Benchmark exercise for fluid flow simulations in a liquid metal fast reactor fuel assembly
Not Available
Merzari, E.↗
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Fast Reactor Methods Development Plan
Fast reactor design and safety analysis methods (computer codes) have been under development for approximately sixty years. During the first forty years, code development was driven by the immediate needs of operational facilities such as Experimental Breeder Reactors I and II and the Fast Flux Test Facility. Development was further driven by programmatic needs to support advanced reactor programs such as the Clinch River Breeder Reactor Project and the Advanced Liquid Metal Reactor Program. By the early 1990s, design and safety analysis codes had been well established to addressed the unique characteristics of fast reactors. During the subsequent twenty years, however, development was nearly stagnant as a result of U.S. Government policy changes that abruptly canceled nearly all of the fast reactor research in the United States. As a result, the state of fast reactor design and safety analysis methods has deteriorated.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Path to a commercial fast reactor option in the United States
Fast reactors represent one technology with the potential to transform the energy sector through increased resource utilization and unique waste management opportunities that are not available through the established light water reactor (LWR) technology. The technology is 'old' by 21 st century standards, as the United States and other nations have funded fast reactor research and development efforts from the birth of nuclear power that have led to deployment at all scales from experimental to commercial. Yet, in spite of great promise, the experience with fast reactors can be characterized as mixed, with both notable successes and failures. Why then, after 50 years of development, is there no commercial fast reactor operating in the United States? Is commercialization of the technology possible? If so, what business case is needed for a commercial utility to operate a fast reactor? This paper will attempt to sketch a possible path for the development and demonstration of technology needed to make fast reactors a real option for commercial operation by mid-century. This path starts with today's proven LWR technology and uranium oxide fuels and takes an evolutionary route to a first commercial fast reactor that complements and supports the existing reliable LWR fleet. This paper is not intended to promote any one technology or approach, but instead seeks to illustrate the value of national level support for an energy technology program oriented toward demonstration and seeks eventual deployment of transformational energy systems on the scale and time-frame that matters. The observations and conclusions presented are derived primarily from the collective experience and expertise of the authors who represent a broad range of commercial perspectives from the utility, vendor, research and development, and regulatory policy communities. (authors)
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Proposed Adaptation of the Standard Review Plan NUREG-0800, Chapter 4 (Reactor) for Sodium-Cooled Fast Reactors and Modular High-Temperature Gas-Cooled Reactors
This report provides a proposed adaptation of a previous regulatory gap analysis of Chapter 4 (Reactor) of NUREG-0800 , Standard Review Plan (SRP) for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light Water Reactor] Edition, for applicability to advanced reactors; specifically, the sodium-cooled fast reactor (SFR) and the modular high temperature gas-cooled reactor (mHTGR). SRP Chapter 4, which addresses reactor components, was selected for the review because of the possible significant differences in advanced non-light water reactor (non-LWR) technologies compared with the current LWR-based Chapter 4. SFR and mHTGR technologies were chosen for this gap analysis because of their diverse designs and the availability of significant historical design detail. This report satisfies M3AT-16OR2001042, “Complete Development of an Advanced Reactor Adaptation of Chapter 4 NUREG-0800,” in support of the US Department of Energy (DOE) Office of Nuclear Energy (DOE-NE) Advanced Reactor Technology program.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Transient behavior of a molten salt fast reactor under two-phase flow conditions with helium bubbling
Not provided.
