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At least 703 records · Page 39

Synthesis of phase-pure U 2 N 3 microspheres and its decomposition into UN

Uranium mononitride (UN) is important as a nuclear fuel. Fabrication of UN in its microspherical form also has its own merits since the advent of the concept of accident-tolerant fuel, where UN is being considered as a potential fuel in the form of TRISO particles. But, not many processes have been well established to synthesize kernels of UN. Therefore, a process for synthesis of microspherical UN with a minimum amount of carbon is discussed herein. First, a series of single-phased microspheres of uranium sesquinitride (U 2 N 3 ) were synthesized by nitridation of UO 2 +C microspheres at a few different temperatures. Resulting microspheres were of low-density U 2 N 3 and decomposed into low-density UN. The variation of density of the synthesized sesquinitrides as a function of its chemical composition indicated the presence of extra (interstitial) nitrogen atoms corresponding to its hyperstoichiometry, which is normally indicated as α-U 2 N 3 . Average grain sizes of both U 2 N 3 and UN varied in a range of 1–2.5 μm. In addition, these had a considerably large amount of pore spacing, indicating the potential sinterability of UN toward its use as a nuclear fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

General Multifidelity Surrogate Models: Framework and Active-Learning Strategies for Efficient Rare Event Simulation

Estimating the probability of failure for complex real-world systems using high-fidelity computational models is often prohibitively expensive, especially when the probability is small. Exploiting low-fidelity models can make this process more feasible, but merging information from multiple low-fidelity and high-fidelity models poses several challenges. Here, this paper presents a robust multi-fidelity surrogate modeling strategy in which the multi-fidelity surrogate is assembled using an active learning strategy using an on-the-fly model adequacy assessment set within a subset simulation framework for efficient reliability analysis. The multi-fidelity surrogate is assembled by first applying a Gaussian process correction to each low-fidelity model and assigning a model probability based on the model's local predictive accuracy and cost. Three strategies are proposed to fuse these individual surrogates into an overall surrogate model based on model averaging and deterministic/stochastic model selection. The strategies also dictate which model evaluations are necessary. No assumptions are made about the relationships between low-fidelity models, while the high-fidelity model is assumed to be the most accurate and most computationally expensive model. Through two analytical and two numerical case studies, including a case study evaluating the failure probability of Tristructural isotropic-coated (TRISO) nuclear fuels, the algorithm is shown to be highly accurate while drastically reducing the number of high-fidelity model calls (and hence computational cost).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Effect of graphite properties in thermal analysis of CHTR: A parametric study

Compact High Temperature Reactor (CHTR) is a {sup 233}U-Thorium fuelled, lead-bismuth cooled reactor. The CHTR core mainly consists of graphite and beryllium oxide (BeO). The CHTR core consists of nineteen prismatic beryllium oxide (BeO) moderator blocks. These 19 blocks contain centrally located graphite fuel tubes. The BeO moderator blocks are surrounded by reflector blocks (partially graphite and partially BeO). The nuclear heat from the core is removed passively by natural circulation of the coolant between top and bottom plenums, upward through the fuel tubes and returning through the downcomer tubes at the periphery. The temperature gradient in fuel tubes, downcomer tubes and BeO is very high and therefore, to take care of the differential thermal expansion, gaps are provided in the core between the tubes and other core components. These gaps affect the heat transfer through the core in radial direction. In addition, there is a large variation in thermal properties of graphite which in turn affects the thermal behaviour of the core in various operating conditions. The fuel of CHTR is TRISO coated particle fuel. These particles are packed in with graphite powder as matrix and made into cylindrical compacts these compacts are packed in the bores of fuel tube. In this study, the effect of the thermal conductivity variation of the graphite on the temperature distribution of the core and density variation of the matrix graphite material in fuel compact on the maximum fuel kernel temperature is studied along with the overall role of graphite properties variation in heat transfer.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

High-Fidelity CFD Simulation of Mixed Convection and Forced Convection in a Pebble Bed Test Reactor Core

