, such as point kinetics, adiabatic, quasi-static and improved quasi-static approximations or direct numerical solution of...
Mr
Alexey Tutukin
(Vladimirovich)
28/06/2017, 17:50
Track 6. Test Reactors, Experiments and Modeling and Simulations
POSTER
Information has been obtained on the flow distribution along parallel paths, on the
situation of the free levels, on the heat and mass transfer processes. Experimental
determination of the hydraulic parameters is only possible for the components of the core by
using full-scale mock-ups. The remaining elements of the circulation loop require calculation
justification supported by...
Dr
Tai Asayama
(Japan Atomic Energy Agency)
28/06/2017, 17:50
Track 2. Fast Reactor Operation and Decommissioning
POSTER
Grade 91 steel is a Code-approved construction material in the ASME and JSME nuclear codes. Applications of Grade 91 steel include intermediate heat exchanger, piping, steam generator tubing and shell, etc. for sodium fast reactor systems. Current creep-fatigue damage evaluation method in the ASME and JSME nuclear code differs in the method to calculate creep damage. In the simplified...
Dr
Hidemasa Yamano
(Japan Atomic Energy Agency)
28/06/2017, 17:50
Track 6. Test Reactors, Experiments and Modeling and Simulations
POSTER
This paper describes basic visualization experiments on eutectic reaction and relocation of boron carbide (B4C) and stainless steel (SS) under a high temperature condition exceeding 1500°C as well as the importance of such behaviors in molten core during a core disruptive accident in a Generation-IV sodium-cooled fast reactor (750MWe class) designed in Japan. At first, a reactivity history was...
Dr
Chiwoong CHOI
(Korea Atomic Energy Research Institute (KAERI))
28/06/2017, 17:50
Track 6. Test Reactors, Experiments and Modeling and Simulations
POSTER
KAERI has joined the International Atomic Energy Agency (IAEA) coordinated research project (CRP) on Benchmark Analysis of an EBR-II Shutdown Heat Removal Test (SHRT). The major goal for this program is to validate MARS-LMR, which is a newly developed safety analysis code for PGSFR. One of benchmark tests is a SHRT-45R, which is an unprotected loss of flow test in an EBR-II. Thus, sodium...
Mr
Moustafa Aziz Ibrahim
(Nuclear and Radiological Regulatory authority)
28/06/2017, 17:50
Track 6. Test Reactors, Experiments and Modeling and Simulations
POSTER
The BN-600 reactor is a sodium-cooled fast breeder reactor, built at Russia. Designed to generate electrical power of 600 MW in total. IAEA has considered the reactor for many phases of benchmark problems.The coordinated research project activities were started in 1999 and included studies fora so-called hybrid BN-600-reactor-type core model, partially fuelled with highly enriched uranium and...
Mr
Yury Naboyshchikov
(JSC "State Scientific Center - Research Institute of Atomic Reactors")
28/06/2017, 17:50
Track 6. Test Reactors, Experiments and Modeling and Simulations
POSTER
At present, a wide range of tests is performed in the BOR-60 reactor in support of reactors under operation, construction and design in Russia and worldwide. Most of the tests are performed in the reactor core regions of the peak dose accumulation rates. However, there is a high demand for irradiation testing to be performed in the BOR-60 blanket.
An important feature of any nuclear...
Mr
Artem Varivtcev
(JSC “SSC RIAR”)
28/06/2017, 17:50
Track 6. Test Reactors, Experiments and Modeling and Simulations
POSTER
At present, different nuclear fuels (NF) to be used in advanced fast neutron reactors (AFR) are tested in the BOR-60 reactor. In such in-pile testing the top priority is to ensure the maximum possible compliance of the target NF irradiation parameters with the design operating parameters. The key monitored parameters in testing experimental fuel elements are the fuel burnup rate and linear...
Mr
Jae-Ho Jeong
(KAERI)
28/06/2017, 17:50
Track 6. Test Reactors, Experiments and Modeling and Simulations
POSTER
The wire effect on three-dimensional flow field and heat transfer characteristics in a helically wrapped fuel assembly mock-up of an SFR (Sodium-cooled Fast Reactor) have been investigated through a numerical analysis using the commercial CFD (Computational Fluid Dynamics) code, CFX. The SFR system has a tight package of the fuel bundle and a high power density. The sodium material has a high...
Mr
Yonghoon Shin
(Seoul National University)
28/06/2017, 17:50
Track 6. Test Reactors, Experiments and Modeling and Simulations
POSTER
We present results of experiments with lead-bismuth eutectic (LBE) non-isothermal natural circulation in a full-height scale test loop, HELIOS, and numerical modeling results performed by a system thermal-hydraulics code. The experimental studies were conducted under steady state as a function of core power conditions from 9.8kW to 33.6kW. Local surface heaters on the main loop were activated...
