Results for “fast reactors” — Page 32

DOE OSTI2015

A burnup credit approach irradiated fast-neutron reactor fuels/ nuclides of interest and fuel storage application - 14248

The concept of taking into account the reduction of the reactivity of nuclear used fuel due to their burnup is referred to as 'Burnup Credit' (BUC). Economic incentives offered by the implementation of a BUC methodology in criticality analyses of Light Water Reactor (LWR) used fuel facilities are nowadays well-demonstrated. The current prospect of reprocessing the fissile and fertile subassemblies of the PHENIX reactor in France, and the development of the 4.-Generation Fast-neutron Reactors (FR) enable research projects in relevant fuel cycle operations. Under these circumstances, the CEA and AREVA-NC have decided to study a burnup credit approach for irradiated fuels of the PHENIX reactor. This paper focuses on a preliminary investigation on the use of burnup credit for used FR fuel operations. The first step consists in analyzing and selecting the nuclides of interest for burnup calculation, on the basis of the 27 BUC nuclides (12 actinides and 15 fissions products) chosen for LWR-MOx fuels [1]. These nuclides are then involved in a criticality calculation (a used fuel pool) to highlight the interest of a BUC approach. This study highlights the interest of a burnup credit approach for FR fuels and shows that the highest contribution to reactivity loss is due to only one fission product: the {sup 149}Sm. The opportunity to take account of only one fission product is highly promising and will considerably simplify the way to determine a penalizing fuel inventory. Moreover, it is to notice that results for different configurations may differ from those obtained for a fuel storage application. Thus, specific analyses have to be carried out in order to study the applicability of the conclusions of this paper. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2015

Decommissioning sodium dump tanks - 15547

The Nuclear Decommissioning Authority (NDA) is a non-departmental public body of the UK Government responsible for managing the effective and efficient clean-up of the UK's nuclear legacy. DSRL (Dounreay Site Restoration Limited, a wholly-owned subsidiary of the Cavendish Dounreay Partnership Ltd, a consortium of Cavendish Nuclear, CH2M HILL and URS) is the site licence company operating under contract to the NDA and funded by the NDA for the clean-up and demolition of the Dounreay Site, Britain's former centre of fast reactor research and development, near Thurso, Scotland, including decommissioning the Dounreay Prototype Fast Reactor (PFR). The plutonium metal fuel PFR was the second and last fast reactor to be built in the UK, with construction commencing in 1968. The reactor closed in 1994 and was de-fuelled, but the reactive sodium remaining in the cooling system required decommissioning, and approximately 12 tonnes of low level waste (LLW) sodium remained in the un-heated dump tanks. This paper highlights the approach used for decommissioning the dump tanks and their contents. The sodium was removed as a solid at ambient temperature, packaged, and staged for destruction at a later date. The tanks and pipework were then cleaned, size reduced and packaged for disposal as LLW. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2025

Software Validation Work With The ZPPR-15 Data

The analysis activities for fast reactors involve using many different pieces of software that are relied upon for their predictive capabilities. For this software to be considered reliable, documented proof that the predictions of the software are accurate is required. In this manuscript, the validation work that covers some of the Argonne software used in fast reactor design activities is discussed and displayed. This validation work includes neutron and gamma flux distributions, reaction rate distributions, and reactivity worth. In an ideal world, a reactor development program would have access to a comprehensive set of experimental facilities to help inform the design aspects of the reactor itself. While thermal-hydraulics experiments, and to a limited degree mechanical experiments, can be carried out today for validation needs, neutronics related experimental facilities are rather impractical because of the lack of experimental facilities. Given the desired time table for construction of new reactors, the reconstitution or creation of new neutronic experimental facilities is untenable and thus those reactor development programs must rely upon any available experimental measurements that are qualitatively similar to the design. While a methodology has been proposed to assess the similarity between the past experimental measurements and the reactor itself, that aspect is beyond the scope of this manuscript. In this manuscript, the focus is entirely placed on the analysis results for a series of experiments carried out at the ZPPR facility in Idaho in the mid-1980s. In this regard, this manuscript only shows the validation of the stated neutronics software for specific loadings of the ZPPR reactor. Because of the fuel form, its proposed enrichment, and the material content of the reactor core, the ZPPR-15 experiments were identified as potential validation data for the reactor. The ZPPR-15 experiments were intended as mockups of a 330 MWe Integral Fast Reactor program which was a follow on program to the Clinch River Breeder Reactor. In the ZPPR-15 series of experiments, measurements of the neutron spectrum, control rod worth, sodium void worth, foil reaction rate distributions, Doppler worth of heated samples, gamma dose, and axial expansion worth were all carried out and published. In many cases, these reactivity coefficients are good candidates to validate the reactivity coefficient calculation scheme used by the analysis software and included in the safety analysis activities of fast reactor development projects today. This manuscript discusses the modeling methodology and accuracy of the calculated experimental results using the LANL software MCNP and the ANL software package ARC (Argonne Reactor Codes). As will be shown, for many of the experimental measurements, the two software packages are found to be good predictive analysis tools for those experiments. In other cases, problems with the analysis methodology or underlying cross section data are exposed which indicates where predictive analysis is not as reliable. Finally, in some of the measurements the conclusion is reached that the experimental measurement cannot be reproduced with the analysis software as it is simply too difficult.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2026