The Hermes low-power [35-MW(thermal)] reactor will be built and operated by Kairos Power LLC (KP) to demonstrate its fluoride salt-cooled high-temperature reactor (FHR) technology. In the KP FHR, the reactor core is composed of randomly packed pebbles with TRISO fuel particles inside with FLiBe flow upward through the core acting as a coolant. Previous numerical and experimental studies have been limited to either a small-size bed or to a lack of detailed measurements for heat transfer. Here, to address the lack of high-fidelity heat transfer data in a real-size FHR core, in this study, we simulated a pebble bed core with 34 374 pebbles randomly packed, similar to the Hermes reactor's size. The core radius was 14 times that of the pebble diameter, while the core height was 45 times. In this work, we were particularly interested in a mixed convection regime, where buoyancy is important. Therefore, we performed several large-eddy simulations at different Reynolds numbers (160 to 1000) with gravitational force included. The spectral element computational fluid dynamics code NekRS with graphics processing unit acceleration was used for this study. The low-Mach number approximation was applied to address property changes in the FLiBe and to account for buoyancy. A pure hexahedral mesh with 60 million elements was generated by the Voronoi cell method. At the polynomial order of 5, the total degrees of freedom was 7.5 billion. The developed case in this work is the first of its kind in terms of size and complexity. The local numerical data across the domain were obtained and compared with empirical correlations. After examining the data, we found the following conclusions. For pressure drop, the Reger correlation predicted less than a 5% error. On the other hand, for heat transfer, the Wakao correlation outperformed the others. Based on our findings, we recommend the use of the Wakao correlation for the Nusselt number calculation, and for pressure drop, the KTA (Kerntechnischer Ausschuss) correclation, among the available experimental correlations. In conclusion, the Reger direct numerical simulation-driven correlation for pressure drops should also be considered, given its best agreement with our calculations.

Mixed Convection↗

Feasibility of Recycling Discharged Microreactor Heavy Metal in Light Water and Sodium-Cooled Fast Reactors: A Neutronics Analysis

Nuclear microreactors (MRs) offer unique advantages, such as rapid deployment, potability, low maintenance requirements, and operational flexibility. Their compact size makes them a promising solution for decentralized power generation, particularly in remote areas, military bases, and disaster-stricken regions. However, MRs face challenges, including unutilized fissile material at the end of life, economic inefficiency, increased heavy metal (HM) waste complicating disposal, and the accumulation of plutonium (Pu) with high 239 Pu concentrations raising proliferation risks. Here, this study investigated the neutronics feasibility of a novel three-stage fuel cycle where discharged HM from MRs is recycled and burned in light water reactors and sodium-cooled fast reactors. This approach converts discharged HM into valuable fuel, enhancing the efficiency of MR deployments while improving the safeguardability of their final waste products. Neutronics analysis demonstrated that the safety characteristics of reactor designs in each stage were minimally impacted by the proposed cycle. For two representative MR designs, a fast-spectrum MR with solid pellet fuel and a thermal-spectrum MR with TRISO (TRi-structural-ISOtropic) fuel compacts, the proposed fuel cycle reduced the uranium disposal mass flow rate by ~60%, decreased the 235 U enrichment of the discharge fuel to ~1 wt%, eliminated plutonium disposal, and increased the cumulative fuel burnup to ~580 gigawatt-day per metric ton of initial heavy metal (GWd/t-iHM) or 60% fissions per initial metal atom. Despite the significant differences between the two MR designs, the performance and infrastructure requirements of the developed fuel cycles were remarkably similar, indicating its generalizability to a broader class of MRs.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Neutronic Evaluation of a Liquid Salt–Cooled Reactor Assembly

This study presents a thorough parametric neutronic analysis of a plate-based tristructual isotropic (TRISO) fuel particle bearing liquid salt–cooled reactor assembly. The analyses presented investigated the effects of altering fuel enrichment, packing fraction, plate region thicknesses, assembly structure thicknesses, assembly size, numbers of plates per assembly, use of burnable poison materials, replacement of assembly and plate carbon material with silicon carbide, and use of uranium nitride fuel kernels. The effects or trends observed included reactivity behavior, discharge burnup, cycle length, and other key design parameters such as moderator temperature coefficients, coolant density coefficients, control blade worth, and impacts upon power peaking (i.e., power and flux distributions).