Prof.
Nikolay Maksimov
(National research nuclear university (Moscow Engineering Physics Institute))
28/06/2017, 17:50
Track 8. Professional Development and Knowledge Management
POSTER
The results of the analysis of terminological subsets of the characteristic lexicon, that is used for semantic identification of scientific and technical information in the field of fast neutron reactors is discussed. Describes the procedure of automatic formation of ordered dictionaries reference lexicon based on the full text processing of documents – scientific reports, dissertations,...
Mr
Zhiwei Zhou
(Engineer)
28/06/2017, 17:50
Track 6. Test Reactors, Experiments and Modeling and Simulations
POSTER
This study developed a thermal hydraulic design code based on powerful capabilities of computer hardware and advanced numerical simulation technology for the optimization design of reactor core assemblies, big and small grid plates and the primary circuit fluid network, for the use of core assemblies thermal hydraulic design, special design of flow distribution in the big and small grid...
Ms
Xiuli Xue
(Senior engineer)
28/06/2017, 17:50
Track 6. Test Reactors, Experiments and Modeling and Simulations
POSTER
For the lacking of applicable code to analyze emergency heat removal capacity of pool-style fast reactors, it is developed according to the design requirement of demonstration fast reactors. The code builds a consolidated platform developing modules of reactor core, sodium pool, relevant components and so on individually from one-dimension to three-dimension, forming program module packages...
Mr
Victor Blandinskiy
(NRC "Kurchatov Institute")
28/06/2017, 17:50
Track 6. Test Reactors, Experiments and Modeling and Simulations
POSTER
Multifunctional fast neutron research reactor MBIR is intended to provide the basis for broad scope of research and experimental activities due to achieving 5E15 n/cm2/s integral neutron flux in central loop channel, i.e. MBIR is considered as neutron source of high intensity.
This work focuses on the possibilities of the use of different fuel in MBIR to provide required neutron flux in loop...
Dr
Velusamy K
(IGCAR)
28/06/2017, 17:50
Track 2. Fast Reactor Operation and Decommissioning
POSTER
During start-up of the reactor, the rate at which the power raising is to be done is a critical issue. As a safe practice, the power raising can be done slowly by increasing the sodium temperature at the rate of 5K/hr, 10K/hr or 20K/hr. But this takes a long time for start-up, i.e., 80, 40 and 20 hours per cycle respectively to raise the sodium temperature from the initial temperature of 453 K...
Mr
David Wootan
(Pacific Northwest National Laboratory)
28/06/2017, 17:50
Track 8. Professional Development and Knowledge Management
POSTER
The FFTF was designed to accommodate up to four CLS. Each CLS was an irradiation testing system capable of operating at 2.3 MWt with its own independently controlled coolant system. An irradiation test in a CLS was inserted into the core within a Closed Loop In-Reactor Assembly (CLIRA). An entire CLS consisted of a CLIRA, a primary and a secondary cooling loop, and a Dump Heat Exchanger...
Mr
Puthiyavinayagam Pillai
(Indira Gandhi Centre for Atomic Research)
28/06/2017, 17:50
Track 2. Fast Reactor Operation and Decommissioning
POSTER
In the case of Indian Prototype Fast Breeder Reactor (PFBR), which is in an advanced stage of commissioning, the fuel pins along with other parts of fuel subassembly are stored in an Interim Fuel Storage Building (IFSB). The final assembling of subassembly is carried out in IFSB and the IFSB is located far away from PFBR site. It is essential to demonstrate safe transportation of the fuel...
Dr
Romain Coulon
(CEA LIST)
28/06/2017, 17:50
Track 2. Fast Reactor Operation and Decommissioning
POSTER
The cleanliness of the primary circuit is a safety requirement for Generation IV nuclear power plants. During operation, fission product concentration into the sodium coolant has to be monitored by dedicated radiation monitoring systems. These systems are based on neutron counting and gamma spectroscopy. Neutron detection allows detecting clad failure when its opening is large enough to...
Prof.
liqin hu
(Institute of Nuclear Energy Safety Technology, CAS · FDS Team)
28/06/2017, 17:50
Track 6. Test Reactors, Experiments and Modeling and Simulations
POSTER
Abstract:
Compared with the other reactor types, the neutron spectrum of fast reactors is hard, affecting the neutronics and safety performance, for which advanced nuclear simulation methods such as Monte Carlo method is necessary for the nuclear design. Super Monte Carlo Simulation Program for Nuclear and Radiation Process (SuperMC) is a general, intelligent, accurate and precise program...