Software Validation Work With The ZPPR-15 Data

The analysis activities for fast reactors involve using many different pieces of software that are relied upon for their predictive capabilities. For this software to be considered reliable, documented proof that the predictions of the software are accurate is required. In this manuscript, the validation work that covers some of the Argonne software used in fast reactor design activities is discussed and displayed. This validation work includes neutron and gamma flux distributions, reaction rate distributions, and reactivity worth. In an ideal world, a reactor development program would have access to a comprehensive set of experimental facilities to help inform the design aspects of the reactor itself. While thermal-hydraulics experiments, and to a limited degree mechanical experiments, can be carried out today for validation needs, neutronics related experimental facilities are rather impractical because of the lack of experimental facilities. Given the desired time table for construction of new reactors, the reconstitution or creation of new neutronic experimental facilities is untenable and thus those reactor development programs must rely upon any available experimental measurements that are qualitatively similar to the design. While a methodology has been proposed to assess the similarity between the past experimental measurements and the reactor itself, that aspect is beyond the scope of this manuscript. In this manuscript, the focus is entirely placed on the analysis results for a series of experiments carried out at the ZPPR facility in Idaho in the mid-1980s. In this regard, this manuscript only shows the validation of the stated neutronics software for specific loadings of the ZPPR reactor. Because of the fuel form, its proposed enrichment, and the material content of the reactor core, the ZPPR-15 experiments were identified as potential validation data for the reactor. The ZPPR-15 experiments were intended as mockups of a 330 MWe Integral Fast Reactor program which was a follow on program to the Clinch River Breeder Reactor. In the ZPPR-15 series of experiments, measurements of the neutron spectrum, control rod worth, sodium void worth, foil reaction rate distributions, Doppler worth of heated samples, gamma dose, and axial expansion worth were all carried out and published. In many cases, these reactivity coefficients are good candidates to validate the reactivity coefficient calculation scheme used by the analysis software and included in the safety analysis activities of fast reactor development projects today. This manuscript discusses the modeling methodology and accuracy of the calculated experimental results using the LANL software MCNP and the ANL software package ARC (Argonne Reactor Codes). As will be shown, for many of the experimental measurements, the two software packages are found to be good predictive analysis tools for those experiments. In other cases, problems with the analysis methodology or underlying cross section data are exposed which indicates where predictive analysis is not as reliable. Finally, in some of the measurements the conclusion is reached that the experimental measurement cannot be reproduced with the analysis software as it is simply too difficult.

Aliberti, Gerardo↗

DOE OSTI2022

A revisitation of the ENEA TAPIRO Fast Source Reactor neutronic capabilities

This work is devoted to the analysis of the neutronic characterization of the ENEA-Casaccia center RSV TAPIRO fast neutrons source reactor performed during the 80's in the frame of a joint ENEA-SCK-CEN Mol collaboration. The Belgian scientist Albert Fabry, who was available to investigate a fast research reactor to perform a full characterization of the system neutron flux, led the TAPIRO experimental campaign. In particular, the paper is focused on the comparison between some 'recommended' Fabry experimental reaction rates traverses and nowadays MCNP Monte Carlo and ERANOS deterministic calculation results. It appears that the quality of the results depends on the penetration depth into the copper reflector. The influence of copper nuclear data are more evident in the ERANOS results. Massimo Salvatores has provided a great contribution to the interpretation of the TAPIRO neutronics: this humble work is dedicated to him.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2021

