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DOE OSTI2015

Executing BISON-CASL Fuel Performance Cases Using VERA-CS Output

As the Consortium for Advanced Simulation of Light Water Reactors (CASL) moves forward with more complex multiphysics simulations, there is increased focus on incorporating fuel performance analysis methods. The coupled neutronics/thermal-hydraulics capabilities within the Virtual Environment for Reactor Applications Core Simulator (VERA-CS) have become relatively stable and major advances have been made in analysis efforts, including the simulations of twelve cycles of Watts Bar Nuclear Unit 1 (WBN1) operation. However, VERA-CS approaches for treating fuel pin heat transfer have well-known limitations that could be eliminated through better integration with the BISON-CASL fuel performance code. Several approaches are being taken to improve or replace the VERA-CS fuel models and improve integration with BISON-CASL. Tiamat is being developed to replace the fuel models with a direct coupling of BISON-CASL and the neutronics and coolant heat transfer of VERA-CS. Tiamat will execute in an inline mode that allows for in-memory transfer of VERA-CS data to BISON-CASL. This will provide fuel performance results during core simulation. However, fuel performance typically undergoes an independent analysis using a stand-alone fuel performance code with manually specified input defined from an independent core simulator solution or set of assumptions. This milestone covers initial efforts to facilitate using VERA for core simulation and fuel performance to operate in this mode of execution. This approach will be used for several CASL fuel performance challenges, such as RIAs and DNB. These issues require time-dependent full core results to establish the boundary conditions for the high-fidelity BISON-CASL simulations. The approach will also be used to improve fuel temperature models in VERA-CS, guide the development of Tiamat, and simplify the use of BISON-CASL within VERA for industry users.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Pros and Cons Analysis of HALEU Utilization in Example Fuel Cycles

The Systems Analysis and Integration campaign assessed the pros and cons of high-assay low-enriched uranium (HALEU) utilization in advanced reactors and associated fuel cycles. The assessment was done for three example fuel cycles at equilibrium states: once-through, limited recycle, and continuous recycle (CR) starting with HALEU. Front- and back-end fuel cycle parameters and the Levelized Cost of Fuel (LCF), which is the Levelized Cost of Electricity excluding reactor cost, of the three example fuel cycles were calculated using a single Analysis Example Reactor. The pros and cons of HALEU utilization were assessed by normalizing the fuel cycle parameters and LCF to a unit of electricity generation (GWe-year) and comparing them with a Basis of Comparison. In this study, a sodium-cooled reactor with sodium-bonded metallic fuel having a burnup of ~100 GWd/t was used as the Analysis Example Reactor because its technology readiness level is high, and the burnup and fuel enrichment are in the middle of those ranges of advanced reactor concepts that are under development. The current once-through Light Water Reactors (OT-LWRs) with <5% low-enriched uranium and 50 GWd/t burnup were used as the Basis of Comparison. In addition, a series of sensitivity analyses was conducted by varying burnup, enrichment, fuel forms, and reactor types to capture the design variations in two once-through Advanced Reactor Demonstration Program (ARDP) reactors, Natrium with sodium-free metallic fuel having a burnup of ~150 GW/t and Xe-100 with Tristructural-Isotropic (TRISO) pebble fuel having a burnup of ~168 GWd/t.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