Mr
Partha Sarathy UPPALA
(IGCAR, INDIA)
28/06/2017, 17:50
Track 6. Test Reactors, Experiments and Modeling and Simulations
POSTER
Experimental Breeder Reactor (EBR-II) was a U-Pu-Zr metal-alloy fueled liquid-metal-cooled fast reactor, extensively used for conducting safety experiments. Out of several tests conducted, the SHRT-17 loss of flow test conducted in 1984 demonstrated the decay heat removal capability by natural circulation in sodium cooled fast reactor with no core damage. In order to utilize the data recorded...
Mr
Nikolay Loginov
(SSC RF-IPPE)
28/06/2017, 17:50
Track 6. Test Reactors, Experiments and Modeling and Simulations
POSTER
Development experience for experimental reactor facility cooled with evaporating liquid metals
Loginov N.I.
State Scientific Center of the Russian Federation – Institute of Physics and Power Engineering. Obninsk, Russia
SSC RF-IPPE developed technical project of experimental reactor facility cooled with evaporating liquid metals at 1988-97. Evaporating sodium used as a coolant of the core...
Ms
Marianne GIRARD
(CEA/FRANCE)
28/06/2017, 17:50
Track 2. Fast Reactor Operation and Decommissioning
POSTER
Within the ASTRID reactor project, CEA, EDF and AREVA, have launched a R&D program focused on the low leak rates detection of sodium on pipes. This program is focused on the development of innovating detectors, multilayer-type and Optic Fiber, involving tests in the FUTUNA-2 sodium loop. This loop is designed to produce very accurate sodium leak rates within a range around 1 cc/min, the tests...
Mr
Alexander Tuzov
(JSC "SSC RIAR")
28/06/2017, 17:50
Track 6. Test Reactors, Experiments and Modeling and Simulations
POSTER
This paper presents the technical parameters and the experimental capabilities of the MBIR reactor; it shows the main reactor engineering solutions of the project and describes the key events related to the implementation of the MBIR construction project.
Mr
Tanju Sofu
(Argonne National Laboratory)
28/06/2017, 17:50
Track 8. Professional Development and Knowledge Management
POSTER
Inherent and passive safety is a key aspect to achieve licensing assurance and reduce plant costs, but it requires demonstration and validation of key features. To address this need, DOE-NE's Advanced Reactor Technologies (ART) program supported development of EBR-II, FFTF, and TREAT safety testing databases, metal fuel irradiation database, EBR-II physics analysis database and materials...
Dr
Tyler Sumner
(Argonne National Laboratory)
28/06/2017, 17:50
Track 6. Test Reactors, Experiments and Modeling and Simulations
POSTER
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 performed...
Mr
Francesco Lodi
(University of Bologna)
28/06/2017, 17:50
Track 6. Test Reactors, Experiments and Modeling and Simulations
POSTER
One of the crucial objectives for the Advanced Lead-cooled Fast Reactor European Demonstrator (ALFRED) is proving the viability of the general concept adopted in the design. This proof passes through the successful operation of ALFRED, demonstrating that the design assumptions provide not only the foreseen performances, but also the aimed reliability. The demonstration of the reliability can...
Dr
Nicolas Stauff
(Argonne National Laboratory)
28/06/2017, 17:50
Track 6. Test Reactors, Experiments and Modeling and Simulations
POSTER
One of the tasks of the OECD/NEA sub-group on Uncertainty Analysis in Modelling (UAM) of Sodium-cooled Fast Reactors (SFR-UAM) under the NSC/WPRS/EGUAM is to perform a code-to-code comparison on neutronic feedback coefficients and associated uncertainties calculated for transient analyses. This benchmark exercise benefits from the results of a previous Sodium Fast Reactor core Feed-back and...
Mr
Vladimir Danilenko
(FSUE «Alexandrov Research Institute of Technology»)
28/06/2017, 17:50
Track 2. Fast Reactor Operation and Decommissioning
POSTER
Diagnostic system of the core of the reactor facility BN-800 (SDRU) of Unit 4 of Beloyarsk NPP is an automated system as part of automatic process control system (APCS), designed for complex control of processes taking place in the reactor in normal operating conditions and violations of normal operation, detection at an early stage of violations of normal operation and damage to the reactor...
Mr
Rajan Babu Vinayagamoorthy
(Director (Technical))
28/06/2017, 17:50
Track 2. Fast Reactor Operation and Decommissioning
POSTER
Abstract
Electric power supply comprising of both OFF site and ON site power supply systems is designed to facilitate functioning of equipment important to safety during normal operation, anticipated operational occurrences and accident conditions. Reliability of both OFF site and ON site electrical power supply systems is of paramount importance as it feeds the power for Reactor...
