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All sourcesNASA NTRSDOE OSTI
DOE OSTI2019

Neutronic Design and Analysis of the Holos-Quad Concept

HolosGen, LLC is proposing a highly innovative micro-reactor concept targeting both civilian and military applications. It consists of an advanced helium-cooled gas reactor using a turbojet-type turbine and compressor to achieve a highly condensed reactor that fits into a commercial ISO container. The Holos-Quad concept being considered for civilian applications will generate 22MWt, using four Subcritical Power Modules (SPMs) that fit into one 40-foot ISO container. It is a high-temperature gas-cooled reactor concept using TRISO fuel distributed in graphite hexagonal blocks, cooled with helium gas at 7MPa and a high outlet temperature. In FY2019, HolosGen, LLC received an ARPA-E MEITNER grant to team up with the Argonne National Laboratory (ANL) design team to demonstrate the feasibility of the Holos-Quad concept in terms of neutronics. In particular, the evaluation study of the Holos design features and alternative options for reactivity control was considered. Through these tasks, the ANL design team has been working on demonstrating the feasibility of the Holos-Quad concept in terms of neutronics to help advance the core design of the Holos-Quad. A rigorous design approach was employed for the Holos-Quad in order to ensure that the input space is fully investigated, and that the best solutions are considered. The first step of this approach consisted of properly defining the Holos-Quad design problem by identifying the competitive design objectives: its economic performance is maximized with longer lifetime and its transportability to remote locations is maximized with reduced weight. The design problem also required to identify the operational constraints (in terms of thermal hydraulics, reactivity feedback and shutdown margins) and the input parameters that could be varied. In the second step, a sensitivity analysis was performed to enable preliminary investigation of the input space to identify the correlations among input and output variables. It allowed, for instance, to find input parameters that have relatively poor impact on the core performance and that could be removed from the optimization problem to accelerate its convergence. This sensitivity analysis showed that the design problem defined is highly constrained since only 117 solutions out of 1,000 simulated cores met all the design constraints. The design optimization was performed in the third step employing a genetic algorithm to effectively explore this highly constrained input space and find global optimal solutions. More than 2,500 cores were fully evaluated to converge to a Pareto Frontier of optimum solutions providing best compromise between the life-time target and the total core weight, as shown in Figure 1. The main benefit of this approach was to reduce the human effort and to be able to automatically re-run the optimization problem when changing the design problem. The multi-objective optimization identified various core solutions that would be optimum solutions for different types of applications.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

Transient Temperature Safety Testing of AGR-2 UCO Compacts 5-1-1, 5-1-2, and 5-3-1

Post-irradiation examination (PIE) and elevated-temperature safety testing are being performed on tristructural-isotropic (TRISO) coated-particle fuel compacts from the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program’s second irradiation experiment (AGR-2). Previous reports provide background information on (1) the fuel fabrication (Barnes and Marshall 2009; Hunn, Montgomery, and Pappano 2010a; Hunn, Montgomery, and Pappano 2010b), (2) the as-fabricated fuel properties (Hunn 2010; Hunn, Savage, and Silva 2010), (3) the irradiation experiment (Collin 2014), and (4) the comprehensive PIE plan (Demkowicz 2013).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Development and Demonstration of an In-Situ Tritium Scavenger