SCALE Demonstration for Sodium-Cooled Fast Reactor Fuel Cycle Analysis

In support of the US Nuclear Regulatory Commission non-light-water reactor fuel cycle demonstration project, SCALE 6.3.1 capabilities for radionuclide characterization, criticality, and shielding were demonstrated for scenarios in the sodium-cooled fast reactor (SFR) nuclear fuel cycle. Three postulated accident scenarios were selected for analysis in this work. As a basis for all scenarios, irradiated fuel inventories were generated using SCALE/ORIGAMI. To cover multiple SFR design choices, two different types of SFR fuel, uranium/transuranic-loaded and U-based fuels, were considered. For the first scenario, SCALE/MAVRIC was used to calculate the radiation shielding and dose rates inside and outside of the containment building due to a drop of a spent fuel assembly from the fuel-handling system during unloading inside the containment building. For the second scenario, potential critical configurations in an electrofiner were investigated through criticality calculations with SCALE/CSAS. For the third scenario, the activity of the waste salt from an electrorefiner was evaluated using SCALE/ORIGEN. The dose rate produced by the analyzed SFR assemblies is similar to that produced by a typical pressurized water reactor (PWR) fuel assembly with a discharge burnup of 50 GWd/MTU, with the same cooling time of 10 days. The criticality analyses suggested that the different electrorefiner configurations have a large margin to criticality. The activity analysis of the electrofiner waste revealed that shielding and cooling may be required for the waste salt that contains transuranics and fission products produced by the electrorefiner because of the high activity of the waste salt. In general, the application of various capabilities in the SCALE code system for SFR fuel inventory generation, criticality, and shielding was successfully demonstrated for the selected scenarios in the SFR nuclear fuel cycle. Additional analyses can be performed to provide more accurate results when more details of the SFR nuclear fuel cycles are available, for example, the dimension of the electrorefiner and the salt compositions during reprocessing.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Nuclear Material Accounting and Control (NMAC) for a Commercial Fuel Debundling Facility in the United States

The concept of fuel debundling proposes that all non-fuel components of a fuel assembly be removed prior to beginning the recycling process. It was introduced to reduce the quantity of and cost of disposing high level radioactive waste during recycling in the United States. However, nuclear material accountancy and control programs could also benefit from fuel debundling due to a reduction in self-shielding material for individual fuel rods compared to a fuel assembly. Current NRC regulations and guidance point to an independent fuel debundling facility with a material-possession-only license as the best option for implementing this concept into a used fuel recycling scheme in the United States. This allows the facility to comply only with simple NMAC requirements under Subpart B of 10 CFR Part 74 without the need to submit an NMAC plan to the NRC, exempting it from more tedious and complex NMAC requirements. However, a recycling facility may desire a Pu mass estimate from measurements of each individual fuel rod to improve accountancy into the dissolver. Many technologies and methodologies were considered for performing NMAC in an independent fuel debundling facility in the United States. Few were able to meet anticipated throughput demands of up to 3,300 rods/day and achieve measurement uncertainties suitable for input accountancy for a recycling facility. The best option was a combination of passive gamma-ray spectroscopy and neutron coincidence counting to meet all requirements for the proposed Fuel Debundling Detector (FUDD). Follow-on work will include a full uncertainty analysis to inform what assumptions can be made to meet input accountancy objectives before beginning experimental activities.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Assessment of Possible Cycle Lengths for Fully Encapsulated Microstructure fueled light water reactor Concepts

The use of TRISO-particle-based dispersion fuel within SiC matrix and cladding materials has the potential to allow the design of extremely safe LWRs with failure-proof fuel. This paper examines the feasibility of LWR-like cycle length for such fuel with the imposed constraint of strictly retaining the original geometry of the fuel pins and assemblies. The motivation for retaining the original geometry is to provide the ability to incorporate the fuel “as-is” into existing LWRs while retaining their thermal–hydraulic characteristics. Another mandatory constraint is use of low enriched uranium (at or below 20 w/o). The feasibility of using this fuel is assessed by looking at two factors: cycle lengths and fuel material failure rates. Other considerations (e.g., safety parameters such as reactivity coefficients, feedback, etc.) were not considered at this stage of the study. The study includes the examination of increases in the TRISO kernel sizes without changing the thickness of any of the coating layers. In addition, cases where the buffer layer thickness is allowed to vary are also considered. The study shows that a naïve use of UO2 (even up to 20 w/o enrichment) results in cycle lengths too short to be practical for existing LWR designs and operational demands. Increasing fissile inventory within the fuel compacts shows that acceptable cycle lengths can be achieved. The increase of fissile inventory can be accomplished through multiple means, including higher particle packing fraction, higher enrichment, larger fuel kernel sizes, and the use of higher density fuels (that contain a higher number of U atoms per unit volume). In this study, starting with the recognized highest packing fraction practically achievable (44%), combinations of the other means have been evaluated. The models demonstrate cycle lengths comparable to those of ordinary LWRs. As expected, TRISO particles with extremely large kernels are shown to fail under all considered scenarios. In contrast, the designs that do not depart too drastically from those of the nominal NGNP HTR fuel TRISO particles are shown to perform satisfactorily and display a high rates of survival under all considered scenarios.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Transition Analysis of Promising U.S. Future Fuel Cycles Using ORION