Mr
Huanjun Zhu
(CEFR)
28/06/2017, 17:50
Track 6. Test Reactors, Experiments and Modeling and Simulations
POSTER
China experimental fast reactor (CEFR) uses uranium dioxide as its first fuel, then it switchs to the MOX fuel. The MOX fuel have been great changes in thermal hydraulic characteristics, therefore it needs to study. In this paper, an resistance characteristic expriment and the CFD simulation of the CEFR-MOX assembly was carried on. Through comparison and analysis, an empirical formula was...
Mr
Vladimir Eliseev
(SSC RF-IPPE)
28/06/2017, 17:50
Track 6. Test Reactors, Experiments and Modeling and Simulations
POSTER
Cores of research reactor facilities (RRF) as opposed to those of power reactors are designed taking into account their research function. This is their unique peculiarity reflected, among other features, in the flexibility (i.e. transformation in safe and reasonable limits) of the core arrangement according to the changing goals of specific experiments. Power generation for research reactor...
Mr
Konstantin Mitrofanov
(JSC “SSC RF – IPPE”)
28/06/2017, 17:50
Track 6. Test Reactors, Experiments and Modeling and Simulations
POSTER
In the present work the set-up created on the basis of the accelerator Tandetron (IPPE) for the experimental studies of the time dependence of delayed neutron activity from neutron induced fission of 235U is described. Measurements were carried out with neutron beam generated by the 7Li(p,n) reaction. The lower limit of the investigated time range was governed by the proton beam switching...
Dr
HUEE-YOUL YE
(Korea Atomic Energy Research Institute)
28/06/2017, 17:50
Track 6. Test Reactors, Experiments and Modeling and Simulations
POSTER
The Prototype Generation IV Sodium-Cooled Fast Reactor (PGSFR) has been developed by Korea Atomic Energy Research Institute (KAERI). The hydraulic part such as the impellar and diffuser of the PHTS pump has been designed to satisfy the requirement of the hydraulic performance. The essential geometric parameters of the impellar and diffuser were determined through the optimal design...
Dr
Giacomo Grasso
(Italian National Agency for New Technology, Energy and Sustainable Economic Development (ENEA))
28/06/2017, 17:50
Track 6. Test Reactors, Experiments and Modeling and Simulations
POSTER
The advancement of the design of ALFRED – the Advanced Lead-cooled Fast Reactor European Demonstrator – beyond the conceptual phase, passes through the analysis of the impact of uncertainties, notably to what concerns safety-related conditions.
Focusing on the design of the core, nuclear data are the main source of uncertainties, so that their evaluation is of utmost importance in order to...
Mr
Francois Baque
(CEA)
28/06/2017, 17:50
Track 2. Fast Reactor Operation and Decommissioning
POSTER
CEA initiated a study in 2008 for improving design rules of Fast Reactors, with French utilities (EDF), French Designers (AREVA) and Non Destructive Examination (NDE) specialists (Aix Marseille University), and dealing with the specific aspect of in-service inspection (ISI).
Thus, at the end of 2012, RCC-MRx specifications for NDE code could be enlarged, extending those performed at...
Mr
Vasiliy Pakholkov
(JSC “Afrikantov OKBM”)
28/06/2017, 17:50
Track 6. Test Reactors, Experiments and Modeling and Simulations
POSTER
Decay heat removal from BN-1200 to atmospheric air is performed using passive DHRS. Several innovative solutions are applied during the DHRS development:
- hydraulic connection of the decay heat exchanger (DHX) with reactor high pressure plenum, installation of check valve at DHX outlet;
- application of slide valve of air heat exchanger with passive elements;
- natural circulation in all...
Dr
Jong Hyuck Won
(KAERI)
28/06/2017, 17:50
Track 6. Test Reactors, Experiments and Modeling and Simulations
POSTER
Korea Atomic Energy Research Institute (KAERI) has been developing an SFR to aim at specific design approval of a Prototype Gen-IV Sodium-cooled Fast Reactor (PGSFR). In the PGSFR, a metal fueled, blanket-free, pool type SFR concept is adopted to acquire the inherent safety characteristics and high proliferation-resistance.
The metal-fueled SFR such as the PGSFR is known to be inherently...
Mr
Konstantin Sergeenko
(Michailovich)
28/06/2017, 17:50
Track 6. Test Reactors, Experiments and Modeling and Simulations
POSTER
Study of turbulent pipe flows is extremely important because of its wide range of
applications. In the past decades, many fundamental theoretical and experimental studies on
wall-bounded flows have been performed: in the pipe, flat channel and boundary layer flow
geometries. However, the internal fluid dynamics in these regions still far from being
understood. Numerical simulation offers...