Application of sensitivity analysis in DYMOND/Dakota to fuel cycle transition scenarios

The ability to perform sensitivity analysis has been enabled for the nuclear fuel cycle simulator DYMOND through its coupling with the design and analysis toolkit Dakota. To test and demonstrate these new capabilities, a transition scenario and multi-parameter study were devised. The transition scenario represents a partial transition from the US nuclear fleet to a closed fuel cycle with small modular LWRs and fast reactors fueled by reprocessed used nuclear fuel. Four uncertain parameters in this transition were studied – start date of reprocessing, total reprocessing capacity, the nuclear energy demand growth, and the rate at which the fast reactors are deployed – with respect to their impact on four response metrics. The responses – total natural uranium consumed, maximum annual enrichment capacity required, total disposed mass, and total cost of the nuclear fuel cycle – were chosen based on measures known to be of interest in transition scenarios and to be significantly impacted by the varying parameters. Furthermore, analysis of this study was performed both from the direct sampling and through surrogate models developed in Dakota to calculate the global sensitivity measures Sobol’ indices. This example application of this new capability showed that the most consequential parameter to most metrics was the share of new build capacity that is fast reactors. However, for the cost metric, the scaling factor of the energy demand growth was significant and had synergistic behavior with the fast reactor new build share.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2015

Investigation of Benefits from U/TRU Recycle - Quantification and Comparison to U/Pu Recycle

The recently completed comprehensive evaluation and screening of nuclear fuel cycle options identified a number of potentially promising fuel cycles for R&D that offer what could be considered by decision-makers as having the potential for significant improvement compared to the current U.S. fuel cycle. The fuel cycles that consistently performed the best were recycle fuel cycles that used self-sustaining fast reactors operating with either U/Pu or U/TRU recycle fuel and also included options where the fast reactors provided fissile materials to support operation of thermal reactors. However, based on the evaluation criteria and metrics used in the study, there was no difference in benefit between recycle of U/Pu and U/TRU (where TRU is plutonium and the minor actinides) while there were differences in the challenges for developing and deploying such fuel cycles, with U/TRU recycle being more challenging. This observation prompted the question as to the desirability of pursuing R&D on U/TRU recycle given that there may not be an increase in benefit. As a result, activities have been pursued to further investigate the performance differences between U/Pu and U/TRU recycle based on considering issues beyond those used in the evaluation and screening study to identify, if possible, areas where there are significant benefits of U/TRU recycle compared to U/Pu recycle. These new considerations focused on several areas, but especially on the impact on disposal of the HLW, which in the case of U/Pu recycle contains all of the minor actinides along with fission products, while in the case of U/TRU recycle only contains the losses of minor actinides from the reprocessing and recycle fuel fabrication operations. This difference in content has several implications. One impact is on the time dependent decay heat which can affect handling and the use of space in a geologic repository. Another impact concerns the HLW form and volume, since presence of minor actinides may adversely affect the ability to reduce HLW volume. The short-term radioactivity and long-term radiotoxicity of the HLW is also affected, which may be of more or less importance depending on the specific geologic disposal environment. To study these potential effects, a range of waste forms and disposal environments were used in the analysis, documenting to what extent the recycle of minor actinides in addition to plutonium may offer further benefit. Another area of investigation concerned the recycle fuel, for the fast reactor and for the thermal reactors they may support. Information to date indicates that U/Pu fuel may be simpler to fabricate and has a much more extensive database than U/TRU fuel, one of the reasons for the increased challenge for developing and deploying a U/TRU fuel cycle, and also indicates that heterogeneous recycle of the minor actinides may be even more difficult as compared to homogeneous recycle. This information was reviewed and updated to reflect the most recent studies for the purpose of informing on all aspects of the differences between U/Pu and U/TRU recycle. The results of all of these investigations will be presented to provide information on the findings concerning the value of U/TRU recycle.