Nuclear Science & Technology↗
Proposed Advanced Reactor Adaptation of the Standard Review Plan NUREG-0800 Chapter 4 (Reactor) for Sodium-Cooled Fast Reactors and Modular High-Temperature Gas-Cooled Reactors
This report proposes adaptation of the previous regulatory gap analysis in Chapter 4 (Reactor) of NUREG 0800, Standard Review Plan (SRP) for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light Water Reactor] Edition. The proposed adaptation would result in a Chapter 4 review plan applicable to certain advanced reactors. This report addresses two technologies: the sodium-cooled fast reactor (SFR) and the modular high temperature gas-cooled reactor (mHTGR). SRP Chapter 4, which addresses reactor components, was selected for adaptation because of the possible significant differences in advanced non-light water reactor (non-LWR) technologies compared with the current LWR-based description in Chapter 4. SFR and mHTGR technologies were chosen for this gap analysis because of their diverse designs and the availability of significant historical design detail.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Upsampling Monte Carlo Reactor Simulation Tallies in Depleted Sodium-Cooled Fast Reactor Assemblies Using a Convolutional Neural Network
The computational demand of neutron Monte Carlo transport simulations can increase rapidly with the spatial and energy resolution of tallied physical quantities. Convolutional neural networks have been used to increase the resolution of Monte Carlo simulations of light water reactor assemblies while preserving accuracy with negligible additional computational cost. Here, we show that a convolutional neural network can also be used to upsample tally results from Monte Carlo simulations of sodium-cooled fast reactor assemblies, thereby extending the applicability beyond thermal systems. The convolutional neural network model is trained using neutron flux tallies from 300 procedurally generated nuclear reactor assemblies simulated using OpenMC. Validation and test datasets included 16 simulations of procedurally generated assemblies, and a realistic simulation of a European sodium-cooled fast reactor assembly was included in the test dataset. We show the residuals between the high-resolution flux tallies predicted by the neural network and high-resolution Monte Carlo tallies on relative and absolute bases. The network can upsample tallies from simulations of fast reactor assemblies with diverse and heterogeneous materials and geometries by a factor of two in each spatial and energy dimension. The network’s predictions are within the statistical uncertainty of the Monte Carlo tallies in almost all cases. This includes test assemblies for which burnup values and geometric parameters were well outside the ranges of those in assemblies used to train the network.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Testing Fast Reactor Fuels in a Thermal Reactor: A Comparison Report
The testing of fast reactor fuels in the thermal-spectrum of the Advanced Test Reactor using cadmium-filtering has been underway since 2003. The objective of this experimental approach has not been to create testing conditions that are fully prototypic of a fast reactor, but to create a thermal profile inside of test fuel rodlets that is nearly prototypic of corresponding fast reactor conditions. By doing so, this experimental approach should allow for the study of fast reactor fuel performance phenomena that are primarily dependent upon the conditions of temperature and/or temperature gradient inside the fuel to be possible using a thermal test reactor. Validation of this assertion has been undertaken by comparing fuel performance phenomena observed in fuels irradiated in cadmium-filtered positions in ATR to similar or identical fuels irradiated under similar conditions in genuine fast reactors. This report documents a preliminary comparison of limited fuel performance data available from the testing of metallic, oxide, and nitride fuels in cadmium-filtered positions in the ATR to the performance of identical or similar fuels irradiated in genuine fast reactors. Although a variety of fuel performance metrics are presented, particular attention is paid to those phenomena that are primarily, or significantly, dependent on conditions of temperature and temperature gradient with the fuel. Although the comparisons presented in this report are at this time limited in scope and preliminary in nature, they do support the assertion that the objectives of the cadmium-filtered testing approach in ATR are both feasible and appear sound. The available evidence supports the assertion that use of cadmium-filtering creates irradiation conditions in the ATR that are both relevant to, and suitable for, fast reactor fuel performance testing.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Testing Fast Reactor Fuels in a Thermal Reactor: A Comparison Report
The testing of fast reactor fuels in the thermal-spectrum of the Advanced Test Reactor using cadmium-filtering has been underway since 2003. The objective of this experimental approach to create a temperature profile inside of test fuel rodlets that is nearly prototypic of corresponding fast reactor conditions. This requires, necessarily, that the power generation profile inside of test fuel rodlets also be nearly prototypic of corresponding fast reactor conditions. By doing so, this experimental approach should allow for the study of fast reactor fuel performance phenomena that are primarily dependent upon the conditions of temperature and/or temperature gradient inside the fuel to be possible using a thermal test reactor. Validation of this assertion has been undertaken by both supporting analyses and by comparing fuel performance phenomena observed in fuels irradiated in cadmium-filtered positions in ATR to similar or identical fuels irradiated under similar conditions in genuine fast reactors. This report documents a preliminary comparison of fuel performance data available from the testing of metallic, oxide, and nitride fuels in cadmium-filtered positions in the ATR to the performance of identical or similar fuels irradiated in genuine fast reactors. Although a variety of fuel performance metrics are presented, particular attention is paid to those phenomena that are primarily, or significantly, dependent on conditions of temperature and temperature gradient with the fuel. Although the comparisons presented in this report are at this time limited in scope and preliminary in nature, they do support the assertion that the objectives of the cadmium-filtered testing approach in ATR are both feasible and appear sound. The analyses and comparisons presented in this report show that ATR irradiations performed using cadmium shrouding are sufficiently prototypic that they can be used with confidence in the development and testing of fast reactor fuels.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Testing Fast Reactor Fuels in a Thermal Reactor: A Comparison Report
