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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↗
DOE OSTI2016

An Innovative Accident Tolerant LWR Fuel Rod Design Based on Uranium-Molybdenum Metal Alloy

The US Department of Energy is developing a uranium-molybdenum metal alloy Enhanced Accident Tolerant Fuel concept for Light Water Reactor applications that provides improved fuel performance during normal operation, anticipated operational occurrences, and postulated accidents. The high initial uranium atom density, the high thermal conductivity, and a low heat capacity permit a U-Mo-based fuel assembly to meet important design and safety requirements. These attributes also result in a fuel design that can satisfy increased fuel utilization demands and allow for improved accident tolerance in LWRs. This paper summarizes the results obtained from the on-going activities to; 1) evaluate the impact of the U-10wt%Mo thermal properties on operational and accident safety margins, 2) produce a triple extrusion of stainless steel cladding/niobium liner/U-10Mo fuel rod specimen and 3) test the high temperature water corrosion of rodlet samples containing a drilled hole in the cladding. Characterization of the cladding and liner thickness uniformity, microstructural features of the U-Mo gamma phase, and the metallurgical bond between the component materials will be presented. The results from corrosion testing will be discussed which yield insights into the resistance to attack by water ingress during high temperature water exposure for the triple extruded samples containing a drilled hole. These preliminary evaluations find that the U-10Mo fuel design concept has many beneficial features that can meet or improve conventional LWR fuel performance requirements under normal operation, AOOs, and postulated accidents. The viability of a deployable U-Mo fuel design hinges on demonstrating that fabrication processes and alloying additions can produce acceptable irradiation stability during normal operation and accident conditions and controlled metal-water reaction rates in the unlikely event of a cladding perforation. In the area of enhanced accident tolerance, a key objective is to establish that the lower stored energy of the U-Mo fuel design can provide the emergency core cooling systems the opportunity to maintain the reactor core in a coolable geometry following an accident.

Accident tolerance fuel↗
DOE OSTI2015

EVALUATION OF ENHANCED ACCIDENT TOLERANT FUELS

The U.S. Department of Energy Advanced Fuels Campaign (AFC) is currently funding research and development on a number of fuel technologies for light water reactors (LWRs) that could provide enhanced performance under accident scenarios relative to the current zirconium alloy-uranium dioxide (UO2) fuel system. The overall mission of the accident tolerant fuel (ATF) research is to develop advanced fuels/cladding with improved performance, reliability and safety characteristics during normal operations and accident conditions, while minimizing waste generation. Evaluating the performance of candidate accident tolerant fuel concepts is a topic of significant discussion among researchers, fuel vendors and utilities to ensure that the developed concepts can meet minimum performance and economic requirements, can be developed within a reasonable time, and can be approved for insertion as a lead fuel rod in a commercial reactor by the 2022 goal. Characterizing the performance enhancements of candidate ATF first requires understanding the performance of the current Zr-UO2 system under equivalent operations and accident scenarios. Proper evaluation of each concept is dependent on development of data through focused out-of-pile and in-core experiments to support modeling of the fuel and cladding behavior in fuel performance and systems analysis codes; complimentary irradiation of ATF concepts will be discussed in a separate paper. This paper will provide an overview of the key evaluation tools and evaluation scenarios currently being considered within the LWR community for ATF. The evaluation toolset includes standard neutronic and thermal-hydraulic analysis for normal operating conditions and transient/accident conditions and analysis of severe accident behavior using a modified version of the MELCOR code. Development of advanced fuel performance analysis using the BISON application based on the Multi-physics Object-Oriented Simulation Environment (MOOSE) environment at Idaho National Laboratory (INL) has also been initiated. Evaluation scenarios (e.g. accident scenarios) currently being discussed for use across the international ATF development teams are also presented.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

U-10Mo Monolithic Fuel Qualification Plan

The Material Management and Minimization (M3) Program’s primary objective within the U.S. Department of Energy/National Nuclear Security Administration is to achieve permanent threat reduction by minimizing and when possible, eliminating weapon-usable nuclear material around the world. Under M3’s Office of Conversion, the U.S. High Performance Research Reactor (USHPRR) Project, is pursuing a fuel qualification and licensing effort focused on converting high-performance research reactors in the United States from using highly enriched uranium (HEU) fuel to using low-enriched uranium (LEU) fuel. The body of this document is focused on defining the activities required for generic qualification of high density, LEU fuel that then allows reactor conversions to proceed for the four high-performance research reactors and the critical assembly that will operate using the U-10Mo monolithic fuel design (see Subsection 1.3): MITR, MURR, NBSR, ATR, and ATRC. Fuel qualification in this report means generic fuel acceptance of the U-10Mo monolithic fuel form by the NRC for use in reactor conversions in the United States, which allows the fuel to be used in subsequent reactor specific licensing requests. The U 10Mo Monolithic Fuel Research, Development, and Qualification Plan utilizes the functions and requirements of the USHPRR Project that was established in the Functions and Requirements Document (F&RD) [1] and expands on these requirements to ensure that planned tests have traceable results that will ensure the requirement has been met. The methods by which data will be collected to show that these requirements have been met are described in this document.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Enhanced CANDU6: Reactor and fuel cycle options - Natural uranium and beyond

