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At least 1,153 records · Page 64
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

EIC Hadron Beamline Vacuum Studies

Ninety percent of the EIC hadron ring vacuum chamber consists of cold-bore sections running through and interconnecting 4.55 K RHIC superconducting (SC) magnets. The EIC operating specification requires shorter bunches and 3x higher intensity beams which are not appropriate for the present RHIC stainless steel cold-bore beam tube. The intensity and emittance of the hadron beams will degrade due to interactions with residual gas or vacuum instabilities arising from the expected resistive-wall (RW) heating, electron clouds, and beam-induced desorption mechanisms. Without strategies to limit RW heating, major cryogenic system modifications are needed to prevent SC magnet quenches. The SC magnet cold-bore beam tubes will be equipped with a high RRR copper-clad stainless steel screen to significantly reduce RW heating and so the effect on the SC magnet cryogenic heat load and temperature. A thin amorphous carbon film applied to the beam-facing copper surface will suppress electron cloud formation. This paper discusses the vacuum requirements imposed by the EIC hadron beams and the plans to achieve the necessary vacuum and thermal stability that ensure acceptable beam quality and lifetime.

43 PARTICLE ACCELERATORS↗
DOE OSTI2024

Comparative study of decay heat calculations with FLUKA and MCNP/CINDER2008

In designing a high-power tungsten target, decay heat driven temperature rise in the spallation volume is a safety concern during maintenance and in loss of coolant accidents. As tungsten hydrates and becomes volatile in steam at above 800 ◦C, it is important to keep the target temperature below this threshold when active cooling is unavailable. Decay heat in a target is calculated with particle transport simulation codes combined with transmutation codes. The calculated decay heat usually differs depending on the nuclear cross sections and the decay particle transport models built in the code architecture. In this paper, we calculated decay heat of a water-cooled tungsten target using popular particle transport codes, FLUKA and MCNP6® paired with CINDER2008 and compared the results. The target-moderator-reflector (TMR) system is modeled with a water-cooled solid tungsten target, water premoderators, liquid hydrogen cold moderators and beryllium reflectors. Water-cooled stainless-steel shielding is modeled around the TMR system. The tungsten volume is clad with a thin layer of erosion/corrosion resistant material. This study provides information about the uncertainty range in decay heat prediction of high-power spallation targets for hazard analysis.

Lee, Yong Joong↗
DOE OSTI2020

Separate Effects Validation for Subcooled Boiling in CTF

In 2010, the U.S. Department of Energy created its first Energy Innovation Hub, which is focused on developing high-fidelity and high-resolution modeling and simulation (M&S) tools for modeling of light water reactors (LWRs). This hub, the Consortium for Advanced Simulation of LWRs (CASL), has developed an LWR simulation tool called the Virtual Environment for Reactor Applications (VERA). The multi-physics capability of VERA is achieved through the coupling of single-physics codes, including CTF (the CASL version of Coolant Boiling in Rod Arrays— Three Field (COBRA-TF)), Michigan Parallel Characteristics Transport (MPACT), BISON, and Materials Performance and Optimization (MPO) Advanced Model for Boron Analysis (MAMBA). As part of its M&S efforts, CASL has identified various challenge problems, including Crud Induced Power Shift (CIPS), Crud-Induced Localized Corrosion (CILC), Pellet-Cladding Interaction (PCI), and Departure from Nucleate Boiling (DNB). This work addresses CASL milestone L2:VVI.P19.03, which focuses on uncertainty quantification of crud, which is relevant to both CIPS and CILC. This is achieved through an analysis and separate effects validation of the thermal hydraulic phenomenon known as subcooled boiling. As part of this work, various sources of experimental data are examined and compared to different options for empirical modeling of subcooled boiling. Through this analysis, a complete understanding of the underlying models and their implementation details are understood. A subset of these data are incorporated into a separate effects validation study of CTF. The Westinghouse Advanced Loop Tester (WALT) and Rohsenow experiments are modeled, and it is shown that the newly-implemented Gorenflo correlation is more accurate than the existing Chen and Thom correlations.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Irradiation enhanced diffusion and diffusional creep in U₃Si₂