The objective of this project was to use an inherent feature of salt-cooled reactors towards the capture and removal of tritium from the primary coolant system: the presence of the graphite moderator. The fuel elements fluoride-salt cooled high-temperature (FHR) reactors consist of a graphite matrix that encapsulates TRISO fuel particles, and the total fuel element surface area of Mark I PB-FHR is 7780 m 2 . Therefore the FHR fuel inherently provides removable graphite of high surface area. The goal of this project was to understand the transport of tritium across the salt-graphite interface. The following was achieved: 1. Electrochemical techniques were demonstrated for studying the entry, diffusion, and trapping of hydrogen isotopes in graphite. 2. Electrochemical techniques were used to demonstrate the relationship between galvanic corrosion between graphite and stainless steel 316 and hydrogen uptake in graphite. 3. Electrochemical techniques were demonstrated for studying the dissolution and diffusion of hydrogen isotopes in molten FLiBe. 4. Thermodynamic reference electrodes were tested for FLiBe and FLiNaK. 5. Electrochemical techniques were demonstrated as an in-salt hydrogen thermal desorption measurement, which had previously only been achieved in gas environments. 6. FLiBe exposure of graphite lead to graphite fluorination and an increase of defects, which indicate an increase of reactive carbon sites available for tritium uptake. 7. TRIDENT modeling was used to perform sensitivity analysis of FHR design parameters on tritium uptake fraction in graphite, showing high sensitivity to tritium diffusivity and solubility. 8. A cross-cutting effort with another NEUP-funded project was initiated for application of the SAM mass-transport module to system-scale tritium modeling.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Characteristics of Fuel Articles for Irradiation Testing in INL-TREAT and MIT-NRL

To obtain valuable data that will help in the design and commissioning of the Transformational Challenge Reactor, a series of graded, increasingly complex irradiation tests are planned that will demonstrate the feasibility of advanced manufactured reactor core components. In the first irradiation test, conventionally- fabricated uranium nitride tristructural isotropic (UN TRISO) fuel kernels will be embedded into compacts that are fabricated using advanced manufacturing methods. These compacts are to be irradiated at Massachusetts Institute of Technology’s Nuclear Reactor Laboratory and Idaho National Laboratory’s Transient Reactor Test Facility. This report presents the characteristics of the proposed samples and experiments.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

Oxidation of Matrix Material in Helium with Varied Moisture Content

Exploration of the oxidation behavior of matrix material from tristructural-isotropic (TRISO) coated particle fuel elements in moisture environments provides critical data needed to predict the material’s response during off-normal accident scenarios. Near-representative, matrix-only samples were fabricated to facilitate oxidation testing in varied moisture concentrations. Oxidation testing was completed over a range of temperatures and oxidizing environments to obtain the relevant oxidation kinetics parameters and empirical oxidation rates necessary to better predict and understand matrix oxidation behavior at relevant conditions. The acquired oxidations rates were fit to established models for steam oxidation of nuclear graphite, and data were acquired following an established process. The data showed good agreement over conditions typical of the kinetic regime, and they showed a departure from the kinetic model fit at elevated temperatures and steam partial pressures in helium carrier gas. The departure was confirmed to be due to the experimental conditions being beyond the kinetic regime based on destructive analysis.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

FY20 Verification of BISON Using Analytic and Manufactured Solutions

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, Consortium for Advanced Simulation of LWRs (CASL), has developed an LWR simulation tool called Virtual Environment for Reactor Applications (VERA). The multi-physics capability of VERA is achieved through the coupling of single-physics codes, including BISON, CTF, MPACT, and MAMBA. BISON is a fuel performance code which models the thermo-mechanical behavior of nuclear fuel using high performance M&S. It is capable of modeling traditional LWR fuel rods, fuel plates, and TRi-structural ISOtropic (TRISO) fuel particles. It can employ three-dimensional Cartesian, two-dimensional axisymmetric cylindrical, or one-dimensional radial spherical geometry. It includes empirical models for a large variety of fuel physics: temperature- and burnup-dependent thermal properties, fuel swelling and densification, fission gas production, cladding creep, fracture, cladding plasticity, and gap/plenum models. This document details a series of code verification test problems that are used to test BISON. These problems add confidence that the BISON code is a faithful representation of its underlying mathematical model. The suite of verification tests are mapped to the underlying conservation equations solved by the code: heat conduction, mechanics, and species conservation. Twenty-two problems are added for the heat conduction solution, two for the mechanics solution, and none for species conservation. Method of Manufactured Solutions (MMS) capability is demonstrated with three problems, and temperature drops across the fuel gap are tested.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Monolithic ATF MiniFuel Sample Capsules Ready for HFIR Insertion