Fast reactor recycle↗

DOE OSTI2025

AFC FAST Status Report Fiscal Year 2025

The Fission Accelerated Steady-state Test (FAST) methodology has been used to test several innovative metallic fuel designs for use in Sodium?cooled Fast Reactors (SFRs). The FAST method utilizes geometrically scaled fuel pins to produce improved burnup rates for fuels while maintaining semi-prototypic thermal conditions. This report provides a summary description of the test matrix and test objectives for the AFC FAST irradiation campaign as well as details of the as-built conditions for all rodlets. This report also include status update on the irradiation conditions for the rodlets as well as post-irradiation examination (PIE) results on low to mid?range burnup fuel rodlets and a summary power history for those rodlets. A brief discussion provides some early interpretation of the results and whether or not FAST is producing results typical of metallic fuels. Lastly, an outlook for the work expected to be completed in fiscal year (FY) 2026 is also included.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2018

Weighted delta-tracking in scattering media

In this work, we expand the weighted delta-tracking routine to include a treatment for scattering. The weighted delta-tracking routine adds survival biasing to normal delta-tracking, improving problem figure of merit. In the original formulation of this method, only absorption events were considered. We have expanded the method to include scattering and investigated the method’s effectiveness with two test cases: a pressurized water reactor pin cell and a fast reactor pin cell. We compare the figure of merit for calculating infinite flux and total cross-section while incrementally changing the amount of weighted delta-tracking used. Thus, we find that this new weighted delta-tracking (WDT) routine has strong potential to improve the efficiency of fast reactor calculations, and may be useful for light water reactor calculations.

97 MATHEMATICS AND COMPUTING↗

DOE OSTI2022

University Contributions to the Versatile Test Reactor (VTR) (FY2022) (Book of Abstracts)

Students were an integral part of the work done on the Versatile Test Reactor (VTR) this fiscal year (FY). In fact, these students participated in the research, development, and deployment of technologies needed to move the design of advanced experimental vehicles forward. This document summarizes the work performed by these students in several areas, including the Extended Length Test Assembly (ELTA)-Sodium-cooled Fast Reactor (SFR); ELTA-Molten Salt Reactor (MSR); ELTA- Lead/lead-bismuth cooled Fast Reactor (LFA); ELTAGas-cooled Fast Reactor (GFR); ELTA-Materials (M); and Cross-Cutting Technologies areas.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2019

Complete Status Report Documenting the Processing, Microstructure and High-Temperature Mechanical Properties of the Nanostructured Ferritic Alloy OFRAC

This report provides a summary of the new nanostructured ferritic alloy, OFRAC (Oak Ridge Fast Reactor Advanced Fuel Cladding), that was developed for the advanced sodium fast reactor fuel cladding and core internals. The composition of OFRAC, which is Fe-12Cr-1Mo-03.Ti-0.3Nb + 0.3Y2O3 (wt. %), was chosen to achieve high temperature strength as well as high temperature creep performance and swelling resistance requirements that past structural materials such as austenitic stainless steels and ferritic/martensitic steels, i.e. HT-9, are not suitable as SFR cladding materials. The mechanical alloying procedures were described for producing OFRAC and the detailed characterization studies revealed uniformly distributed, high concentration of Y-Ti-O enriched oxide particles, or nanoclusters. The Nb addition that was intended to sequester interstitial C and N that are commonly introduced as contamination during ball milling of powder was found to be very effective in forming coarse size Nb carbonitrides along with excess Ti that removes these interstitial elements from the bcc-Fe matrix of OFRAC. The tensile tests on the OFRAC in the extruded condition showed excellent balance in strength and ductility from room temperature to 800°C. Both the microstructure and tensile properties were not significantly degraded after annealing the extruded OFRAC samples for 8 h at 1050°C and 1150°C. The latter temperature is ~76% of the melting point of OFRAC. These results demonstrated the stability of the microstructure and tensile properties to extremely high temperatures. The thermal creep resistance at elevated temperatures and stresses was investigated using strain rate jump testing. The results showed that OFRAC possessed high stress exponents at high stresses and temperatures of 550°C, 600°C, 700ºC and 800°C. This thermal dependence of minimum creep rate indicates that OFRAC will show quite low creep rates at higher temperatures and stresses compared to HT-9 and stainless steels. These significant improvements indicate that the NFA OFRAC alloy is a compelling and viable choice for sodium fast reactor fuel cladding.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2019