U₃Si₂ is a candidate as an advanced fuel for light water reactors (LWRs). Compared to UO₂, there are significant data gaps for the thermophysical and thermomechanical properties of U₃Si₂. Atomic scale simulations can help fill in those gaps, while also providing a more mechanistic understanding of the material properties. Creep is an important property for fuel performance as it influences the pellet-cladding gap, which in turn affects the temperature of the fuel (a key parameter in many other aspects of fuel behavior). In this work, we employed a DFT-informed cluster dynamics framework to predict point defect concentrations in U₃Si₂. The concentrations, diffusivities, and volumes of these defects were used to develop a diffusional creep model based on bulk (Nabarro-Herring) and grain boundary (Coble) diffusion processes. It was found that irradiation enhanced Nabarro-Herring creep is dominant at low temperatures, while Coble creep dominates for high temperatures. The model compares well against available experimental data and has been implemented in BISON. A demonstration case using simple power profiles has been carried out, showing that negligible creep occurs due to the low temperatures experienced by U₃Si₂, a consequence of its high thermal conductivity.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Research Needs for Uranium-Zirconium- Based Metallic Fuels

The purpose of this report is to summarize the needs and present recommendations related to the future direction for U-Zr-based metallic fuel research (including binary U-Zr and ternary U-Pu-Zr alloys). These needs and recommendations were determined by subject matter experts from various institutions during a two-day workshop held at the University of Florida in November, 2019. During open-floor discussions, the highest priority gaps in our understanding of U-Zr-based fuels were down-selected, and near- and long-term needs that directly impact the implementation of these metallic fuels were identified. The identified near-term needs include investigation of the following phenomena: i) swelling and fission gas release, ii) fuel-cladding chemical interaction, iii) phase evolution/constituent redistribution, and iv) thermal properties of the fuels. The long-term needs are: i) investigation of fuel creep and plasticity and ii) fission product (lanthanide) transport. In addition, there was general agreement that all institutions and subject matter experts would benefit from an open-source metallic fuels database with thermophysical property and microstructural data, along with fuel operation/irradiation history, which should be regularly updated with vetted information from new experimental and computational investigations and used to advance metallic fuels research and development. Finally, we recommend that metallic fuel research should be ongoing and that fuel qualification and fuel optimization should be equally prioritized; research combining experiments with modeling and simulation has the largest potential impact.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Non-destructive Examinations of ATF-2 Baseline Rodlets

This report contains the results of non-destructive examinations of Accident Tolerant Fuels 2 (ATF-2) rodlets irradiated in the Advanced Test Reactor (ATR) Loop 2A. The experiment is part of the U.S. Department of Energy Nuclear Technology Research and Development (NTRD) program’s Advanced Fuels Campaign (AFC). The rodlets were composed of UO2 pellets and Zr-4 cladding. The data have been collected to provide baseline PIE data to support future testing in TREAT and to have performance data against which the data of the ATF concepts can be compared directly. The analyses performed included: visual examinations, axial gamma scanning, neutron radiography and profilometry. All the data collected showed a performance consistent with expectations for this fuel system at low burnup.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Evaluation of Radiolysis Data for Hydrogen Gas Generation During Gamma Irradiation of Pre-Corroded and Pristine Aluminum Samples - An Aluminum SNF Dry Storage Study Interim Report

Information and data from radiolysis testing to measure hydrogen (H₂) generated from hydrated oxides on aluminum exposed to ⁶⁰Co radiation were reviewed to evaluate hydrogen generation with radiation dose. Hydrogen generation rate is a primary input to the coupled thermal-chemical system model of the performance of aluminum-clad spent nuclear fuel in a dry storage canister (ASNF-in-canister) over its storage life. Hydrogen generation data and hydrogen generation rate (hydrogen generated per absorbed radiation dose, $\Delta$(H₂)/$\Delta$(dose)) were reported in two reports for radiolysis testing [1, 2]. The testing involved first-time data¹ for radiolytic yield of hydrogen from hydrated-oxides-on-aluminum substrates, and from pristine (non-corroded) aluminum substrates. The laboratory-grown hydrated-oxides-on-aluminum substrates contained a ~ 5 μm film consisting of the trihydroxide bayerite (Al(OH)₃) with amounts of the oxyhydroxide boehmite (AlOOH). These specimens were assumed to also contain physisorbed water (unquantified). The pristine substrates were flat coupons of aluminum that were ground to a 600-grit finish but were not immersed in water to grow a hydrated oxide. The pristine substrates were tested as companion specimens to provide information on hydrogen generation from material without chemisorbed water and were assumed to contain physisorbed water (unquantified).