Mr
David Wootan
(Pacific Northwest National Laboratory)
28/06/2017, 17:50
Track 8. Professional Development and Knowledge Management
POSTER
The Fast Flux Test Facility (FFTF) is the most recent liquid-metal reactor (LMR) to operate in the United States, having operated from 1982 to 1992 and played a key role in LMFBR development and testing activities. In addition to irradiation testing capabilities, FFTF provided long-term testing and evaluation of plant components and systems for LMFBRs. The FFTF was highly successful and...
Dr
Dmitry Fomichev
(JSC NIKIET)
28/06/2017, 17:50
Track 6. Test Reactors, Experiments and Modeling and Simulations
POSTER
Liquid metals (LM), such as sodium, lead or lead-bismuth eutectic (LBE), are preferred candidate coolants for advanced fast nuclear reactors next generation. Despite the comprehensive amount of experimental and calculated data, obtained by Russian as well as EU scientists in previous 30-40 years, the investigation of hydraulic and heat transfer characteristics of the fuel pin bundles is one of...
Dr
Sandro Pelloni
(Paul Scherrer Institute)
28/06/2017, 17:50
Track 6. Test Reactors, Experiments and Modeling and Simulations
POSTER
Safety robustness by all means is still an open issue for Generation IV Sodium-cooled Fast Reactors (SFR) that needs to be demonstrated in particular with respect to severe accidents. For reliable safety analyzes it is important among other things not only determining 3D maps of reactivity coefficients which can then be used in corresponding transient analyzes within a point kinetics code; in...
Ms
Ksenia Kalugina
(JOINT-STOCK COMPANY «N.A. DOLLEZHAL RESEARCH AND DEVELOPMENT INSTITUTE OF POWER ENGINEERING»), Mrs
Varvara Yufereva
(JSC «N.A. Dollezhal Research and Developmant Institute of Power Engineering»)
28/06/2017, 17:50
Track 6. Test Reactors, Experiments and Modeling and Simulations
POSTER
Criticality calculations for BFS-1 test facility with lead were performed using Monte-Carlo code MCU-BR to verify some evaluated neutron data files for fast spectra. These data files are RUSFOND, ENDF/B-VII.1, JEFF-3.2, JENDL-4.0, CENDL-3.1 and some combined data. The continuous energy treatment (ACE format) was used. Critical assemblies include the pellets consisted from fissionable...
Dr
Emil Fridman
(Helmholtz-Zentrum Dresden-Rossendorf)
28/06/2017, 17:50
Track 6. Test Reactors, Experiments and Modeling and Simulations
POSTER
The nodal diffusion code DYN3D is under extension for Sodium cooled Fast Reactor (SFR) applications. As a part of the extension a new model for axial thermal expansion of fuel rods was developed. The model provides a flexible way of handling the axial fuel rod expansion that is each sub-assembly and node can be treated independently. In the current paper the new model will be described in...
Mr
boris abramov
(Joint Stock Company "State Scientific Centre of the Russian Federation – Institute for Physics and Power Engineering named after A. I. Leypunsky")
28/06/2017, 17:50
Track 6. Test Reactors, Experiments and Modeling and Simulations
POSTER
More precise definitions of the perturbation theory formulas
for reactivity effects calculations
B.D. Abramov
Leading Scientist of IPPE, Obninsk, Russia. E-mail: abramov@ippe.ru
Abstract
The paper is devoted to the issues of development and modification of methods of evaluation of reactivity effects in nuclear...
Mr
Valery Bereznev
(IBRAE)
28/06/2017, 17:50
Track 6. Test Reactors, Experiments and Modeling and Simulations
POSTER
CORNER and ODETTA codes for neutrons and photons transport based on discrete ordinates using finite differences and finite elements methods have been developed as a part of the new generation codes for the construction and validation of the perspective FBR safety.
Modern CONSYST software is used for the preparation of the macroscopic cross sections. Both eigenvalue (keff) and fixed source...
Mr
Chirayu Batra
(IAEA)
28/06/2017, 17:50
Track 6. Test Reactors, Experiments and Modeling and Simulations
POSTER
Nuclear Reactor Modeling and Simulation Toolkit (NuReMoST) is a new generation system for numerical simulation of the nuclear reactor. The system integrates typical simulation tasks, such as preparation of the design and geometry data, setup of the initial and boundary conditions, meshing tools, visualization functions, as well as several other related utilities and services. The core point of...
norihiro doda
(Japan Atomic Energy Agency)
28/06/2017, 17:50
Track 6. Test Reactors, Experiments and Modeling and Simulations
POSTER
After reactor shutdown in sodium-cooled fast reactors, natural circulation in the heat transport systems can be expected to remove the core decay heat in the case of station blackout. For reactor safety, the core hot spot temperature during decay heat removal by natural circulation should be evaluated. In order to evaluate the core hot spot temperature, Japan Atomic Energy Agency is developing...