SAM User's Guide

The System Analysis Module (SAM) is a modern system analysis tool being developed at Argonne National Laboratory for advanced non-LWR safety analysis. It aims to provide fast-running, whole-plant transient analyses capability with improved-fidelity for Sodium-cooled Fast Reactors (SFR), Lead-cooled Fast Reactors (LFR), and Molten Salt Reactors (MSR) or Fluoride-cooled High-temperature Reactors (FHR). SAM takes advantage of advances in physical modeling, numerical methods, and software engineering to enhance its user experience and usability. It utilizes an object-oriented application framework (MOOSE), and its underlying meshing and finite-element library (libMesh) and linear and non-linear solvers (PETSc), to leverage the modern advanced software environments and numerical methods. This document provides a user’s guide, which will help users understand the input description and core capabilities of the SAM code. A brief overview of the code is presented, as well as how to obtain and run it. The input syntax for various parts of the code is provided. Additionally, a number of example problems, starting with simple unit component problems to problems with increasing complexity, are provided. Because the code is still under active development, this SAM User’s Guide will evolve with periodic updates.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗

DOE OSTI2021

SAM User’s Guide

The System Analysis Module (SAM) is a modern system analysis tool being developed at Argonne National Laboratory for advanced non-LWR safety analysis. It aims to provide fast-running, whole-plant transient analyses capability with improved-fidelity for Sodium-cooled Fast Reactors (SFR), Lead-cooled Fast Reactors (LFR), and Molten Salt Reactors (MSR) or Fluoride-cooled High-temperature Reactors (FHR). SAM takes advantage of advances in physical modeling, numerical methods, and software engineering to enhance its user experience and usability. It utilizes an object-oriented application framework (MOOSE), and its underlying meshing and finite-element library (libMesh) and linear and non-linear solvers (PETSc), to leverage the modern advanced software environments and numerical methods. This document provides a user’s guide, which will help users understand the input description and core capabilities of the SAM code. A brief overview of the code is presented, as well as how to obtain and run it. The input syntax for various parts of the code is provided. Additionally, a number of example problems, starting with simple unit component problems to problems with increasing complexity, are provided. Because the code is still under active development, this SAM User’s Guide will evolve with periodic updates.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2025

SAM User's Guide

The System Analysis Module (SAM) is a modern system analysis tool being developed at Argonne National Laboratory for advanced non-LWR safety analysis. It aims to provide fast-running, whole-plant transient analyses capability with improved-fidelity for Sodium-cooled Fast Reactors (SFR), Lead-cooled Fast Reactors (LFR), and Molten Salt Reactors (MSR) or Fluoride-cooled High-temperature Reactors (FHR). SAM takes advantage of advances in physical modeling, numerical methods, and software engineering to enhance its user experience and usability. It utilizes an object-oriented application framework (MOOSE), and its underlying meshing and finite-element library (libMesh) and linear and non-linear solvers (PETSc), to leverage the modern advanced software environments and numerical methods. This document provides a user’s guide, which will help users understand the input description and core capabilities of the SAM code. A brief overview of the code is presented, as well as how to obtain and run it. The input syntax for various parts of the code is provided. Additionally, a number of example problems, starting with simple unit component problems to problems with increasing complexity, are provided. Because the code is still under active development, this SAM User’s Guide will evolve with periodic updates.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2022

Report on Fuel Cycle Facility Requirements for Deployment of Demonstration Reactors and Potential Evolutionary Fuel Cycle Scenarios