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2020

Risk-Informed Safety Analysis for Accident Tolerant Fuels

Accident Tolerant Fuels (ATF) are being tested by different nuclear vendors and research organization and their introduction in light water reactors fleet is planned for the second half of the 2020`s. In the framework of the US-DOE Light-Water Reactor Sustainability program, Risk-Informed Safety Analysis pathway (LWRS-RISA), research activities are being conducted at the Idaho National Laboratory (INL) for developing tools and methods that can help the industry in quantifying the ATF introduction benefits. In this paper we describe the developed risk-informed methodology, the codes improvements and we present some results for selected accidental conditions. The developed methodology combines INL state-of-the-art deterministic Best Estimate tools like RELAP5-3D code, and Probabilistic Risk Analysis tools like RAVEN and SAPHIRE codes. The analyses are performed on a three-loops pressurized water reactor (PWR) simulating accidental conditions like Station Blackout and LB-LOCA and considering near-term ATF (FeCrAl and Chromium-coated clads). Finally, we show, through our methodologies, how the delta-Core Damage Frequency can be assessed.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Evaluation of Hydrogen Generation in High Burnup Demonstration Dry Storage Cask

This report provides a best-estimate evaluation of residual water content (post-dry out) in the High Burnup (HBU) LWR Spent Fuel Demonstration project TN-32 cask, and evaluates the radiolysis of the residual free water, and the physisorbed and chemisorbed waters on the surfaces of the fuel and cask internal contents. The evaluation of radiolytic breakdown of those waters with gamma radiation causing the generation of hydrogen gas (H 2 ) is made using available literature data and models. This evaluation is part of the overall materials performance evaluation of the SNF-in-canister system, and is part of the technical bases for their continued safe dry storage. The TN-32 cask contents included 32 HBU LWR spent fuel assemblies each with 264 fuel rods clad in zirconium alloys, aluminum neutron absorber components, and aluminum and stainless steel structural components. The residual free and surface (physisorbed/chemisorbed) waters are ascribed to water vapor in the free volume and to components’ surfaces, respectively. The total potential radiolytic hydrogen inventory from the water vapor and from waters ascribed to surfaces has been calculated assuming all the water produced molecular H 2 . The residual water that is chemically incorporated into the bulk of a hydrated oxide, i.e., chemisorbed water, and its total potential hydrogen inventory has been calculated. These calculations are at the physical limit of material available and are used for a bounding assessment purpose only.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Metallic Fuel Benchmark Simulations Based on the X430 Experiments

To gauge the accuracy of current BISON metallic fuel performance models and identify models that require further development, Oak Ridge National Laboratory has developed benchmark problems based on two pins of the X430 experiment. These benchmarks recreate the experimental conditions of the X430 series of experiments with as much accuracy as possible given the available data. The benchmarks were implemented in the BISON nuclear fuel performance code and the simulation results were compared with the experimental results to quantify the simulation accuracy. Findings in this work were consistent with previous benchmark problems. The temperature profiles, burnup, and fission gas behavior were all as accurate as can reasonably be expected. The cladding radial growth was accurate enough to not affect the results of other measurements. The fuel axial growth was accurate for U-Zr fuel but not for U-Pu-Zr fuel, which appears to be caused by a lack of U-Pu-Zr data. Swelling correlations have not been developed for U-Pu-Zr fuel because there were not enough data on this fuel type, so BISON used the U-Zr correlations. Finally, recommendations were made for near-term BISON development.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

A Metallic Fuel Performance Benchmark Problem Based on the IFR-1 Experiment

Metallic nuclear fuels are an active area of research and development for use in advanced reactors. Robust and accurate metallic fuel performance models are necessary for the design, safety analysis, and licensing of such reactors. However, metallic fuel performance models are not as mature as ceramic fuel performance models and require additional development. In order to better gauge the accuracy of existing models, identify high-priority models for development, and quantify the improvements made by future model development, ORNL has developed a benchmark problem based on the IFR-1 experiment. This work gathers all relevant information on the IFR-1 experiment and uses it to develop the benchmark problem. It simulates the problem using the fuel performance code BISON and compare the results to post-irradiation examination data from the experiment. Then a sensitivity analysis is performed on the BISON model to determine the benchmark problem’s sensitivity to uncertainty in the input parameters. The results suggest that BISON’s mechanical models require additional development. Plastic deformation is under-predicted in the cladding and axial swelling is over-predicted in the fuel. These problems may be related. Additionally, there is a bias in the temperature which may be a result of uncertainty in the input parameters rather than an issue with the fuel performance models.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