The Enhanced CANDU 6{sup R} (ECo{sup R}) is the updated version of the well established CANDU 6 family of units incorporating improved safety characteristics designed to meet or exceed Generation III nuclear power plant expectations. The EC6 retains the excellent neutron economy and fuel cycle flexibility that are inherent in the CANDU reactor design. The reference design is based on natural uranium fuel, but the EC6 is also able to utilize additional fuel options, including the use of Recovered Uranium (RU) and Thorium based fuels, without requiring major hardware upgrades to the existing control and safety systems. This paper outlines the major changes in the EC6 core design from the existing C6 design that significantly enhance the safety characteristics and operating efficiency of the reactor. The use of RU fuel as a transparent replacement fuel for the standard 37-el NU fuel, and several RU based advanced fuel designs that give significant improvements in fuel burnup and inherent safety characteristics are also discussed in the paper. In addition, the suitability of the EC6 to use MOX and related Pu-based fuels will also be discussed. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Nuclear fuel behavior at high temperature: new insights from post-test examinations on the VERDON-1 sample

In order to reduce the uncertainties concerning Source Term assessment under Light Water Reactor (LWR) severe accident conditions, 4 VERDON tests were recently performed. They have been defined in the framework of the International Source Term Program and carried out at CEA Cadarache, in the LECA-STAR facility. Among these 4 VERDON tests, VERDON-1 test was devoted to high burn-up UO{sub 2} fuel behavior and FP release under H{sub 2} reducing conditions at very high temperature. This test was performed with the Release Circuit, which is devoted to the precise characterization of the FP release, in terms of FP release kinetics and fission products (FPs) released fraction. The main components of the circuits are a furnace heating the fuel sample, a filter trapping the emitted aerosols and gamma spectrometry detectors for online and post-test measurements. VERDON-1 test was conducted up to about 2600 Celsius degrees without reaching global fuel relocation, allowing thereby post-test micro-analyses. After a brief description of the fuel sample nature and the experimental set-up, we present the progress of the experimental sequence and the associated gamma spectrometry measurements. We then focus on the major results of the VERDON-1 tests regarding the fuel global behavior. These results are obtained by two complementary techniques. First, the fuel sight gamma station can be used either on-line to detect fuel relocation or just after the test, as a longitudinal bench, to get preliminary information on the global shape of the fuel sample inside the furnace. Secondly, these information are completed by post-test examinations performed on the fuel sample after extraction from the furnace. Complete qualitative and quantitative characterizations, including EPMA (Electron Probe Micro-Analysis) and SEM (Scanning Electron Microscopy) analysis, were performed for the first time at the CEA Cadarache on a sample that has been submitted to such severe accident conditions. These results show an advanced state of sample degradation, with locally previously molten areas, meaning that fuel global relocation was imminent by the end of thermal-hydraulic sequence. The relocation temperature is compared to that of previous VERCORS tests and discussed regarding the atmosphere of the test and possible UO{sub 2}-ZrO{sub 2} interaction. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Selected Profiles of Burnup and Isotopic Concentrations in Pebble Bed High Temperature Reactor Fuel

Pebble Bed High Temperature Reactors (PB-HTRs) feature a unique fuel loading, burnup, and discharge process that is markedly different that a light water reactor or any other batch-loaded reactor that uses fixed fuel elements. During PB-HTR operation, small spherical fuel elements are dropped onto the top of the bed at small and regular intervals, 'trickle' through the core, and pass out of the core via a discharge chute at the bottom. Partially burnt pebbles are circulated back to the top of the core for another pass through the flux field. The process continues until the pebble achieves a specified discharge burnup. This results in high fuel utilization relative to a comparable batch-loaded fuel assembly which may or may not achieve full burnup depending upon where it is placed in the core. On the other hand, the radial and axial variation in the flux field and the random placement of each element on top of the bed mean that there will be a distribution of the final burnup and isotopic concentrations in spent fuel. This leads to some uncertainty in burnup analysis that has consequences for burnup monitoring (and thus efficient operation), source term quantification, and material accountability. The PEBBED code was developed specifically for the design and analysis of PB-HTRs. Although it is still under development, the code captures the important physics of cores fueled by dropping in, and if so configured, recirculating small graphite spheres in which the Tri-Isotropic (TRISO) coated particle fuel is embedded. In this study, the capabilities of the code for computing the isotopic contents of PBR fuel were exploited to examine and compare the variability of selected isotopes, in this case U-235, Pu-239, and Cs-137, in the fuel discharged from different PBR concepts. Results indicate the sensitivity of certain isotopes to the core geometry, mainly the presence of an inner reflector. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