The testing of fast reactor fuels in the thermal-spectrum of the Advanced Test Reactor using cadmium-filtering has been underway since 2003. The objective of this experimental approach has not been to create testing conditions that are fully prototypic of a fast reactor, but to create a thermal profile inside of test fuel rodlets that is nearly prototypic of corresponding fast reactor conditions. By doing so, this experimental approach should allow for the study of fast reactor fuel performance phenomena that are primarily dependent upon the conditions of temperature and/or temperature gradient inside the fuel to be possible using a thermal test reactor. Validation of this assertion has been undertaken by comparing fuel performance phenomena observed in fuels irradiated in cadmium-filtered positions in ATR to similar or identical fuels irradiated under similar conditions in genuine fast reactors. This report documents a preliminary comparison of limited fuel performance data available from the testing of metallic, oxide, and nitride fuels in cadmium-filtered positions in the ATR to the performance of identical or similar fuels irradiated in genuine fast reactors. Although a variety of fuel performance metrics are presented, particular attention is paid to those phenomena that are primarily, or significantly, dependent on conditions of temperature and temperature gradient with the fuel. Although the comparisons presented in this report are at this time limited in scope and preliminary in nature, they do support the assertion that the objectives of the cadmium-filtered testing approach in ATR are both feasible and appear sound. The available evidence supports the assertion that use of cadmium-filtering creates irradiation conditions in the ATR that are both relevant to, and suitable for, fast reactor fuel performance testing.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
An overview of the activities carried out by the IAEA Technical Working Group on Fast Reactors
The Technical Working Group on Fast Reactors (TWG-FR) of the International Atomic Energy Agency (IAEA) is a group of experts to provide advice and support programme implementation, reflecting a global network of excellence and expertise in the area of advanced technologies and R and O for fast reactors and sub-critical hybrid systems. Consistent with its mandate, the TWG-FR assists in defining and carrying out the Agency's activities in the field of nuclear power technology development for fast reactors and sub-critical hybrid systems. It promotes the exchange of information on national and multi-national programmes and new developments and experience, with the goal to identify and review problems of importance and to stimulate and facilitate cooperation, development and practical application of fast reactors and sub-critical hybrid systems. Finally, the TWG-FR provides Member States with information about the current status and development trends of advanced technologies for fast reactors and sub-critical hybrid systems. The TWG-FR coordinates its activities with other IAEA projects and initiatives, especially those of the Technical Working Group on Nuclear Fuel Cycle Options (TWG-NFCO), the Dept. of Nuclear Safety and Security and, last but not least, the International Project on Innovative Nuclear Reactors and the Fuel Cycle INPRO. It also cooperates with FR-related initiatives implemented within the framework of other international programmes (e.g. GIF, OECD/NEA, ESNII, etc.). The paper presents the recent relevant activities of the TWG-FR as well as the status of the current Coordinated Research Projects jointly carried out by a number of organizations representing the various Member States within the TWG-FR. In particular they concern benchmark analyses of: - the sodium natural circulation tests performed during Phenix End-of-Life experiments; - the sodium natural convection in the upper plenum of the Monju reactor vessel; - an EBR-II Shutdown Heat Removal Test. (authors)
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
The IAEA international conference on fast reactors and related fuel cycles: highlights and main outcomes
The 'International Conference on Fast Reactors and Related Fuel Cycles', which is regularly held every four years, represents the main international event dealing with fast reactors technology and related fuel cycles options. Main topics of the conference were new fast reactor concepts, design and simulation capabilities, safety of fast reactors, fast reactor fuels and innovative fuel cycles, analysis of past experience, fast reactor knowledge management. Particular emphasis was put on safety aspects, considering the current need of developing and harmonizing safety standards for fast reactors at the international level, taking also into account the lessons learned from the accident occurred at the Fukushima- Daiichi nuclear power plant in March 2011. Main advances in the several key areas of technological development were presented through 208 oral presentations during 41 technical sessions which shows the importance taken by fast reactors in the future of nuclear energy.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Qualification of Digitized Legacy Fast Reactor Data
The Integral Fast Reactor (IFR) fuel compatibility test program (1984-1994) included a variety of fuel pin examinations conducted at the Hot Fuel Examination Facility (HFEF) and the Alpha-Gamma Hot Cell Facility (AGHCF). Hard copy data records of these examinations have been recovered, scanned, and preserved in PDF format. Many hard copy records are now qualified in accordance with an NRC-approved Quality Assurance Program Plan (QAPP), and there is an ongoing effort to qualify additional legacy records. This legacy fuel performance data is vital to support design and licensing of fast reactors with validation of state-of-the-art codes and advanced methods for design and analysis. Stakeholders can most easily utilize this data when the PDF scans have been converted into digital data tables. However, qualification of the scanned hard copy data does not qualify the digital data file resulting from the digitization of the data contained in the record; the subject matter expert (SME) must make a review of the digitized data table as well before it can be designated as qualified. This report outlines a peer review process to qualify the digital data file(s), typically in CSV format, corresponding to hard copy records in accordance with the existing QAPP.