The Relationship Between Dose Rate and Decay Heat for Spent Nuclear Fuel Casks

Decay heat and dose rate are two important limits used for determining the allowable contents of spent fuel (SNF) in dry storage systems and transportation packages. While the decay heat limit is used to maintain fuel cladding integrity and ensure retrievability, dose rates are used to demonstrate compliance with regulatory requirements on radiation protection. Because both dose rate and decay heat result from decay of radioisotopes in SNF, this study is an attempt to examine the relationship between dose rate and decay heat for a given cask design. Dose rates were evaluated for 198 cask configurations, that include various SNF system designs (e.g., storage, transfer, transport), SNF characteristics (e.g., fuel types, burnup, cooling time), and loading maps (e.g., uniform loading, zone loading), while a constant decay heat was maintained. The decay heat was calculated using US Nuclear Regulatory Commission (NRC) Regulatory Guide (RG) 3.54, Revision 2, and verified using ORIGEN sequence within SCALE code system. The ORIGEN outputs were used as source terms in the dose analysis using 198 different configurations. A computer script was developed to calculate the cooling time necessary to achieve a given decay heat for a given enrichment, assembly average burnup, assembly mass, and in-core history using a rootfinder algorithm. Initially the computer script was developed to provide cooling time and burnup calculations directly to the analysis of dose rates and decay heats, so a comparison between Used Nuclear Fuel-Storage Transportation and Disposal Analysis Resource Data System (UNF-ST&DARDS) results and RG3.54 data was made; results are included in the appendix to this document. However, an iterative approach was used to compute cooling time, and its accuracy did not depend on the results of the RG3.54r2 algorithm, although the algorithm was still used. The results of the evaluation presented herein clearly demonstrate that a given decay heat does not correspond to a unique dose rate for a variety of cask and package designs. There is no clear pattern to develop a correlation between decay heat and the source terms. Depending on burnup, enrichment, cask type, and loading pattern, dose rates varied for the exact same decay heat—in some cases by 400% for a given cask. For cases in which decay heat was held constant through selection of the appropriate cooling time, dose rates would decrease with increasing burnup, and in other cases, dose rates would increase. The large variation in dose rates for a constant decay heat indicates that casks loaded based on decay heat—that is allowing any burnup, cooling time, and enrichment combinations that yield the qualified decay heat limit(s) —cannot ensure that an Independent Spent Fuel Storage Installation or a spent fuel transportation package will meet the regulatory limits set forth by the respective regulations, i.e., 10 CFR 72 or 10 CFR 71.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Augmented Monitoring and Condition Assessment Program (AMCAP) Material Test Reactor (MTR) (Fuel Inspection Program Report)

The AMCAP MTR Fuel Inspection Program, a special inspection program comprised of four inspection campaigns, examined a total of ten (10) pre-selected aluminum-clad, aluminum-based fuel core spent nuclear fuel assemblies (ASNF) stored in the L Area Disassembly Basin (L Basin) at the Savannah River Site. A full description of the inspections and the results are reported. The fuel had been stored in bundled-tube storage (Vertical Tube Storage) for periods of 18 to 21 years. The prior service experience of the individual 10 assemblies varies, but all included irradiation followed by wet storage at international research reactor sites prior to shipment to the US and storage in L Basin. The 10 assemblies were expected to be among the “worst"’ in terms of prior corrosion damage of the entire inventory of the direct-bundle-stored ASNF in L Basin. The inventory of MTR ASNF in L Basin will continue to be stored in the bundled-tube configuration or in slug-storage buckets with inserts pending retrieval for ultimate disposition. The MTR fuel inspections focused on collecting information for characterization of the material condition of the ASNF considering various types of aluminum fuel corrosion degradation of its assembled materials in water storage. A custom-designed Fuel Inspection Table was used to stage the fuel for remote, enhanced visual examination (close-up video imaging & recording) with controlled lighting and positioning that enables reproducibility of imaging conditions. The inspections were conducted by fuel subject matter expert staff from Spent Fuel Project Engineering (SFPE) and the Savannah River National Laboratory (SRNL). Stills captured from the video records were used to compare the corrosion evolution from previous records, as available. This evaluation of the inspection results including the comparison to the previous inspection results demonstrate that the water quality and the storage configuration of ASNF in L Basin do not cause aggressive corrosion degradation of the fuel; mitigation of the prior corrosion degradation of the fuel also appears to have been achieved with the good water quality conditions of L Basin. Recommendations are made for future inspection of the fuel to trend corrosion degradation and demonstrate continued safe wet storage of the ASNF in L Basin. The next fuel examination is recommended to be performed in 5 years.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Status Report on Development of a Cask to Enable Pulsed Neutron Characterization of Irradiated Fuel