The US Department of Energy Office of Fuel Cycle Technologies performed an evaluation and screening (E&S) study of nuclear fuel cycle options to help prioritize future research and development decisions. Previous work for this E&S study focused on establishing equilibrium conditions for analysis examples of 40 nuclear fuel cycle evaluation groups (EGs) and evaluating their performance according to a set of 22 standardized metrics. Following the E&S study, additional studies are being conducted to assess transitioning from the current US fuel cycle to future fuel cycle options identified by the E&S study as being most promising. These studies help inform decisions on how to effectively achieve full transition, estimate the length of time needed to undergo transition from the current fuel cycle, and evaluate performance of nuclear systems and facilities in place during the transition. These studies also help identify any barriers to achieve transition. Oak Ridge National Laboratory (ORNL) Fuel Cycle Options Campaign team used ORION to analyze the transition pathway from the existing US nuclear fuel cycle—the once-through use of low-enriched-uranium (LEU) fuel in thermal-spectrum light water reactors (LWRs) —to a new fuel cycle with continuous recycling of plutonium and uranium in sodium fast reactors (SFRs). This paper discusses the analysis of the transition from an LWR to an SFR fleet using ORION, highlights the role of lifetime extensions of existing LWRs to aid transition, and discusses how a slight delay in SFR deployment can actually reduce the time to achieve an equilibrium fuel cycle.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Uncertainty Quantification and Sensitivity Analysis Applications to Fuel Performance Modeling

Best-estimate fuel performance codes such as BISON currently under development at the Idaho National Laboratory, utilize empirical and mechanistic lower-length-scale informed correlations to predict fuel behavior under normal operating and accident reactor conditions. Traditionally, best-estimate results are presented using the correlations with no quantification of the uncertainty in the output metrics of interest. However, there are associated uncertainties in the input parameters and correlations used to determine the behavior of the fuel and cladding under irradiation. Therefore, it is important to perform uncertainty quantification and include confidence bounds on the output metrics that take into account the uncertainties in the inputs. In addition, sensitivity analyses can be performed to determine which input parameters have the greatest influence on the outputs. In this paper we couple the BISON fuel performance code to the DAKOTA uncertainty analysis software to analyze a representative fuel performance problem. The case studied in this paper is based upon rod 1 from the IFA-432 integral experiment performed at the Halden Reactor in Norway. The rodlet is representative of a BWR fuel rod. The input parameters uncertainties are broken into three separate categories including boundary condition uncertainties (e.g., power, coolant flow rate), manufacturing uncertainties (e.g., pellet diameter, cladding thickness), and model uncertainties (e.g., fuel thermal conductivity, fuel swelling). Utilizing DAKOTA, a variety of statistical analysis techniques are applied to quantify the uncertainty and sensitivity of the output metrics of interest. Specifically, we demonstrate the use of sampling methods, polynomial chaos expansions, surrogate models, and variance-based decomposition. The output metrics investigated in this study are the fuel centerline temperature, cladding surface temperature, fission gas released, and fuel rod diameter. The results highlight the importance of quantifying the uncertainty and sensitivity in fuel performance modeling predictions and the need for additional research into improving the material models that are currently available.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Status of LEU U-10Mo Monolithic Fuel Testing and Qualification Efforts to Support US High Performance Research Reactor Conversions

The US High Performance Research Reactor Project has begun a new phase of fuel testing and post irradiation examination (PIE) of plate-type low-enriched uranium (LEU) U-10Mo Monolithic fuel. The testing campaigns are intended to first confirm that the fuel performance behavior of commercially produced fuel matches that of the lab-scale fabricated fuel tested during development, followed by qualification of the U-10Mo monolithic fuel system through the US Nuclear Regulatory Commission. Planned tests include miniature-plate, full-size plate, and fuel element irradiations, during which fuel test specimens fabricated with qualified processes will be irradiated under conditions that bound the NRC-regulated reactors in order to demonstrate that the fuel meets the overarching fuel performance requirements of 1) maintaining mechanical integrity 2) maintaining geometric stability and 3) exhibiting stable and predictable behavior. The MP-1 test, with the objective of confirming fuel performance behavior, is currently undergoing irradiation in the Advanced Test Reactor and PIE in INL hot cells. Status of upcoming qualification and licensing test design activities and highlights of initial as-fabricated characterization and testing results from the MP-1 test will be provided.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Tracking radioactivity content in the spent fuel pool in PWRS - 15150