Mr
Evgeny Rozhikhin
(IPPE)
28/06/2017, 17:50
Track 6. Test Reactors, Experiments and Modeling and Simulations
POSTER
Specialists involved in the process of validation and verification of codes and cross sections for the physics of fast reactors traditionally used the benchmarks presented in the “Cross Section Evaluation Working Group Benchmark Specifications” BNL-19302 (ENDF-202) handbook first issued in 1974 and last updated in 1991. This handbook presents simplified homogeneous models of experiments with...
Dr
Stefano Monti
(IAEA)
28/06/2017, 17:50
Track 8. Professional Development and Knowledge Management
POSTER
The Technical Working Group on Fast Reactors (TWG-FR) of the International Atomic Energy Agency (IAEA) was established in 1967 and since then it has been a foundation of the agency’s activities in the field of fast reactor. For last five decades the group of experts under the umbrella of TWG-FR have provided advice and supported the implementation of the programme. The TWG-FR assists in...
Dr
Federico Puente Espel
(Instituto Nacional de Investigaciones Nucleares)
28/06/2017, 17:50
Track 6. Test Reactors, Experiments and Modeling and Simulations
POSTER
In 2014, Mexico was honored with its acceptance as an observer member in the Technical Working Group on Fast Reactors of the IAEA. Afterwards, the Mexican participation in the fast reactors activities augmented when in 2016 the Mexican Team was accepted in the UAM SFR Benchmark of the OECD/NEA. The first technical specifications of the mentioned Benchmark consisted of four sodium-cooled fast...
Mr
David Wootan
(Pacific Northwest National Laboratory)
28/06/2017, 17:50
Track 8. Professional Development and Knowledge Management
POSTER
Significant cost and safety improvements can be realized in advanced liquid metal reactor designs by emphasizing inherent or passive safety through crediting the beneficial reactivity feedbacks associated with core and structural movement. This passive safety approach was adopted for the Fast Flux Test Facility (FFTF), and an experimental program was conducted to characterize the structural...
Mr
Balija Sreenivasulu
(IGCAR)
28/06/2017, 17:50
Track 2. Fast Reactor Operation and Decommissioning
POSTER
Large volume of liquid sodium is being handled in primary and secondary coolant circuits in Fast Breeder Reactors (FBRs). In the steam generator section, sodium is separated from high pressure steam/water by a thin wall of ferritic steel. In the event of any sudden leak, high pressure steam/water comes in contact with liquid sodium resulting into sodium-water reaction. Such an eventuality...
Mr
Igor Zhemkov
(JSC “SSC RIAR”)
28/06/2017, 17:50
Track 6. Test Reactors, Experiments and Modeling and Simulations
POSTER
The fast research reactor BOR-60 is one of the world’s leading research reactors in large-scale testing of fuel elements, FAs and control rods of different design options, advanced fuel compositions and structural materials, as well as in tryout of the closed fuel cycle technologies and transmutation of minor actinides. BOR-60 is a unique experimental reactor with a neutron spectrum ranging...
Mr
Qi Zhou
(China Institute of Atomic Energy)
28/06/2017, 17:50
Track 6. Test Reactors, Experiments and Modeling and Simulations
POSTER
Research on MOX fuel has lasted for a long time in China Institute of Atomic Energy (CIAE). Right now the focused topics are the manufacture and irradiation performance test for the China Experimental Fast Reactor (CEFR) MOX assembly, large batch production for CEFR MOX assembly and the CEFR core transition from Uranium fuel to MOX fuel. In order to determine the uncertainty of the CEFR MOX...
Prof.
Zhigang Zhang
(Harbin Engineering University)
28/06/2017, 17:50
Track 6. Test Reactors, Experiments and Modeling and Simulations
POSTER
Based on the structure and physical characteristics of primary coolant system (PCS) of China Experimental Fast Reactor (CEFR), a series of reasonable mathematical and physical models were set up. A set of stable and highly effective numerical methods were used to solve the models. Then the real-time thermal-hydraulic analysis codes for PCS of CEFR have been developed with modular method by...