Several advanced nuclear fuel concepts are being considered throughout the nuclear industry. Implementation of these concepts could increase accident tolerance or enhance performance beyond the capabilities of the UO₂-based nuclear fuel currently in use. Qualification and deployment of any new fuel requires rigorous irradiation testing to demonstrate performance under representative normal and off-normal operating conditions. The traditional approach for qualifying new fuels requires exhaustive execution of many integral fuel tests. However, due to the long timeframe for executing these integral tests and the limited number of available materials test reactors, this approach is becoming impractical. To accelerate fuel qualification, Oak Ridge National Laboratory developed the MiniFuel irradiation vehicle for use in conducting accelerated separate effects irradiation testing of a wide range of nuclear fuel materials in the High Flux Isotope Reactor (HFIR). The first MiniFuel irradiations performed in the facility tested sol gel–derived uranium nitride kernels and tristructural isotropic (TRISO)-coated particle fuels. This report describes the preparation and assembly of the first set of monolithic MiniFuel irradiations conducted to support accident-tolerant fuel (ATF) development and accelerated fuel qualification. Two irradiation targets containing a variety of UO₂ and U₃Si₂ disk fuel specimens were fabricated and assembled for irradiation to burnups of 8-10 and 28-40 MWd/kg U. The target irradiation temperature is 450-550°C. Irradiation of U₃Si₂ will provide new data regarding the irradiation performance of a candidate ATF to complement current ATF-1 integral experiments being performed in the Advanced Test Reactor. UO₂ samples were included as a reference so that the results from the MiniFuel experiments can be compared with the extensive UO₂ fuel performance database. The monolithic MiniFuel capsules were successfully assembled, welded, and tested per HFIR requirements and are ready for insertion into the reactor. Pictures of the assembly process are included in this report. The experiment is planned for insertion into the HFIR during cycle 487 in April 2020.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Evaluation of Microreactor Inhalation Dose Consequences

The new generation of microreactors are intended to provide useful, very reliable and safe safe energy for a variety of applications such as supplying to remote locations. Microreactors range from less than 1 MW th to more than 20 MW th . A range of technologies is being considered, including variations in neutron energy spectrum (fast or thermal), variations in primary and secondary coolant choices (e.g. metal cooled, gas cooled, and unique power cycle reactors. The reactors tend to use new fuel forms and types with unique benefits such as metal or TRISO fuel for higher fuel density and safety, respectively. The notional image of a microreactor is a reactor that fits in a seavan or intermodal shipping container. This would allow conventional shipping options, including aircraft, and match other mobile power sources. A development of a capability to test these reactors is needed, whether in a facility or in an open area. INL is in the process of determining acceptable test areas for these various reactors and an understanding of potential bounding consequences is important in informing siting and required infrastructure capability decisions. This ECAR considers the bounding release fractions of these types of reactors at 20 MW th and documents the dose consequences to the public and collocated worker. This is a scoping analysis for the generalized type of reactor, and the analysis will be refined based on the specific reactor design.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

Compilation of Data from ORNL Characterization of German Reference Fuel Composite EUO 2358-2365

This document was prepared to document the distribution of additional data summarized in a previous report entitled "Results from ORNL Characterization of German Reference Fuel from the EUO 2358-2365 Composite," ORNL/TM-2005/546 (also known as ORNL/CF-04/06), April 2004. This previous report was a compilation of the characterization data obtained on a sample of TRISO-coated 500 µm diameter UO2 produced in Germany and obtained by the AGR program for use as a historical reference material. This sample came from the EUO 2358-2365 composite studied by General Atomics (GA) and referenced in GA document #910852 “Acceptance Test report for German Fuel Particles.” The ORNL designation for the material characterized was AGR-06. This update includes the individual particle data from measurements of kernel diameter, particle diameter, and coating thickness. This individual particle data was not included in the previous report, which only contained the summary data from the image analysis.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

AGR-1 Irradiated Compact 6-1-1 PIE Report: Evaluation of As-Irradiated Fuel Performance Using Leach Burn Leach, IMGA, Materialography, and X-ray Tomography