Over the last decade, significant progress has been made in managing radioisotopes generated in the PWR primary systems. These include limiting the use of cobalt containing materials, along with improved fuel performance and zinc injection to passivate the corrosion layers to limit the accumulation of Co-58 and Co-60 on primary coolant surfaces. Attention here is directed to the spent fuel pool where accumulation of spent fuel and activated material represents a growing source of mobile radioactivity supplemented with periodic exchanges of water from the reactor cooling system. Characterization of this source is increasingly important to effective management of disposal options. This paper examines the radioactive resin source term as determined from routine sampling of the coolant and fuel pool liquids. Targeted resin stream sampling and radiochemical analyses conducted at Diablo Canyon serve to inform and provide context for this examination. This examination is supplemented with computer modelling using the DW James Consulting computer program, 3R-SCAN. At most operating plants daily reactor coolant samples are counted using gamma spectroscopy and recorded. These samples serve to track fuel performance and maintain water chemistry. A secondary use for this data is to maintain a day to day tracking of releases of radioisotopes important to waste disposal. Given the coolant concentrations, plant operating history, and corroborative mathematical models and sampling, accumulation of these radioisotopes in coolant cleanup system ion-exchangers can be predicted. Fuel pool sampling is less frequent, but at the same time, things change more slowly there. The fuel pool contains about ten times the volume of the reactor coolant but its cleanup system process rate is comparable to reactor coolant letdown, i.e. 5.047 - 8.833 liters per second [lps] (80-140 gallons per minute [gpm]). A spent fuel pool clean-up demineralizer flow rate is typically about 5.047 lps (80 gpm). Concentrations of longer lived radioisotopes in the spent fuel pool can be comparable to those in reactor coolant. Releases of corrosion products and fission products from the stored fuel occur at a low rate but given the overall inventory can be significant. Recognized scaling factor ratios, particularly those to Co-60, are skewed by the aging of the material in the pool. An additional consideration stems from refueling operations themselves. Prior to refueling, it is common practice to conduct a forced oxidation of the corrosion films within the reactor. This spikes the activity within the reactor coolant. Most of this activity is removed by the reactor coolant purification system prior to removing the reactor head. This paper examines the accumulation of activity in the fuel pool purification ion exchange resins and how it impacts the overall classification of those ion exchange resins. The following discussion focuses on the transport of nickel and cobalt isotopes particularly as it relates to determination of Ni-63 which has become increasingly prominent in disposal classification. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Development of the Licensing Procedure and Regulatory Framework for the Spent Fuel Storage facility in Korea - 16423

The Public Engagement Commission on SNF management (PECOS) published the recommendation report on the spent fuel management of Korea in June, 2015. Based on this report, the future national policy for spent fuel management of Korea has been established, taking into consideration the national/international trends on policy and technology development. Regardless of the recommendation report of PECOS, Spent fuel storage casks are necessary in Korea. Therefore, various types of spent fuel storage casks are now being developed by Korea vendors. For example, Concrete storage casks and dual purpose metal casks are now under development by the KORAD (Korea Radioactive waste agency) for future use in the interim storage facility. Unfortunately, there is no independent licensing procedure for spent fuel storage casks in Nuclear Safety Act (NSA) because is the casks are considered an integral part of the interim storage facility safety equipment. Furthermore, the definition of radioactive waste in the NSA was changed recently. Now, spent fuel is separated from the definition of 'radioactive waste'. Spent fuel can only become radioactive waste by the decision through the meeting of the relative ministers of Korea government. For that reason, the licensing procedure for interim spent fuel storage facilities was separated from the current 'disposal facility, etc.' in accordance with the article 63 of NSA. The aim of this study is to develop the licensing procedure for the storage cask, and also to develop a revised draft of interim storage facility procedure within the NSA. Once this independent licensing procedure on the spent fuel storage cask is introduced to the NSA, it is expected that developers or operators can develop and commercialize the storage cask in Korea. In this study, an independent licensing procedure for the spent fuel storage facility was developed including the storage cask licensing procedure for the NSA. The procedure was composed of the permit for construction and operation of interim storage facility, criteria for permit, inspection criteria, periodic safety review, and decommissioning criteria, etc. The detailed contents of these requirements were also developed to introduce the regulatory framework for the interim storage facility. Particularly, periodic safety review and decommissioning criteria were developed in accordance with the recommendation of IRRS (International Regulatory Review Service) of IAEA. The developed licensing procedure for spent fuel storage facility will be introduced in NSA in response to the revised definition of NSA and the recommendation report of PECOS. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Challenges of SLOWPOKE-2 HEU Irradiated Fuel Transportation from Canada to USA - 19123(2)