Mr
Svyatoslav Sikorin
(Nikolaevich)
28/06/2017, 17:50
Track 6. Test Reactors, Experiments and Modeling and Simulations
POSTER
Benchmark criticality experiments on small-sized fast multiplying systems with HEU fuel were performed using “Giacint” critical facility of the Joint Institute for Power and Nuclear Research – Sosny of the National Academy of Sciences of Belarus. The critical assemblies’ cores comprised fuel assemblies, each of which consisted from 19 fuel rods of two types and had no the clad. The first one...
Mr
Igor Petrov
(JSC “Afrikantov OKBM”)
28/06/2017, 17:50
Track 2. Fast Reactor Operation and Decommissioning
POSTER
PRC – Russia cooperation in CEFR RP construction commenced in 1992.
In 1992 – 1995, specialists of RF from “Experimental Design Bureau of Mechanical Engineering” developed “CEFR unit concept” and technical requirements for the reactor and its main components.
In 1995 – 1996, Russian companies’ specialists developed detailed design of CEFR nuclear power plant.
Designing solutions implemented...
Dr
Jewhan Lee
(KAERI)
28/06/2017, 17:50
Track 6. Test Reactors, Experiments and Modeling and Simulations
POSTER
To support the development of Prototype Gen IV Sodium-cooled Fast Reactor (PGSFR), the Sodium Integral Effect Test Loop for Safety Simulation and Assessment (STELLA) program has been launched and the basic design of STELLA-2 facility was completed in 2015. The STELLA-2 is a scaled facility including all the major systems and components in PGSFR and is able to simulate the transient behavior....
Dr
Barbara Vezzoni
(Karlsruhe Institute of Technology, Germany)
28/06/2017, 17:50
Track 6. Test Reactors, Experiments and Modeling and Simulations
POSTER
The Karlsruhe Institute of Technology (KIT) and the Kyushu University (KU) have participated to the Coordinated Research Project (CRP) of the International Atomic Energy Agency (IAEA) on the Experimental Breeder Reactor II (EBR-II) Shutdown Heat Removal Test (SHRT). Two SHRT tests (SHRT-17 and SHRT-45R) representative, respectively, of Protected Loss of Flow (PLOF) and Unprotected Loss of Flow...
Mr
Pengrui Qiao
(China Institute of Atomic Energy, Beijing, China)
28/06/2017, 17:50
Track 6. Test Reactors, Experiments and Modeling and Simulations
POSTER
Unprotected loss of flow accident (ULOF) is the most typical severe accident in sodium cooled fast reactor, which is focused by scholars civil and abroad. Metal fuel has different safety characteristics with the oxide fuel as the important development direction of future sodium fast reactor, accident analysis of which is also a research focus at home and abroad. This paper bases on one...
Dr
Georgy Tikhomirov
(National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)), Mr
Mikhail Ternovykh
(National Research Nuclear University MEPhI (Moscow Engineering Physics Institute))
28/06/2017, 17:50
Track 6. Test Reactors, Experiments and Modeling and Simulations
POSTER
G.V. Tikhomirov, M.Y. Ternovykh, I.S. Saldikov, Y.S. Khomiakov, I.R.Suslov,Y.E. Shvetsov, A.S. Gerasimov, R.B. Bahdanovich, A.D. Smirnov, A.L. Farkhulina, E.J. Proshkina, V.I. Romanenko, V.S. Kharitonov, M.V. Bajaskhalanov
System of benchmarks of the class “tests” with prototype for neutron-physical and thermal hydraulic calculations of the BN-type reactors with nitride uranium-plutonium...
Mr
Mikhail Kriachko
(Joint Stock Company "State Scientific Centre of the Russian Federation – Institute for Physics and Power Engineering named after A. I. Leypunsky")
28/06/2017, 17:50
Track 2. Fast Reactor Operation and Decommissioning
POSTER
In the analysis of nuclear and radiation safety of existing and designed BN reactors considerable attention is paid to the problems associated with the formation of radioactive waste (RW) during their operation and decommissioning.
This paper describes the approaches to determine radiation characteristics of non-fuel compositions and structural elements of fuel assemblies (FA) and non-fuel...
Mr
Konstantin Mitrofanov
(JSC “SSC RF – IPPE”)
28/06/2017, 17:50
Track 6. Test Reactors, Experiments and Modeling and Simulations
POSTER
At the present time, the most perspective processes, which could form the basis of the technology of transmutation of radionuclides, are the processes associated with using of nuclear reactors, as well as sub-critical systems with high neutron flux generated using charged particle accelerators. The delayed neutrons have an important role in the safe management and kinetics of nuclear power...