The AGR-1 irradiation test train consisted of six independently monitored capsules. All capsules contained twelve compacts in three stacks of four compacts each [Pope 2010]. Compact 6-1-1 was located in capsule 6, at the top of the test train. The compact was in stack 1 (one of the two stacks closest to the core) in position 1 (at the bottom of the 4 compact stack). Compact 6-1-1 was identified in the fabrication data package as LEU01-46T-Z32 [Hunn et al. 2006]. Details of the compact fabrication and measured properties can be found in that data package. The compact lot was made using TRISO-coated particles designated as AGR-1 Baseline fuel. These particles contained 19.7% enriched mixed uranium oxideuranium carbide (UCO) kernels with a O/U atomic ratio of ~1.36 and a C/U atomic ratio of ~0.33 [Pope 2010]. Details of the coated particle fabrication and measured properties can be found in the data package for the AGR-1 Baseline particles [Hunn and Lowden 2006]. The compact was irradiated at the INL Advanced Test Reactor (ATR) to an average calculated burn-up of 15.1% fissions per initial metal atom (FIMA) and an average calculated fast fluence of 3.0 x 10 25 n/m 2 [Sterbentz 2011] at a calculated volume averaged and time averaged temperature of approximately 1111°C [Hawkes 2012]. Capsules were disassembled and preliminary compact and graphite holder inspection was performed [Demkowicz et al. 2011]. Compact 6-1-1 was shipped to ORNL and post-irradiation examination (PIE) was performed according to examination plan PLN-3624 [Demkowicz 2010-2]. All the examination plan objectives listed below were completed except item 6, which was deleted due to limited funding. 1. Determine the inventory of fission products retained in the compact matrix, but not contained by the SiC and determine the number of exposed kernels or particles with defective SiC layers using the deconsolidation-leach-burn-leach method. 2. Gamma count all particles to quantify 137Cs inventory for comparison with calculated values and gamma spectrometric measurements made on the whole compact. Also, attempt to quantify the inventory of any other fission products for which gamma emissions are observed. 3. Characterize each particle based on the 137 Cs/ 144 Ce ratio to identify particles with low Cs retention that may be the result of a failed SiC coating. 4. Gamma count a subset of particles to determine the 110m Ag inventory for comparison with calculated values and gamma spectrometric measurements made on the whole compact. Also, attempt to quantify the inventory of other fission products observed. 5. Microscopically inspect selected particles to determine the condition of the kernels and coatings. This could include kernel migration, coating microstructures and defects, fission product attack of the SiC layer, and fission product migration in the coating layers. 6. Perform high temperature fission product release tests on a subset of particles in order to calibrate the collection efficiency of the Core Conduction Cooldown Test Facility (CCCTF) deposition cup for selected fission products. Gamma count particles before and after the heating test to quantify the released fission product fractions. 7. Archive remaining particles for later use. Compact 6-1-1 LBL began with compact deconsolidation in the Building 3525 hot cell, followed by two nitric acid leaches of the particles and matrix residue using a Soxhlet extraction apparatus. The matrix residue was then burned and the ash was leached twice with boiling nitric acid. The deconsolidated particles underwent an IMGA analysis to determine gamma-emitting isotope inventories for individual particles; the most conspicuous elements with reduced inventories, indicating fission product release, were silver and europium. During the IMGA analysis, particles with low silver were identified and a subset of these was set aside for materialography and x-ray. After the completion of the IMGA study, about 10% of the particles (421) were set aside as an archive and the remainder burned back and leached twice in the Soxhlet apparatus. Figure 1 shows a flow chart detailing the PIE of Compact 6-1-1.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Advanced Moderation Module for High-Temperature Micro-Reactor Applications