The SLOWPOKE-2 (Safe Low-Power Kritical Experiment) was designed by AECL (Atomic Energy of Canada Limited) in the late 1960's. The core consists of HEU (highly enriched uranium) fuel with an enrichment of 93%. The fuel cage typically has approximately 300 fuel pins assembled in a triangular pitch between the two plates. The fuel pins are made from uranium-aluminum alloy cladded with aluminum. The SLOWPOKE-2 reactor is rated at a nominal 20 kW. with varying neutron flux, maximum being 1 E12 n/cm{sup 2}/s. The mass of the fuel cage is 5.06 kg. The life of the core can be extended by adding layers of beryllium reflector. The transportation flask for the irradiated fuel is a Type B(U)F. The shipment classification is UN3328 Class 7. To date, 4 SLOWPOKE-2 reactors have been decommissioned. The fuel from all these reactors was transported to the U.S. The CNSC (Canadian Nuclear Safety Commission) joint Canada-United States guide for approval of Type B(U) and fissile material transportation packages (RD-364) satisfies the requirements stipulated in International Atomic energy Agency (IAEA) TS-R-1 1996 edition (revised) and is acceptable to the DOT (Department of Transportation) and NRC (Nuclear Regulatory Commission) for complying with U.S. code of federal regulations 10 CFR Part 71. The export license is issued by the CNSC. The transportation licence is issued by CNSC after approval from TC (Transport Canada) and also by DOT after approval from NRC. The final home of the HEU SLOWPOKE-2 irradiated fuel is in Savannah River Site of DOE (Department of Energy). The transportation licence requires that the Type B(U)F flask is structurally sound in case of a crash. The US DOT provides a CAC (Competent Authority Certificate) which describes the irradiated fuel specifications that can be transported in the designated flask. The U.S. DOT also requires a leak test certification in accordance with ANSI N14.5-2014 to verify a leakage rate not to exceed 1 E-7 std-cm{sup 3}/s. As most SLOWPOKE-2 facilities are not located on the existing approved routes for nuclear fuel transportation, a route survey needs to be performed and submitted to the NRC. The route survey report is assessed by the NRC and issues a route approval for the transportation of the fissile material. This technical paper will discuss the detailed processes required to obtain approvals in order to transport the HEU SLOWPOKE-2 irradiated fuel from Canada to U.S. based upon recent experiences of decommissioning SLOWPOKE-2 reactors in Canada. It will provide details on the requirements of various regulators in Canada and the U.S. that govern and regulate this shipment including transportation, safety and security. It will also present lessons learned from previous experiences that can be utilized for best practice on other similar projects. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

Development of Instrumented Advanced Test Reactor Irradiation Capsule Experiment for In-situ Thermal Conductivity Measurements of High-Density Fuels