We present the design of a shielding cask that allows pulsed neutron characterization of irradiated fuel rodlets prior to their destructive examination in hot cells. The goal is to provide complementary and informative measurements that will inform subsequent hot cell PIE examinations by identifying typical and atypical regions with respect to microstructure, tomographic data, or isotope densities. Time-of-flight neutron diffraction has the potential to offer efficient, non-destructive and non-contact microstructural characterization of irradiated fuel pins with spatial resolution of 1 mm3 to 1 cm3 while for energy-resolved neutron imaging (and by extension tomography) a resolution of 100 μm was demonstrated. The potential results include crystallinity vs. amorphous volumes and microstructural information such as phase compositions and textures from the diffraction data as well as distances (e.g. pellet to cladding), cracks, and isotopic distributions of minor actinides, fission products as well as fission gas partial pressures e.g. in the plenum from energy resolved neutron imaging. The capability will identify regions of atypical behavior and provide cost effective bulk characterization of entire samples prior to destructive evaluation in hot cells. This effort is a collaboration between LANL, INL, ORNL, and UC Berkeley. A key facet of the initiative is the ability to cost effectively ship irradiated fuels from INL to the pulsed spallation neutron source at LANSCE. The irradiated fuel would be loaded in the custom designed cask (designated RaMHaM) at INL hot cells which then could be shipped in a BRR Type B shipment container between INL and LANL. No hot cell is required after the loading of the sample at INL in the pathway laid out in this report, greatly simplifying handling at LANL.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Review of the General-Purpose Heat Source Module Reduction and Monitoring Processing at INL

A thorough review of the Module Reduction and Monitoring (MRM) process was conducted to evaluate the efficacy of processing. The degree of reduction was calculated using the mass balance of oxygen contained in carbon monoxide (CO) and carbon dioxide (CO 2 ) removed from all fuel processed between January 2014 through August 2019. The plutonia was slightly reduced from O/Pu ratio of 2.0 to a range of 1.991 to 1.999. The review evaluated publications generated when MRM processing was developed and performed at the Mound Laboratory. A similar degree of reduction was measured at Mound in the 1982 paper authored by Ernie Johnson as well as MRM processing of IHS-60 and some Cassini-era fuel. The original 1982 paper documenting the reduction process reported a mass balance reduction to 1.9931 yet reported a total reduction to ~1.98 (Johnson 1990). The IHS-60 and Cassini reductions were calculated at 1.995 and 1.997 using the number of moles of CO and CO 2 removed as published in a clad pressurization investigation report (Merten, et all 1995). MRM process changes are recommended to improve the accuracy of the CO partial pressure calcualtion.

42 ENGINEERING↗
DOE OSTI2020

Initial development for the introduction of provisions for irradiation effects in Division 5 rules

Heat pipe microreactor concept involves the use of a solid core block to encapsulate the nuclear fuel and the passive extraction of heat from the core block to the heat exchanger block by advanced heat pipes. The core block is a key component of the heat pipe microreactor design. It integrates the functions of reactor vessel, core structural components and fuel cladding of traditional Gen IV advanced reactor designs into one single structure. The effects of irradiation on the structural failure modes are outside the scope of the rules of the ASME Boiler and Pressure Vessel Code, Section III, Division 5, Subsection HB, Subpart B for metallic Class A components. Methods to account for the irradiation effect were investigated in order to develop design methods that can be applied to the core block design in the heat pipe microreactor. This report presents an initial development of an approach to account for the irradiation effects in the design rules.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