Nuclear Science & Technology↗

Kinetics of silver release from microfuel with taking into account the limited-solubility effect

The effect of a limited solubility of silver in silicon carbide on silver release from a microfuel with a TRISO coating is studied. It is shown that a limited solubility affects substantially both concentration profiles and silver release from a microfuel over a broad range of temperatures. A procedure is developed for obtaining fission-product concentration profiles in a microfuel and graphs representing the flow and integrated release of fission products on the basis of data from neutron-physics calculations and results obtained by calculating thermodynamics with the aid of the Ivtanthermo code and kinetics with the aid of the FP-Kinetics code. This procedure takes into account a limited solubility of fission products in protective coatings of microfuel.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

Multidimensional Fuel Performance Code: BISON

BISON is a finite element based nuclear fuel performance code applicable to a variety of fuel forms including light water reactor fuel rods, TRISO fuel particles, and metallic rod and plate fuel (Refs. [a, b, c]). It solves the fully-coupled equations of thermo-mechanics and species diffusion and includes important fuel physics such as fission gas release and material property degradation with burnup. BISON is based on the MOOSE framework (Ref. [d]) and can therefore efficiently solve problems on 1-, 2- or 3-D meshes using standard workstations or large high performance computers. BISON is also coupled to a MOOSE-based mesoscale phase field material property simulation capability (Refs. [e, f]). As described here, BISON includes the code library named FOX, which was developed concurrent with BISON. FOX contains material and behavioral models that are specific to oxide fuels.

Hales, Jason↗

Representative Source Terms and the Influence of Reactor Attributes on Functional Containment in Modular High-Temperature Gas-Cooled Reactors

Modular high-temperature gas-cooled reactors (MHTGRs) offer a high degree of passive safety. The low power density of the reactor and the high heat capacity of the graphite core result in slow transients that do not challenge the integrity of the robust TRISO fuel. Another benefit of this fuel form and the surrounding graphite is their superior ability to retain fission products under all anticipated normal and off-normal conditions, which limits reactor accident source terms to very low values. In this paper, we develop estimates of the source term for a generic MHTGR to illustrate the performance of the radionuclide barriers that comprise the MHTGR functional containment. We also examine the influence of initial fuel quality, fuel performance/failure, reactor outlet temperature, and retention outside of the reactor core on the resultant source term to the environment.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

COMBINE7.1 - A Portable ENDF/B-VII.0 Based Neutron Spectrum and Cross-Section Generation Program

COMBINE7.1 is a FORTRAN 90 computer code that generates multigroup neutron constants for use in the deterministic diffusion and transport theory neutronics analysis. The cross-section database used by COMBINE7.1 is derived from the Evaluated Nuclear Data Files (ENDF/B-VII.0). The neutron energy range covered is from 20 MeV to 1.0E-5 eV. The Los Alamos National Laboratory NJOY code is used as the processing code to generate a 167 fine-group cross-section library in MATXS format for Bondarenko self-shielding treatment. Resolved resonance parameters are extracted from ENDF/B-VII.0 File 2 for a separate library to be used in an alternate Nordheim self-shielding treatment in the resolved resonance energy range. The equations solved for energy dependent neutron spectrum in the 167 fine-group structure are the B3 or B1 zero-dimensional approximations to the transport equation. The fine group cross sections needed for the spectrum calculation are first prepared by Bondarenko self-shielding interpolation in terms of background cross section and temperature. The geometric lump effect, when present, is accounted for by augmenting the background cross section. Nordheim self-shielded fine group cross sections for a material having resolved resonance parameters overwrite correspondingly the existing self-shielded fine group cross sections when this option is used. COMBINE7.1 coalesces fine group cross sections into broad group macroscopic and microscopic constants. The coalescing is performed by utilizing fine-group fluxes and/or currents obtained by spectrum calculation as the weighting functions. The multigroup constants may be output in any of several standard formats including INL format, ANISN 14** free format, CCCC ISOTXS format, and AMPX working library format. ANISN-PC, a one-dimensional (1-D) discrete-ordinate transport code, is incorporated into COMBINE7.1. As an option, the 167 fine-group constants generated by zero-dimensional COMBINE portion in the program can be used to calculate regionwise spectra in the 1-D ANISN portion, all internally to reflect the 1-D transport correction. The regionwise spectra are then used to generate mutigroup regionwise neutron constants. The 1-D neutron transport can be performed up to three stages, e.g., from a TRISO fuel to PEBBLE to 1-D full core wedge. In addition, COMBINE7.1 has now the capability of adjoint flux calculation through the 1-D ANISN transport. Photon transport capability is also added. For this, a photon production and photo-atomic cross section library, MATNG.LIB, was generated in MATXS format through NJOY code. The photon production cross section matrix is of 167 neutron - 18 photon groups. Photo-atomic cross sections, including heating, are in 18 energy groups.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Fusion-fission hybrid for fissile fuel production without processing