A Preliminary Overview of Metal Fuel Models in BISON

BISON is a modern nuclear fuel performance code that has been under development at Idaho National Laboratory (INL) since 2009. It is a finite element-based code capable of solving the energy, mechanics, and species diffusion equations in a fully-coupled sense. BISON also utilizes thermal and mechanical contact algorithms and variety of nonlinear material models. BISON is applicable to steady and transient fuel behavior and supports the use of one-, two-, and three-dimensional meshes. Although thus far used primarily to simulate light water reactor fuel, BISON is designed to be a multi-fuels code. As such, BISON has been used to analyze TRISO-coated particle fuel and metal fuel in rod and plate form, design and interpret fuel irradiation experiments and investigate novel fuel concepts. This publication provides a preliminary list of metal fuel-related material models recently implemented in BISON. Some preliminary results from a simulation of the EBR-II X447 experiment are shown, which begins to demonstrate how all these models work together. A follow-up publication will be presented with more detail at the ANS meeting in New Orleans during the embedded topical meeting titled 'Nuclear Fuels and Structural Materials for the Next Generation Nuclear Reactors'. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Thermal Hydraulics Safety Analyses of TRIGA Reactor Upon Fuel Reconfiguration

Safety analyses of the Training, Research, Isotope, General Atomics (TRIGA) reactor at Kansas State University(KSU) is performed using thermal hydraulic simulations based on changes to reactivity or the volumetric heat generation rate due to changes in the positioning of the enriched fuel elements in the reactor core. If the reactivity or neutronics parameters are known, then they can be used to simulate the thermal energy state of the material which is crucial to assess the safety of the nuclear fuel inside a nuclear reactor. Several thermal hydraulics models and codes have been developed in the past to assess reactor safety. RELAP (Reactor Excursion and Leak Analysis Program) is one such tool widely used throughout the world for safety analyses. Although RELAP is considered a reliable tool by regulators, vendors, and operators, it lacks the numerical fidelity provided by 3-D computational fluid dynamics (CFD) models. This work involves the critical task of assessing safety of a research reactor upon fuel reconfiguration based on models developed in RELAP and comparing them with a commercially available 3-D CFD program. The TRIGA Mark-II reactor, designed by General Atomics and located at Kansas State University, is a class of small pool type inherently safe research reactor licensed to operate at steady state power of up to 1.25 MW. It has a cylindrical core measuring 0.45 meters diametrically and 0.38 meters axially positioned in a tank measuring 1.98 meters diametrically and 6.25 meters axially. It has uranium zirconium hydride (UZrH) fuel rods housed in stainless steel cladding material arranged in five concentric rings (B,C,D,E and F rings, in that order, moving outwards from the center). TRIGA fuel is nominally 20% enriched, but the reactivity of the fuel can be changed by using burnable poisons (not used at KSU), or by varying the total weight percent of uranium in the UZrH fuel meat. The TRIGA reactor is designed to be an inherently safe reactor and is designed to be cooled by natural convection. Thus it is of paramount importance to avoid the film boiling regime; failing to do so will subject the fuel to severe conditions, possibly resulting in its failure due to fuel/clad interactions. A high DNBR ratio is a primary consideration in the reactor core design and thus any modification in the core should ensure that this criteria is always met. In order to fully utilize the capabilities of KSU TRIGA reactor facility, it should be operated at its full capacity, i.e. 1.25 MW, which requires increasing the fuel loading in some of the E- and F-ring elements from 8.5% U (nominal) to 12.0% U (nominal). Only four elements are to be switched to 12% loading, causing local power peaking in these elements. The degree of power peaking to be expected in these 12%-loaded fuel elements was calculated based on a detailed MCNP (Monte Carlo N-Particle code) model of the reactor core. Some preliminary results of the calculations are presented in this paper based on changes in peaking factors and power levels. To predict reactor thermal behavior, axial changes in the peaking factors obtained from MCNP models were used to calculate volumetric energy density functions for a single channel one dimensional RELAP model and a three dimensional COMSOL model of the TRIGA MARK-II reactor. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Exploring the Use of Muon Momentum for Detection of Nuclear Material Within Shielded Spent Nuclear Fuel Dry Casks