Development and deployment of micro-reactors that provide competitive efficiency, prevailing compactness, and inherent safety is an immediate focus of the U.S. nuclear industry. For thermal neutron micro-reactors operating at elevated temperature for optimized efficiency, such as micro molten-salt reactors (MSRs), heat-pipe reactors, and very-high temperature reactors (VHTRs), high-performance moderator based on metal hydride can enhance the neutron economy and therefore achieve reduced weight and enhanced portability. As unclad metal hydride inevitably decomposes at elevated temperature, an enclosure is required for hydride moderator to deliver desired performance at elevated temperatures. Conventional moderator enclosure solutions based on high-temperature alloys introduce extraneous neutron penalty into the micro-reactor, affecting the neutronic benefits provided by the hydride moderator. Additionally, the compatibility between the high-temperature alloy enclosure and the micro-reactor matrix is also a potential issue. In this report, we disclose an Advanced Moderator Module (AMM) concept enabled by an innovative enclosure solution combining the advantages of refractory metals, ceramic matrix composites (CMCs), and advanced coating technology to serve as hydrogen permeation barrier up to very-high temperatures. A schematic description of the AMM structure is illustrated in Figure A1. The AMM contains a moderating material core made of metal hydride with high thermal stability, such as YH 2-x . The hydride core is enclosed by an H 2 barrier layer coated on a ductile refractory metal liner to minimize hydrogen loss during high-temperature operation. A ceramic matrix composite (CMC) cladding is adopted to provide further structural strength, especially during power transients. Between the CMC cladding and metal liner, an extra diffusion barrier coating is inserted to suppress the chemical interaction at elevated temperatures. Hence, based on a series of innovative material solutions, the AMM is capable of containing the metal hydride core at elevated temperature (>900oC) inside coated and lined CMC envelop with negligible hydrogen loss throughout the microreactor lifetime. The benefit of the AMM technology was assessed based on a comprehensive multi-stage reactor-physics analysis. Advanced moderation based on hydride metals (such as YH2-x) enables reaching optimum moderation with higher fuel content than traditional VHTR technology, which is required to design compact micro-reactor cores targeting long-life operation. The AMM enclosure technology provides lower thermal neutron absorption rates than traditional solutions (based on stainless steel or Mo-based alloy such as TZM), which reduces the fissile enrichment requirements by 6-8% on a TRISO-fueled design based on the EMPIRE core. Finally, combining the hydride moderator with neutron transparent enclosure solutions provides significant potential to boost neutronics performance of micro-reactors in terms of increased core lifetime or reduced size and weight by 30-50% on a micro-reactor based on the Holos Quad technology. The progress and plans for on-going development and demonstration efforts are also discussed in this report. The current demonstration of AMM is focused on the hydrogen diffusion barrier demonstration under thermal cycling, while radiation tolerance demonstration is planned. The next step of the demonstration plan will be the assembly of miniature AMM for high-temperature testing, which can be used as a prototype for future pilot scale demonstration.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

BISON BWR Fuel Modeling Capability: Material Models (FY2020)

Until this year the BISON fuel performance code has been developed primarily for pressurized water reactor (PWR), accident tolerant fuel (ATF), and advanced reactor (e.g., tri-structural isotropic (TRISO), metallic) analyses. The main focus of this work is to develop and document required material models that would aid predicting cladding failure with a specific hydride distribution to be investigated for boiling water reactor (BWR) applications in the next fiscal year. In terms of materials, the primary differences between a BWR and PWR is the use of gadolinia doping of the UO 2 fuel, the use of Zircaloy-2 for the cladding, and the addition of a liner on the inner surface of the cladding for improved pellet-clad mechanical interaction (PCMI)performance. The liner is typically made of pure zirconium or a lower tin content zirconium alloy (e.g. Zr-0.3%Sn). This work details the development and documentation of material and behavioral models for BWR analysis such as the addition of material and behavior models for gadolinia doped UO 2 , pure zirconium, Zr-0.3%Sn, and Zircaloy-2 as well as the ease-of-use additions to allow internal meshing capabilities to include liners. Another important aspect affecting cladding performance is hydrogen diffusion and hydride precipitation in the cladding. A model was added previously for non-lined, homogeneous claddings. In this work, instructions are provided on how to apply the model to BWR liner cladding. Representative experimental values are provided for the hydrogen diffusion parameters and precipitation-dissolution kinetics for typical liner materials of interest.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Assessment of Existing Transportation Packages for Use with HALEU