Idaho National Laboratory (INL) is developing a first-of-a-kind leadout instrumented capsule experiment design to enable in-situ measurement capabilities in the Advanced Test Reactor (ATR) core. The Ceramic Advanced Thermal Evolution Research (CRATER) experiment supports the aLEU program objective to accelerate fuel performance irradiation testing for identifying alternative high-assay, low enriched uranium (HALEU) fuel systems. CRATER is a fueled, instrumented capsule experiment to measure in-situ temperature and thermal conductivity of ceramic fuels. Two ceramic fuel types will be used, uranium mono-nitride (UN) and uranium mono-carbide (UC), with a third metallic fuel used for comparison (UMo). The three fuel specimens will use a stainless-steel cladding. Programmatic objectives include linear heat generation rates (LHGR) of 210 ± 25 Watts per cm. and an inner clad temperature of 300-450 °C. The evolution of fuel thermal conductivity during irradiation has never been successfully measured in-situ for these systems and this experiment is designed to use advances in measurement sciences to characterize how thermal transport properties evolve while in reactor. Neutronic simulations of the experiment and its surrounding reactor environment were conducted using the Monte Carlo N-Particle Transport code (MCNP) and result in optimized fuel enrichment to meet target linear heat generation rates (LHGRs) influencing fuel temperatures, and fuel burnup requirements. Fabrication research and development (R&D) efforts are underway to produce annular right cylinder UC and UN pellets using carbothermic reduction and nitridation (or hydride-dehydride-nitride) synthesis methods, followed double-action die cold isostatic pressing.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Analysis of transient fission gas behaviour in oxide fuel using BISON and TRANSURANUS

The modelling of fission gas behaviour is a crucial aspect of nuclear fuel analysis in view of the related effects on the thermo-mechanical performance of the fuel rod, which can be particularly significant during transients. Experimental observations indicate that substantial fission gas release (FGR) can occur on a small time scale during transients (burst release). To accurately reproduce the rapid kinetics of burst release in fuel performance calculations, a model that accounts for non-diffusional mechanisms such as fuel micro-cracking is needed. In this work, we present and assess a model for transient fission gas behaviour in oxide fuel, which is applied as an extension of diffusion-based models to allow for the burst release effect. The concept and governing equations of the model are presented, and the effect of the newly introduced parameters is evaluated through an analytic sensitivity analysis. Then, the model is assessed for application to integral fuel rod analysis. The approach that we take for model assessment involves implementation in two structurally different fuel performance codes, namely, BISON (multi-dimensional finite element code) and TRANSURANUS (1.5D semi-analytic code). The model is validated against 19 Light Water Reactor fuel rod irradiation experiments from the OECD/NEA IFPE (International Fuel Performance Experiments) database, all of which are simulated with both codes. The results point out an improvement in both the qualitative representation of the FGR kinetics and the quantitative predictions of integral fuel rod FGR, relative to the canonical, purely diffusion-based models, with both codes. The overall quantitative improvement of the FGR predictions in the two codes is comparable. Furthermore, calculated radial profiles of xenon concentration are investigated and compared to experimental data, demonstrating the representation of the underlying mechanisms of burst release by the new model.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

MOOGLE: A Multi-Objective Optimization tool for three-dimensional nuclear fuel assembly design

MOOGLE is a new genetic algorithm methodology for the three-dimensional design of nuclear fuel assemblies. MOOGLE uses common fuel rod types as the decision variable to develop a suite of three-dimensional fuel assemblies to provide optimized solutions to the design problem. Pressurized Water Reactor (PWR) fuel assemblies were optimized using IFBA and gadolinium (Gd 2 O 3 ) as burnable poisons in order to compare how burnable poison choice affects optimization results. Boiling Water Reactor (BWR) fuel bundles were also optimized using three unique fuel rod palettes to study how the size of the design space affects optimization results. Burnable poison analysis showed that utilizing IFBA and Gd 2 O 3 as burnable poisons produced the best and widest range of optimized solutions. Further, BWR fuel bundle optimization results indicate that the inclusion of additional fuel rod types produced a wider solution space but did not improve optimization results for regions explored using fewer unique fuel rods. These tests demonstrate MOOGLE's ability to analyze the tradeoffs between the inclusion of different fuel elements and their effects on assembly performance.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

A review of neutronics and thermal hydraulics–based screening methods applied to accelerated nuclear fuel qualification