Two scenarios are typically envisioned for thorium fuel cycles: “open” cycles based on irradiation of 232 Th and fission of 233 U in situ without reprocessing or “closed” cycles based on irradiation of 232 Th followed by reprocessing, and recycling of 233 U either in situ or in critical fission reactors. This study evaluates a third option based on the possibility of breeding fissile material in a fusion-fission hybrid reactor and burning the same fuel in a critical reactor without any reprocessing or reconditioning. This fuel cycle requires the hybrid and the critical reactor to use the same fuel form. TRISO particles embedded in carbon pebbles were selected as the preferred form of fuel and an inertial laser fusion system featuring a subcritical blanket was combined with critical pebble bed reactors, either gas-cooled or liquid-salt-cooled. The hybrid reactor was modeled based on the earlier, hybrid version of the LLNL Laser Inertial Fusion Energy (LIFE) system, whereas the critical reactors were modeled according to the Pebble Bed Modular Reactor (PBMR) and the Pebble Bed Advanced High Temperature Reactor (PB-AHTR) design. An extensive neutronic analysis was carried out for both the hybrid and the fission reactors in order to track the fuel composition at each stage of the fuel cycle and ultimately determine the plant support ratio, which has been defined as the ratio between the thermal power generated in fission reactors and the fusion power required to breed the fissile fuel burnt in these fission reactors. It was found that the maximum attainable plant support ratio for a thorium fuel cycle that employs neither enrichment nor reprocessing is about 2. This requires tuning the neutron energy towards high energy for breeding and towards thermal energy for burning. A high fuel loading in the pebbles allows a faster spectrum in the hybrid blanket; mixing dummy carbon pebbles with fuel pebbles enables a softer spectrum in the critical reactors. This combination consumes about 20% of the thorium initially loaded in the hybrid reactor (~200 GWd/tHM), partially during hybrid operation, but mostly during operation in the critical reactor. The plant support ratio is low compared to the one attainable using continuous fuel chemical reprocessing, which can yield a plant support ratio of about 20, but the resulting fuel cycle offers better proliferation resistance as fissile material is never separated from the other fuel components.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

AGR-1 Safety Test Predictions using the PARFUME code

The PARFUME modeling code was used to predict failure probability of TRISO-coated fuel particles and diffusion of fission products through these particles during safety tests following the first irradiation test of the Advanced Gas Reactor program (AGR-1). These calculations support the AGR-1 Safety Testing Experiment, which is part of the PIE effort on AGR-1. Modeling of the AGR-1 Safety Test Predictions includes a 620-day irradiation followed by a 300-hour heat-up phase of selected AGR-1 compacts. Results include fuel failure probability, palladium penetration, and fractional release of fission products. Results show that no particle failure is predicted during irradiation or heat-up, and that fractional release of fission products is limited during irradiation but that it significantly increases during heat-up.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

BISON Theory Manual: The Equations behind Nuclear Fuel Analysis

BISON is a finite element-based nuclear fuel performance code applicable to a variety of fuel forms including light water reactor fuel rods, TRISO particle fuel, and metallic rod and plate fuel. It solves the fully-coupled equations of thermomechanics and species diffusion, for either 1D spherical, 2D axisymmetric or 3D geometries. Fuel models are included to describe temperature and burnup dependent thermal properties, fission product swelling, densification, thermal and irradiation creep, fracture, and fission gas production and release. Plasticity, irradiation growth, and thermal and irradiation creep models are implemented for clad materials. Models are also available to simulate gap heat transfer, mechanical contact, and the evolution of the gap/plenum pressure with plenum volume, gas temperature, and fission gas addition. BISON is based on the MOOSE framework and can therefore efficiently solve problems using standard workstations or very large high-performance computers. This document describes the theoretical and numerical foundations of BISON.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Verification of the CENTRM Module for Adaptation of the SCALE Code to NGNP Prismatic and PBR Core Designs