Commercial light water reactor operators and fuel vendors in the United States are pursuing changes to fuel that include increased 235 U enrichment. Economic studies generally anticipate maximum near-term fuel assembly designs with up to 8 wt.% 235 U. Many next-generation nuclear reactor designs require high-assay low-enriched uranium (HALEU) (19.75 wt.% > 235 U > 5 wt.%) fuel. One necessary element for the commercial-scale use of HALEU is the ability to safely transport large quantities of enriched fuel material in multiple forms. However, there is uncertainty as to whether subcriticality requirements can be satisfied with existing package designs and whether existing critical benchmark experiment data are sufficient to support criticality safety code validation for HALEU transportation applications. This study assesses the potential to use currently licensed transportation packages for the transportation of increased enrichment unirradiated U fuel forms. The assessment uses selected package designs that represent the five categories of fuel form-boiling water reactor pins and assemblies, pressurized water reactor pins and assemblies, UF 6 , U-metal and tristructural isotropic (TRISO) particles, and UO 2 pellets or powder-and focuses on demonstrating subcriticality and identifying benchmark critical experiments appropriate for use in criticality computer code validation. Key quantities of interest that relate to subcriticality are limiting conditions (e.g., optimum moderation), package or package array k eff , package capacity, and package transportation array size. The SCALE TSUNAMI-3D and TSUNAMI Indices and Parameters codes are used for sensitivity and uncertainty calculations and for identification of candidate critical benchmark experiments for code validation. The primary metric for identifying candidate benchmarks is the similarity coefficient, c k . For each fuel form category, a representative package is evaluated. Results provided for each package evaluation include enrichment and packaging limits (e.g., maximum transportation array size as a function of enrichment) and benchmark critical experiment similarity coefficients. Results indicate that there are viable means for increasing enrichments into the HALEU range across the spectrum of fuel forms with differing increase amounts available for different packages. Sources of subcriticality margin to offset increased enrichment reactivity include reduced transportation array size, reduced fissile mass, burnable absorber credit, and safety analysis margin harvesting. For all packages except the DN-30, numerous critical benchmark experiment candidates for validation were identified.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Evaluation of Binder Jetting and Chemical Vapor Infiltration as a Production Method for Silicon Carbide Parts for Nuclear Applications

Application of silicon carbide (SiC) is often sought within nuclear energy systems due to its high thermal conductivity, mechanical and environmental stability at elevated temperatures, corrosion resistance, and exceptional displacement damage irradiation tolerance. As such, the transformational challenge reactor (TCR) program is considering its application in its additively manufactured core. SiC may be used as the matrix for the TCR fuel blocks that comprise of TRISO fuel particles embedded inside this material or used as core non-fuel-bearing peripherical structures such as bottom support manifolds. The core peripherical structures serves to support the overall core geometry, direct and guide coolant flow, and also reflect back leaked neutrons into the core. SiC has very low neutron absorption and adequate scattering cross sections that makes it an ideal material for utilization in the TCR core. Like many ceramics, however, SiC is difficult and at times impractical to shape into complex geometries via traditional machining due to its brittleness. Further, large structures comprising nuclear-grade SiC, i.e. highly pure and crystalline, are costly. Additive manufacturing (AM), however, offers the ability to shape materials by adding layer-by-layer rather than subtracting, contributing new possibilities for previously difficult-to-shape materials. Thus, exploring AM as a method to shape SiC offers a highly cost-effective method and is strategic to developing the next generation of components for nuclear reactors. This report outlines an evaluation of binder jet additive manufacturing combined with chemical vapor infiltration (CVI) to produce nearly-dense silicon carbide artifacts in complex geometries suitable for advanced nuclear reactor designs. The focus is placed on the binder jet printing technologies at large scale for nonfuel-bearing structures relevant to the TCR core and other advanced reactors. Among all AM processes, binder jet delivers the highest resolution capability while still maintaining large print sizes, making it the ideal candidate for nuclear component production.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

ECAR-3873 Acceptance of Nonconforming Fuel for the AGR-5/6/7 Irradiation Experiments