This paper reviews the state-of-the-art engineering approach for using thermal hydraulic (TH) and neutronics modeling and simulation (M&S) tools to perform rapid screening studies of novel nuclear fuel concepts within the context of accelerated fuel qualification. Global research efforts have introduced nuclear fuel and material concepts that mark a significant departure from traditional reactor materials. The number of new technologies being considered for development for light water reactors and advanced reactor types has created the need for an accelerated fuel qualification procedure. A key component of this procedure is the rapid identification of the most promising fuel concepts using computational screening studies. Advanced TH and neutronic M&S tools should be leveraged to efficiently determine whether the reactor performance and safety characteristics of a given concept warrant additional studies or whether the concept requires modification or elimination. This paper reviews best practices for performing these TH and neutronics screening studies at various stages during a fuel concept's progression through the qualification procedure. The motivation behind standardizing this approach is to minimize time and resources spent on qualification activities for fuel concepts that could be quickly refined or eliminated from consideration based on their reactor physics and TH characteristics. Adoption of this screening procedure—which focuses primarily on nuclear fuels but may be applicable to other reactor materials—will also help accelerate new material qualification by generating boundary conditions crucial to fuel performance evaluations and highlighting needed areas of separate effects experimentation. This article reviews the motivation behind the introduction of novel nuclear fuel concepts, provides incentive for utilizing TH- and neutronics-based screening studies, describes the screening approach and methodology, and includes discussion on how to interpret screening results to provide recommendations for the continued development of a given concept.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Summary and Assessment of Metal Fuel Capabilities in Bison

The work described in this report was performed under funding from the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program. This report is issued in satisfaction of the level 2 milestone M2MS-18IN0201012 on Release Bison with Initial Metallic Fuel Performance Capability. Metallic fuel development began in Bison a few years after light water reactor development began (~ 2009). The initial development was funded by the Advanced Fuels Campaign. With recent interest from industry, the Nuclear Regulatory Commission, and the Versatile Test Reactor program, the Nuclear Energy Modeling Simulation Program has increased funding for metallic fuel development and validation in FY18. The material models with the most mature development are binary uranium-zirconium and ternary uranium-plutonium-zirconium alloys for fuel and HT9 for cladding. The first material models were incorporated from open literature sources, which include mechanical and thermal material properties that are functions of temperature, porosity, and zirconium concentration, swelling, fission gas release, zirconium diffusion, creep, and sodium coolant channel boundary conditions. These material models have been utilized to simulate ERB-II fuel pins and the results compared to measurements. The EBR-II experiment measurements mostly consist of cladding strain and zirconium redistribution. As such, comparisons to EBR-II measurements are currently limited to these two figures of merit. Simulations and comparisons to experiment measurements have also been done for transient testing of EBR- II fuel pins in TREAT, which included temperature measurements. Preliminary comparisons between Bison calculations and experiment measurements are favorable. These comparisons have also highlighted the importance of the fuel swelling and fission gas release models on cladding strain. Also, Bison metallic fuel simulations are currently being run in support of ATR experiments and VTR exploratory design calculations. Further development and evaluation of individual material models and fuel system models are planned for FY19.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