The generation of multigroup cross sections lies at the heart of the very high temperature reactor (VHTR) core design, whether the prismatic (block) or pebble-bed type. The design process, generally performed in three steps, is quite involved and its execution is crucial to proper reactor physics analyses. The primary purpose of this project is to develop the CENTRM cross-section processing module of the SCALE code package for application to prismatic or pebble-bed core designs. The team will include a detailed outline of the entire processing procedure for application of CENTRM in a final report complete with demonstration. In addition, they will conduct a thorough verification of the CENTRM code, which has yet to be performed. The tasks for this project are to: Thoroughly test the panel algorithm for neutron slowing down; Develop the panel algorithm for multi-materials; Establish a multigroup convergence 1D transport acceleration algorithm in the panel formalism; Verify CENTRM in 1D plane geometry; Create and test the corresponding transport/panel algorithm in spherical and cylindrical geometries; and, Apply the verified CENTRM code to current VHTR core design configurations for an infinite lattice, including assessing effectiveness of Dancoff corrections to simulate TRISO particle heterogeneity.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

AGR-2 safety test predictions using the PARFUME code

This report documents calculations performed to predict failure probability of TRISO-coated fuel particles and diffusion of fission products through these particles during safety tests following the second irradiation test of the Advanced Gas Reactor program (AGR-2). The calculations include the modeling of the AGR-2 irradiation that occurred from June 2010 to October 2013 in the Advanced Test Reactor (ATR) and the modeling of a safety testing phase to support safety tests planned at Oak Ridge National Laboratory and at Idaho National Laboratory (INL) for a selection of AGR-2 compacts. The heat-up of AGR-2 compacts is a critical component of the AGR-2 fuel performance evaluation, and its objectives are to identify the effect of accident test temperature, burnup, and irradiation temperature on the performance of the fuel at elevated temperature. Safety testing of compacts will be followed by detailed examinations of the fuel particles to further evaluate fission product retention and behavior of the kernel and coatings. The modeling was performed using the particle fuel model computer code PARFUME developed at INL. PARFUME is an advanced gas-cooled reactor fuel performance modeling and analysis code (Miller 2009). It has been developed as an integrated mechanistic code that evaluates the thermal, mechanical, and physico-chemical behavior of fuel particles during irradiation to determine the failure probability of a population of fuel particles given the particle-to-particle statistical variations in physical dimensions and material properties that arise from the fuel fabrication process, accounting for all viable mechanisms that can lead to particle failure. The code also determines the diffusion of fission products from the fuel through the particle coating layers, and through the fuel matrix to the coolant boundary. The subsequent release of fission products is calculated at the compact level (release of fission products from the compact). PARFUME calculates the release fraction as the ratio of the number of atoms released from the compact to the amount produced in the compact fuel kernels and through uranium contamination. These safety test predictions aim at giving guidance for the upcoming safety tests to be conducted at ORNL and INL. The AGR-2 compacts modeled in this study belong to the four U.S. capsules of the AGR-2 irradiation test train. For each capsule, the compacts with the lowest and highest burnups were selected to provide a prediction envelope. The objective of this document is to present results of calculations and analyses made by PARFUME on these selected AGR-2 compacts. These results include: • Fuel failure probability • Palladium penetration • Fractional release of fission products Details associated with completion of these calculations are provided in the remainder of this document: the AGR-2 Safety Testing plan is briefly introduced in Section 2, PARFUME modeling is outlined in Section 3, results are described in Section 4, conclusions are given in Section 5, and references are listed in Section 6.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