Prof.
qiang zhao
(Harbin Engineering University)
28/06/2017, 17:50
Track 6. Test Reactors, Experiments and Modeling and Simulations
POSTER
China Experimental Fast Reactor (CEFR) is the first fast neutron breed reactor in China, which is different with PWR. In order to research the operational performance of CEFR, the real-time simulator was developed. The simulation of core physics is an important part of the simulator.
The neutron dynamic model used in the simulator is three dimensions and four groups neutron diffusion model,...
Mr
Alexandr Semchenkov
(JSC NIKIET, Moscow/Russia)
28/06/2017, 17:50
Track 6. Test Reactors, Experiments and Modeling and Simulations
POSTER
One of the most important elements of the BREST-OD-300 reactor facility is a steam generator (SG), which is a vertical heat exchanger with twisted pipes, immersed in liquid lead. To justify heat-hydraulic performance of SG and reliability of circulation was conducted complex of computational and experimental works.
Computational research were conducted with the help of numerical model of SG...
Mr
David Wootan
(Pacific Northwest National Laboratory)
28/06/2017, 17:50
Track 8. Professional Development and Knowledge Management
POSTER
An important goal of the U.S. Department of Energy’s Office of Nuclear Energy is to preserve the knowledge that has been gained in the United States on Liquid Metal Reactors by collecting, organizing and preserving technical information that could support the development of an environmentally and economically sound nuclear fuel cycle. The FFTF is the most recent LMR to operate in the United...
Mr
Song LI
(engineer)
28/06/2017, 17:50
Track 6. Test Reactors, Experiments and Modeling and Simulations
POSTER
Abstract:In order to obtain the heat transfer characteristics of the helium gap in the conditions of different thickness and power line in the high temperature range, on the basis of the previous research, the original test device was improved, through the theoretical design of double helium clearance, the test device can perform experiments under high temperature conditions. Compared with the...
Ivan A. Kodeli
(Institut Jožef Stefan, Jamova 39, 1000 Ljubljana, Slovenia)
28/06/2017, 17:50
Track 6. Test Reactors, Experiments and Modeling and Simulations
POSTER
Uncertainty Analysis of Kinetic Parameters for Design, Operation and Safety Analysis of SFRs
I.-A. Kodeli
Institut Jožef Stefan, Jamova 39, 1000 Ljubljana, Slovenia
G. Rimpault, P. Dufay, Y. Peneliau, J. Tommasi
French Alternative and Atomic Energy Commission (CEA), Cadarache Center, 13108 Saint-Paul-lez-Durance, France
E. Fridman
Helmholtz-Zentrum Dresden-Rossendorf (HZDR),...
Dr
Thomas Fanning
(Argonne National Laboratory)
28/06/2017, 17:50
Track 6. Test Reactors, Experiments and Modeling and Simulations
POSTER
Argonne has developed SAS4A/SASSYS-1 models for the benchmark analyses of the Experimental Breeder Reactor II (EBR-II) Balance-of-Plant (BOP) tests that represented protected and unprotected loss of heat sink conditions. The analyses were performed to support the validation of simulation tools and models used for SFR development. Previous benchmark results for the two BOP tests were in good...
Mr
Evgeny Seleznev
(NUCLEAR SAFETY INSTITUTE OF RUSSIAN ACADEMY OF SCIENCES)
28/06/2017, 17:50
Track 2. Fast Reactor Operation and Decommissioning
POSTER
The GEFEST800 code has been developed to carry out neutronic calculations of nu-clear power plant operation for sodium cooled fast breeder reactor BN-800 (stationary and transient from minimum controllable power level to full reactor power with drive control rods and fuel burning). The code allows to calculate the following parameters: keff; maximum reac-tivity reserve; effective reactivity of...
Mr
Sergey Rogozhkin
(JSC “Afrikantov OKBM”)
28/06/2017, 17:50
Track 6. Test Reactors, Experiments and Modeling and Simulations
POSTER
Using CFD codes for numerical simulation of thermohydraulic processes occurring in fast sodium reactors, specific character of heat transfer in liquid metals and complicity of computational model development should be taken into account due to integral layout of reactor equipment.
Application of universal non-Russian CFD codes (CFX, Star-CD, Fluent, etc.) does not enable to take into account...
Mr
Dmitriy Gremyachkin
(JSC “SSC RF – IPPE”)
28/06/2017, 17:50
Track 6. Test Reactors, Experiments and Modeling and Simulations
POSTER
The evaluated fission product yields data are an important characteristic. The validation method of the evaluated fission product yields data is based on comparing the characteristics of delayed neutrons produced by the summation method with appropriate recommended data. The total delayed neutron yields and the mean half-life of delayed neutron precursors were used as the delayed neutron...