AGR-3/4 TRISO Fuel Compact Ceramography

The combined third and fourth irradiation in the Advanced Gas Reactor (AGR) program (AGR-3/4) contained tristructural isotropic (TRISO)-coated particle fuel and designed-to-fail (DTF) fuel particles. The DTF particles were only coated with highly anisotropic pyrocarbon (PyC) so they would purposely fail during the AGR-3/4 irradiation and provide a source of fission products for measurement. To observe the post-irradiation morphology of these DTF particles and the TRISO-coated particles, three AGR-3/4 compacts were mounted in epoxy, sectioned above their centerlines, and ground/polished. Three rounds of grinding/polishing and optical microscopy were performed so that the particles could be observed at multiple planes. Each compact contained approximately 1,872 TRISO particles and exactly 20 DTF particles. The three AGR-3/4 compacts examined covered a wide temperature range and featured both the hottest average irradiation temperature (1375°C) and the coldest/lowest burnup (872°C and 5.5% fissions per initial metal atom) of any compacts to undergo post-irradiation examination (PIE) in the AGR program to date. A total of 29 DTF particles were located and observed via microscopy across the three compacts. All the DTF particles observed via microscopy had completely failed PyC coatings. Some different DTF kernel and PyC morphologies were observed that have been attributed to the differences in irradiation temperature and/or burnup. At low and medium irradiation temperatures (roughly 850 to 1050°C), it appears irradiation-induced dimensional changes in the DTF PyC caused it to fracture and fold in on itself, and the kernel deformed to accommodate this PyC deformation. The extent of DTF kernel deformation to accommodate DTF PyC buckling tended to be greater for the medium-temperature fuel compared to the cold fuel. In the high-irradiation-temperature compact (1375°C), the DTF PyC layer appears to have completely reacted chemically with the kernel material. The only material surrounding the DTF particles in this hot compact resembles the compact matrix material. Observations of the TRISO-coated driver fuel particles and fuel compact graphitic matrix were also made. The TRISO particle kernels in the medium-temperature compact (1047°C), and many in the high-temperature compact (1375°C), showed morphologies consistent with what was seen in AGR-1 and AGR-2. In other TRISO particles from the high-temperature compact, a spatial gradient in the kernel appearance was evident. Here the cool side of the kernel (away from the center of the compact) had a much darker appearance (like that of the buffer and pyrocarbon). This is the first observation of this kind of kernel spatial gradient in AGR fuel. It is believed that the high irradiation temperature of that compact (an average of 1375°C) activates or accelerates an unidentified chemical reaction, and the temperature gradient within the fuel causes a sharp spatial gradient. The low-temperature/low-burnup compact also showed unique kernel morphologies where the TRISO-coated particles had clearly distinguishable oxide and carbide phases at the center of the kernel, an oxide rind surrounding this, and remnants of a carbide skin surrounding the oxide rind. These are features observed in the as-fabricated fuel that are no longer present in higher burnup fuel. Occasional gaps between the outer pyrolytic carbon (OPyC) and graphitic matrix material were observed in all compacts. Gaps were most often found in the small, matrix-filled spaces between adjacent particles. Finally, buffer morphologies in AGR-3/4 TRISO particles were like those observed in AGR-1 and AGR-2, and the buffer fracture frequencies in AGR-3/4 and AGR-2 TRISO particles were plotted versus fast neutron fluence and irradiation temperature. In AGR-3/4 the buffer fracture rate was highest (23%) for an irradiation temperature of 1047°C and a fast fluence of 5.18E25 n/m 2 . Increasing the irradiation temperature to 1375°C reduced the buffer fracture rate to 14%. Reducing the irradiation temperature to 872°C and the fast fluence by a factor of 3 reduced the buffer fracture rate to 7%. The temperature and fluence dependencies observed for AGR-3/4 buffer fracture were consistent with those observed in AGR-2. Lower fluences and/or higher temperatures significantly reduced the buffer fracture frequencies. Lower buffer fracture frequencies were also observed at lower temperatures as long as the fluence was also significantly reduced. This is believed to be due to lower fluence resulting in less irradiation-induced dimensional change (shrinkage) of the buffer, and a higher temperature promoting enhanced creep relaxation of stresses within the buffer, leading to less buffer fracture.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Assessment of Alternative Fueling Infrastructure in the United States

NHTSA uses the Corporate Average Fuel Economy (CAFE) Model to analyze potential CAFE standards and their impact on emissions and vehicle fleet composition. The CAFE model analyzes the application of potential technologies to the current automotive industry vehicle fleet to determine the feasibility of future CAFE standards and the associated costs and benefits of the standards. A significant portion of the engineering input development is related to the effectiveness (energy consumption reduction) of each fuel-saving technology and the combination of several fuel-saving technologies, including AFVs. The purpose of this report is to deepen NHTSA's understanding of alternative fueling infrastructure and its potential impact on the adoption of alternative fuel vehicles (AFVs) so that AFVs can be more accurately and comprehensively incorporated into the CAFE Model. This report analyzes the current state of alternative fueling infrastructure in the United States and its relationship to the light-, medium-, and heavy-duty AFV markets; explores the costs associated with alternative fueling infrastructure; investigates trends driving the deployment of alternative fueling infrastructure; explores how the adoption of various vehicle and fuel technologies may look in the future; and analyzes the evolution of alternative fueling corridors.

33 ADVANCED PROPULSION SYSTEMS↗