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Demonstration of MOOSE-based Griffin reactor physics, code for heterogeneous lead-cooled fast reactor analysis
The MOOSE-based reactor physics code Griffin was assessed on a heterogeneous pin-resolved model of a prototype lead-cooled fast reactor assembly. This model was developed in preparation for future use in MOOSE-based multiphysics calculations for computing hot channel factors. Heterogeneous multigroup cross sections were prepared using the fast reactor multi-group cross section processing code MC{sup 2}-3 using a two-step method. Griffin simulations were performed using the DFEM-SN solver on 576 cores on Argonne's LCRC cluster. Diffusion-based acceleration methods were applied (NDA and CMFD). Reference solutions were generated with continuous energy MCNP and the hybrid MOC/finite element solver PROTEUS-MOC for code-to-code comparison. Space-angle convergence studies were conducted to observe convergence in k-eigenvalue and axial pin power distributions. The fully resolved Griffin calculation was within 68 pcm of the MCNP eigenvalue and exhibited max 1.2% relative error in the axial pin power distribution. Griffin produced nearly identical results to PROTEUS-MOC when using the same 9-group multigroup cross-section set. Griffin demonstrated favorable scaling in wall-clock time and memory usage when using diffusion-based acceleration methods. Griffin is capable of simulating the pin-resolved heterogeneous LFR assembly with good accuracy and performance, and is suitable for future use in coupled high-fidelity hot channel factor simulations. (authors)
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Power flattening on modified CANDLE small long life gas-cooled fast reactor
Gas-cooled Fast Reactor (GFR) is one of the candidates of next generation Nuclear Power Plants (NPPs) that expected to be operated commercially after 2030. In this research conceptual design study of long life 350 MWt GFR with natural uranium metallic fuel as fuel cycle input has been performed. Modified CANDLE burn-up strategy with first and second regions located near the last region (type B) has been applied. This reactor can be operated for 10 years without refuelling and fuel shuffling. Power peaking reduction is conducted by arranging the core radial direction into three regions with respectively uses fuel volume fraction 62.5%, 64% and 67.5%. The average power density in the modified core is about 82 Watt/cc and the power peaking factor decreased from 4.03 to 3.43.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Modern nanostructured ferritic alloys: A compelling and viable choice for sodium fast reactor fuel cladding applications
Sodium fast reactor cores present a truly challenging environment for the fuel cladding and core structural materials that limit achievable fuel burnup in these systems. Modern nanostructured ferritic alloys have the potential to deliver the desired high performance characteristics where historic austenitic and ferritic/martensitic alloys fall short. Herein, a new nanostructured ferritic alloy, OFRAC ( O ak Ridge F ast R eactor A dvanced Fuel C ladding), is developed and demonstrated in cladding geometry with a length > 1 m. The alloy composition and microstructure are tailored to deliver dramatically improved strength and creep resistance. At 600 °C the ultimate tensile strength is ~600 MPa and the stress to induce a steady state strain rate of 10 -6 s -1 is 500 MPa vs. 200 MPa for traditional ferritic/martensitic alloys. A very high defect sink density was engineered in these alloys that is expected to further enhance the traditional good swelling and irradiation creep resistance of ferritic/martensitic steels. These significant improvements along with the demonstrated viability for seamless cladding production indicate that this alloy is a compelling and viable choice for sodium fast reactor fuel cladding applications.
36 MATERIALS SCIENCE↗
Selection of materials for sodium fast reactor steam generators
Sodium Fast Reactor (SFR) is considered in France as the most mature technology of the different Generation IV systems. In the short-term the designing work is focused on the identification of the potential tracks to demonstrate licensing capability, availability, in-service inspection capability and economical performance. In that frame materials selection for the major components, as the steam generator, is a particularly key point managed within a French Research and Development program launched by AREVA, CEA and EDF. The choice of the material for the steam generator is indeed complex because various aspects shall be considered like mechanical and thermal properties at high temperature, interaction with sodium on one side and water and steam on the other side, resistance to wastage, procurement, fabrication, weldability and ability for inspection and in-situ intervention. The following relevant options are evaluated: the modified 9Cr1Mo ferritic-martensitic grade and the Alloy 800 austenitic grade. The objective of this paper is to assess for both candidates their abilities to reach the current SFR needs regarding material design data, from AFCEN RCC-MRx Code in particular, compatibility with environments and manufacturability. (authors)
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