Low-enriched fuel compacts fabricated for the AGR-5/6/7 irradiation experiments did not conform to all fuel specifications. Reasons why the fuel is acceptable for use in the AGR-5/6/7 irradiation are discussed. Revision 1 alters the characterization data for the dispersed uranium and exposed kernel defect fractions. Additional characterization data was obtained for fuel compacts analyzed by Oak Ridge National Laboratory. This enabled statistical outliers to be rejected and to confirm that results obtained from BWX Technologies Nuclear Operations Group were statistically independent of the Oak Ridge results and should be excluded from the analyses. Revision 2 alters the characterization data for the TRISO lot, J52R-16-98005, used in the fabrication of the AGR-5/6/7 fuel. Data from the BWXT J52R-16-98005 Certification package is used to calculate the reported values.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Cross Section Generation Capability in Griffin

The Griffin code is a Multiphysics Object-Oriented Simulation Environment (MOOSE) based reactor multiphysics analysis application jointly developed by Idaho National Laboratory and Argonne National Laboratory. The code includes a variety of steady-state solvers for fixed-source, k-eigenvalue, adjoint, and subcritical multiplication, as well as transient solvers for point-kinetics, improved quasi-static, and spatial dynamics. The code reads multigroup cross sections in the ISOXML format generated from external deterministic or Monte Carlo cross section generation codes. The implementation of the cross section generation capability in Griffin was initiated last year by plugging in the cross section application programming interface (CSAPI) and reviewing the methodologies for treating particulate fuels. The focus this year was on improving the CSAPI integration and implementing advanced self-shielding methods for applications to advanced reactor problems with TRISO fuels. First, the process for cross section library generation was updated to accurately and rigorously produce isotopic cross section data. Second, the equivalent Dancoff factor cell method performing slowing down calculations on the fly for the resonance treatment was implemented in CSAPI to improve the accuracy of effective multigroup cross sections in the resonance energy range. Third, the iterative local spatial self-shielding method was implemented under the calculation framework of the equivalent Dancoff factor cell method to accurately deal with the double heterogeneity effect of particulate fuel. The updated CSAPI with the advanced self-shielding methods, together with the cross section libraries generated based on the improved process, were tested for pin-cell, unit-cell, and fuel assembly problems with various resonance self-shielding conditions based on very high temperature reactor, high temperature test reactor, and Empire benchmark cores, indicating that the updated CSAPI in Griffin is able to produce multigroup cross sections accurately and efficiently. We also showed that the methodology worked well for pebble bed fuel from HTR-10, but the capability still needs to be fully integrated into CSAPI. In the future, further benchmark tests will be performed for various thermal reactor core problems, including particulate fuel-based pebble bed reactors.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2021

ECAR-5021 Source Term Estimates for SCO Micro-Reactor Designs

The purpose of this document is to provide source term estimates and the associated technical basis for Strategic Capabilities Office (SCO) microreactor designs that use TRISO fuel. This estimate can be used for relevant environmental and safety analyses that will be done as part of the project for all of the selected suppliers for a safety design strategy. This is a combined quantitative and qualitative analysis.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

Griffin Software Development Plan

Griffin is a MOOSE-based reactor physics application for advanced reactor multiphysics modeling and simulation. The application is developed in a consistent multiphysics environment with strong software quality assurance. Griffin inherited most of the capabilities of MAMMOTH/Rattlesnake and is adopting the capabilities from PROTEUS that are needed in the code. The toolset includes a variety of deterministic radiation transport solvers for fixed source, k-eigenvalue, ad-joint, and subcritical multiplication, as well as transient solvers for point-kinetics, improved quasi-static, and spatial dynamics. The code contains the cross-section preparation capabilities applicable to fast and thermal reactors, including TRISO-fueled reactors. Core management capabilities include core performance, fuel depletion and shuffling, equilibrium core calculation, pebble-bed reactor run-in and equilibrium core, molten-salt reactor delayed neutron precursor drift, and control rod and drum movement with cusping correction. This software development plan presents the current and future capabilities and features in Griffin for the design and analysis of non-light-water reactor systems in steady-state and transient conditions.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