The inspection of the storage containers that house used nuclear fuel is an important issue facing the nuclear industry; currently, there are limited options available to provide for even minimal inspections. During the long storage life of the fuel, intermediate handling and transportation activities may take place that can create concerns regarding fuel integrity (safety) and continuity of knowledge (safeguards and security). Accidents and natural disasters also contribute to this concern. The reliable and safe management of used nuclear fuel is among the critical tasks to be addressed for the advancement of fission-based nuclear energy in the United States and elsewhere. Since the early 1950's, when the first nuclear power plant began to produce electricity, approximately 65,000 metric tons of used nuclear fuel have been generated in the United States. The postponement of the opening of a permanent repository until 2048 will lead to increased used fuel accumulation at reactor sites. A rapid increase in storage installations with more than 9,000 casks storing {approx}420,000 assemblies is anticipated by 2050 even if no new nuclear power plants are built. After used nuclear fuel has been placed inside a cask, the cask is sealed, not allowing for visual inspection. Conventional methods, e.g., x-rays, are limited in that they cannot penetrate dense, well-shielded objects. More sophisticated techniques, such as penetrating neutrons or proton radiography, require the use of an expensive accelerator. Efforts by Ziock et al. using neutron and gamma-ray imagers concluded that '...the signatures are deemed insufficient to uniquely identify individual casks of the same type...'. It was found that the gamma-ray and neutron images were dominated by diffusion and thermalization effects resulting from the large amount of shielding. One promising mode of inspection would use cosmic-ray muons. These muons are naturally generated, are able to penetrate dense materials, and are incident from all directions above the horizon. Tomography performed using these muons could potentially provide efficient, inexpensive, remote assessment of used nuclear fuel. However, current muon applications are limited in that they have not included momentum measurement. If momentum measurement could be incorporated, that would represent a big step forward because it would allow more accurate estimation of the muon observables, e.g., scattering variance. This approach has the potential to reconstruct the contents of a cask with improved resolution and identify partial defects, fuel assemblies, and individual fuel rods.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

LOCA and LOFA Safety Margin Analysis for U-10Mo Monolithic LEU Fuel Conversion of the MURR{sup R} Reactor

The University of Missouri Research Reactor (MURR{sup R}) is a multi-disciplinary research and education facility providing a broad range of analytical and irradiation services to the research community and the commercial sector. The source of neutrons for the experimental facilities of MURR{sup R} is a compact 10 MW core fueled with highly-enriched uranium (HEU) fuel. The continued presence of HEU fuel in civilian installations such as research and test reactors poses a threat to national and international security. Therefore, most research and test reactors, both domestic and international, have started a program of conversion to the use of low-enriched uranium (LEU) fuel. MURR{sup R} is one of five U.S. high-performance research and test reactors (USHPRRs) that are actively collaborating with the U.S. National Nuclear Security Agency (NNSA) Office of Material Management and Minimization (M3) to develop a suitable LEU fuel replacement for the currently required HEU fuel. A new type of high-density LEU fuel based on an alloy of uranium and molybdenum is currently undergoing qualification testing, and is expected to allow the conversion of all USHPRRs. As a part of the LEU conversion of MURR{sup R}, an analysis of core behavior under a set of accident conditions for a core fueled with an LEU monolithic alloy of uranium and 10 wt% molybdenum (U-10Mo) has been conducted. This paper presents the newest results for the loss-of-coolant accidents (LOCAs) and loss-of-flow accidents (LOFAs). Other classes of accidents have also been investigated but were not included here. Namely, reactivity insertion accidents (RIAs), fuel failure during operations accidents (FFAs), and fuel handling accidents (FHAs)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Microstructurally validated stable and predictable swelling in low-enriched uranium monolithic U-10Mo fuel mini-plates

Qualification of the low-enriched uranium (LEU) monolithic U-10 wt%Mo (U-10Mo) plate-type fuel system requires a demonstration of a stable and predictable fuel swelling behavior over the anticipated operating conditions of the United States high-performance research reactors (USHPRRs) selected for conversion to LEU operation. This will allow each reactor to develop appropriate safety margins that will retain fuel element lifetime coolability. Additionally, the fuel system must maintain performance attributes when fabricated at a commercial scale. The Mini-plate 1 experiment represents the first irradiation test of commercially fabricated miniaturized monolithic LEU U-10Mo fuel plates. Here, the swelling behavior within this experiment was compared against that of historical fuel developmental tests to reveal that the commercially fabricated fuel performed within the current recommended U-10Mo swelling model's predictions. Additionally, the fuel microstructural evolution was evaluated to link initial conditions to subtle variations detected in the swelling response, providing validation and confidence that the fuel system is robust.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