A series of fuel cycle scenarios studies were performed to inform on fuel cycle capacities and facilities needed for large-scale deployment of the Advanced Reactor Demonstration Program (ARDP) reactors and potential future evolutionary fuel cycle scenarios. The reactor deployment and evolutionary fuel cycle scenarios from the present to 2100 were developed based on the following assumptions: 1) achievement of a net-zero emissions economy in the United States by 2050, which requires a nuclear energy generation capacity of ~250 GWe by 2050, 2) the U.S. economic growth of 1% per year from 2051 to 2100, which results in ~340 GWe of nuclear energy capacity in 2100, and 3) commercial-scale recycling and high burnup fuel technologies are available after 2050. Thus, evolutionary fuel cycles with those advanced nuclear technologies start after 2050. A single once-through fuel cycle scenario was assumed from the present to 2050 to achieve a net-zero emissions economy in the United States, and the following four evolutionary fuel cycle scenarios from 2051 to 2100 were considered, 1) Once through fuel cycle with ARDP reactors (Natrium and Xe-100), 2) Once-through fuel cycle with Breed-and-Burn (B&B) fast reactors, 3) Recycling fuel cycle of used metallic fuel in fast reactors, and 4) Recycle fuel cycle of both used uranium oxide and metallic fuels in fast reactors. The projected front-end and back-end fuel cycle capacity demands are compared with the current domestic and global (if needed) fuel cycle capacities.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2022

A multiphysics model of the versatile test reactor based on the MOOSE framework

The traditional modeling approach for sodium fast reactor cores relies on separate physics models, where the fuel performance, thermal–hydraulics, and neutronics calculations required to predict the core physics characteristics for nominal conditions are decoupled by relying on user-imposed boundary conditions. Here, this paper aims at evaluating the impact of multiphysics simulations for predicting the core characteristics of the Versatile Test Reactor, which is being designed as a 300-MWt sodium-cooled fast reactor. The purpose of the Versatile Test Reactor is to accelerate the testing of advanced nuclear materials in the United States. The proposed multiphysics model relies on the Griffin reactor physics code, the SAM thermal–hydraulic system code, the BISON fuel performance code, as well as generic Multiphysics Object-Oriented Simulation Environment capabilities implemented in the open-source tensor mechanics module. For k eff calculations, the introduction of a tight coupling between the neutronics, thermo-mechanical and thermal–hydraulics models induces a change of around 543 pcm in the eigenvalue, compared to the traditional standalone neutronics calculation where approximate temperature profiles are used. The multiphysics model is then employed for quantifying the impact of the thermal conductivity uncertainties on some of the key figures of merit, such as the fuel centerline temperature, assembly powers, and keff for nominal core conditions. As anticipated, uncertainties on fuel thermal conductivity mostly impact the fuel centerline temperature, and to a lesser extend the k eff .

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2024

ARC-100 Reactor Security-by-Design Summary

This report applies the security-by-design methodology developed in a previous National Nuclear Security Administration–sponsored work to the Advanced Reactor Concepts 100 (ARC-100) sodium-cooled fast reactor (SFR) design. The report contains no proprietary information specific to the ARC 100 reactor. The insights developed in this report are high-level, and generally applicable to other sodium fast reactor designs. The information presented here is the result of a qualitative safety-based analysis and would not inform any potential adversary beyond what would be found in a docketed safety analysis report. The scope of this present report covers ARC-100’s reactor core, used fuel storage, and used fuel assembly wash station. These systems are also compared to a generic SFR design assumed in the previous study. The security assessment results show changes in structures, systems, and components (SSCs) safety importance relative to the generic SFR SSCs. However, the consequence assessment results are the similar to a previously assessed generic SFR. Several SSCs have higher importance rankings than others, and it is recommended that protection efforts are prioritized for these SSCs. This work will continue in the Fiscal Year 2025 for the remaining ARC-100 systems, including cesium trap, sodium cold trap, noble gas decay tanks (dewar bottles), and used fuel dry storage facility, to provide safety-and-security-by-design insights and recommendations on non-core systems. Results from this work will furnish a technical justification for the feasibility of these solutions for the ARC reactor's design and, where applicable, identify any regulatory benefits conferred by the proactive design aspect within a risk management framework. This initiative will contribute to a more secure design of the ARC reactor and support its licensing process.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2012

Computational study: Reduction of iron corrosion in lead coolant of fast nuclear reactor

In this paper we report molecular dynamics simulation results of iron (cladding) corrosion in interaction with lead coolant of fast nuclear reactor. The goal of this work is to study effect of oxygen injection to the coolant to reduce iron corrosion. By evaluating diffusion coefficients, radial distribution functions, mean-square displacement curves and observation of crystal structure of iron before and after oxygen injection, we concluded that a significant reduction of corrosion can be achieved by issuing about 2% of oxygen atoms into lead coolant.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