AGR-2 Irradiation Test Final As-Run Report, Rev 2

This document presents the as-run analysis of the AGR-2 irradiation experiment. AGR-2 is the second of the planned irradiations for the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. Funding for this program is provided by the U.S. Department of Energy as part of the Very High Temperature Reactor (VHTR) Technical Development Office (TDO) program. The objectives of the AGR-2 experiment are to: (a) Irradiate UCO (uranium oxycarbide) and UO 2 (uranium dioxide) fuel produced in a large coater. Fuel attributes are based on results obtained from the AGR-1 test and other project activities. (b) Provide irradiated fuel samples for post-irradiation experiment (PIE) and safety testing. (c) Support the development of an understanding of the relationship between fuel fabrication processes, fuel product properties, and irradiation performance. The primary objective of the test was to irradiate both UCO and UO 2 TRISO (tri-structural isotropic) fuel produced from prototypic scale equipment to obtain normal operation and accident condition fuel performance data. The UCO compacts were subjected to a range of burnups and temperatures typical of anticipated prismatic reactor service conditions in three capsules. The test train also includes compacts containing UO 2 particles produced independently by the United States, South Africa, and France in three separate capsules. The range of burnups and temperatures in these capsules were typical of anticipated pebble bed reactor service conditions. The results discussed in this report pertain only to U.S. produced fuel. In order to achieve the test objectives, the AGR-2 experiment was irradiated in the B-12 position of the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) for a total irradiation duration of 559.2 effective full power days (EFPD). Irradiation began on June 22, 2010, and ended on October 16, 2013, spanning 12 ATR power cycles and approximately three and a half calendar years. The test contained six independently controlled and monitored capsules. Each U.S. capsule contained 12 compacts of either UCO or UO2 AGR coated fuel. No fuel particles failed during the AGR-2 irradiation. Final burnup values on a per compact basis ranged from 7.26 to 13.15% FIMA (fissions per initial heavy-metal atom) for UCO fuel, and 9.01 to 10.69% FIMA for UO 2 fuel, while fast fluence values ranged from 1.94 to 3.47 x 10 25 n/m 2 (E >0.18 MeV) for UCO fuel, and from 3.05 to 3.53 x 10 25 n/m 2 (E >0.18 MeV) for UO 2 fuel. Time-average volume-average (TAVA) temperatures on a capsule basis at the end of irradiation ranged from 987°C in Capsule 6 to 1296°C in Capsule 2 for UCO, and from 996 to 1062°C in UO 2 -fueled Capsule 3. By the end of the irradiation, all of the installed thermocouples (TCs) had failed. Fission product release-to-birth (R/B) ratios were quite low. In the UCO capsules, R/B values during the first three cycles were below 10 -6 with the exception of the hotter Capsule 2, in which the R/Bs reached 2 x 10 -6 . In the UO 2 capsule (Capsule 3), the R/B values during the first three cycles were below 10 -7 . R/B values for all following cycles are not reliable due to gas flow and cross talk issues.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

AGR-2 Irradiation Test Final As-Run Report

This document presents the as-run analysis of the AGR-2 irradiation experiment. AGR-2 is the second of the planned irradiations for the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. Funding for this program is provided by the U.S. Department of Energy as part of the Very High Temperature Reactor (VHTR) Technical Development Office (TDO) program. The objectives of the AGR-2 experiment are to: (a) Irradiate UCO (uranium oxycarbide) and UO 2 (uranium dioxide) fuel produced in a large coater. Fuel attributes are based on results obtained from the AGR-1 test and other project activities. (b) Provide irradiated fuel samples for post-irradiation experiment (PIE) and safety testing. (c) Support the development of an understanding of the relationship between fuel fabrication processes, fuel product properties, and irradiation performance. The primary objective of the test was to irradiate both UCO and UO 2 TRISO (tri-structural isotropic) fuel produced from prototypic scale equipment to obtain normal operation and accident condition fuel performance data. The UCO compacts were subjected to a range of burnups and temperatures typical of anticipated prismatic reactor service conditions in three capsules. The test train also includes compacts containing UO 2 particles produced independently by the United States, South Africa, and France in three separate capsules. The range of burnups and temperatures in these capsules were typical of anticipated pebble bed reactor service conditions. The results discussed in this report pertain only to U.S. produced fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

AGR-2 Irradiation Test Final As-Run Report

This document presents the as-run analysis of the AGR-2 irradiation experiment. AGR-2 is the second of the planned irradiations for the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. Funding for this program is provided by the U.S. Department of Energy as part of the Very High Temperature Reactor (VHTR) Technology Development Office (TDO) program. The objectives of the AGR-2 experiment are to: 1. Irradiate UCO (uranium oxycarbide) and UO 2 (uranium dioxide) fuel produced in a large coater. Fuel attributes are based on results obtained from the AGR-1 test and other project activities. 2. Provide irradiated fuel samples for post-irradiation experiment (PIE) and safety testing. 3. Support the development of an understanding of the relationship between fuel fabrication processes, fuel product properties, and irradiation performance. The primary objective of the test was to irradiate both UCO and UO 2 TRISO (tristructural isotropic) fuel produced from prototypic scale equipment to obtain normal operation and accident condition fuel performance data. The UCO compacts were subjected to a range of burnups and temperatures typical of anticipated prismatic reactor service conditions in three capsules. The test train also includes compacts containing UO 2 particles produced independently by the United States, South Africa, and France in three separate capsules. The range of burnups and temperatures in these capsules were typical of anticipated pebble bed reactor service conditions. The results discussed in this report pertain only to U.S.-produced fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