30 cm Drop Modeling

This report documents modeling and analysis of the 30 cm cask drop scenario for spent nuclear fuel (SNF) casks. The 30 cm drop is an important hypothetical dynamic loading scenario for SNF because 10 CFR 71.71 defines testing requirements for normal conditions of transport (NCT) of SNF packages, and free drop (of 30 cm for the heaviest packages) tests are required. Other NCT conditions specified in 10 CFR 71.71 include vibration normally incident to transport, and DOE’s multimodal transportation test (MMTT) and related analysis concluded that mechanical shock and vibration incident to transportation is negligible for SNF structural integrity because the loads are so small. The SNF cask 30 cm drop has a lot more potential to cause damage to SNF because the kinetic energy and momentum of the system is stopped abruptly during the cask impact event. Test data and model predictions both agree that the SNF assemblies are subjected to significantly higher loads in a cask drop event than normal transportation shock and vibration, but currently there is no reason to expect gross structural damage of the fuel assembly or widespread SNF cladding failure as a result of a 30 cm cask drop. The ultimate goal of this work is to close the knowledge gap of stress profiles on SNF rods and fuel assembly components during NCT, and this report describes the progress toward reaching that goal as well as a plan for finishing this work over the next year or two. Key remaining tasks include refining the fuel assembly model, calculating a selection of specific fuel assembly cases to populate the damage model, and finally running the damage model to predict the full general response range of 30 cm drop loads on SNF.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Demonstration of NEAMS Multiphysics Tools for Fast Reactor Applications

The SHARP toolkit is a high-fidelity reactor simulation tool developed under the U.S. Department of Energy, Office of Nuclear Energy Advanced Modeling and Simulation (NEAMS) Campaign. SHARP toolkit is comprised of the neutronics module PROTEUS thermal hydraulics module Nek5000, and structural mechanics module Diablo. During FY17 and FY18, the PROTEUS and Nek5000 components of SHARP were applied to solve challenging sodium-cooled fast reactor (SFR) problems. In particular, selected hot channel factors (HCF) for a prototype metal-fueled SFR design (the AFR-100) were analyzed in high fidelity, and the “SHARP zooming capability” for SFRs was developed and demonstrated to reduce computational expense for full core problems in cases where detailed data is needed in selected fuel assemblies. After the previous success applying SHARP to challenging SFR problems, the focus in FY19 and FY20 expanded to additional fast reactor applications including lead cooled fast reactors (LFR) and sodium cooled fast reactors (SFR). The specific technical tasks were (1) assessment of hot channel factors for LFR, for which no data currently exists, and (2) demonstration of zooming capability in assemblies of the Versatile Test Reactor (VTR). First-of-a-kind hot channel factor (HCF) estimation for LFR with high fidelity codes (PROTEUS/Nek5000) was successfully demonstrated in this study which began in FY19 and continued in FY20. Selected HCF were computed and compared with SFR data (AFR-100, EBR-II). The findings confirm that different reactor types, design parameters and uncertainties lead to different HCFs. Careful estimation of HCF for a specific design is necessary to obtain appropriate HCFs. In addition to improvement in HCF accuracy, high fidelity tools generate data to help the designer better understand the mechanism of the impact from these uncertainties. For example, the impact of cladding thickness manufacturing tolerance resulted in non-intuitive effects in the corner pins of the LFR assembly. This procedure of computing HCF using high fidelity models shows promise and flexibility for being repeated for any arbitrary reactor of choice. Along with the application on SFR and LFR, the capability of the tools has also been matured to deal with different reactor types and designs. Progress was made towards extending the previously demonstrated SHARP zooming capability to non-fueled SFR assemblies. In particular, in FY19 a gamma transport capability was implemented in both high fidelity PROTEUS solvers in order to accurately account for heat deposition caused by gamma particles, which accounts for ~10% of total core power. Neutronics verification cases were carried out for a candidate Versatile Test Reactor (VTR) design using the new gamma transport capability in PROTEUS. Comparisons were made with continuous energy MCNP calculations and shown to agree well. The models for the full core design with heterogeneous control and fuel assemblies is in progress for PROTEUS-SN and completed with MCNP. The MCNP power distributions were transferred to Nek5000 to perform thermal hydraulic calculations of the control and fuel assembly.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