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DOE OSTI2019

Fuel Safety Research Plan for Accident Tolerant Fuels

This document outlines a strategic plan for a safety research program for accident tolerant fuels. Accident tolerant fuels (ATFs) are nuclear fuel materials and designs for existing and future light water reactors. This effort is under the Department of Energy Office of Nuclear Energy’s Nuclear Technology Research and Development Advanced Fuels Campaign. The development of ATF is an industry lead initiative with fuel designs being championed by Westinghouse, Framatome, and General Electric. Industry goals for ATF deployment include implementing the first batch reloads of coated cladding and doped fuel designs into commercial reactors by 2023 and having the first batch reloads of more advanced designs into commercial reactors by 2026. Follow on goals include implementing the first full cores of coated cladding and doped fuel designs into commercial reactors by 2026 and implementing full cores of more advanced designs by 2030. The objectives of the national laboratory research programs within the Advanced Fuels Campaign related to ATF are: 1. Develop and maintain of unique nuclear testing infrastructure, which includes test reactors, hot cell PIE equipment, and unique out of pile separate effects testing facilities. 2. Respond directly to industry requests for support by partnering with the ATF vendors and executing experiments and/or analysis on their behalf. 3. Conduct independent scientific investigations into the performance of the different ATF concepts that are relevant to Industry designs. To achieve these objectives, the safety research program will execute transient irradiation experiments at the TREAT facility at Idaho National Laboratory with the aim of identifying the relevant transient fuel behaviors for each fuel design in order to establish appropriate fuel safety criteria and thresholds. Two test series are described in this plan, (1) an Integral Test Series, and (2) a Focused Effects Test Series. In fulfillment of the first objective for the safety research program, Integral Effects Tests will be developed and demonstrated using standard UO2 Zircaloy fuel designs in both the fresh and previously irradiated case. The Integral Test Series will consist of three campaigns to evaluate transient fuel performance in Reactivity Initiated Accidents, Loss of Coolant Accidents, and in generalized Power Cooling Mismatch scenarios. Development of these test programs also enables the safety research program to fulfill its second objective by making these three test programs available to each industry team. Initially, priority will be given in the integral test series to the coated cladding and doped fuel concepts as their deployment schedule is the most aggressive. To fulfill the third objective of the safety research program, a Focused Effects Test Series is also being developed which will examine specific transient fuel behaviors that are of general interest to all the ATF designs under development. These tests look at transient fuel behavior related to pellet cladding mechanical interaction, coated cladding swelling, rupture, and time at temperature behavior, non oxide fuel melting and mechanical interaction with Silicon Carbide cladding, UO2 and doped UO2 fuel fragmentation and transient fission gas release, and finally fuel coolant interaction. This document provides a general overview of the objectives of the safety research program as well as providing background into how programs like this have been used in the past to develop the criteria necessary for the safe operation of the present fuel designs. An overview of the materials and methods used for safety testing is presented. Finally the Integral and Focused Effects tests series are introduced and described. This document will be maintained as a “living” document with annual updates to reflect evolving requirements and needs.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Out-of-Pile Testing and Instrumentation Transient Water Irradiation System

Current initiatives to increase the burnup of conventional nuclear fuels past the approximate 62 GWd/t limit have been spurred on by direct savings to refueling and waste storage. The technical justification for a new license limit requires extensive qualification through experimental testing. Unlike beginning-of-life fuels, high-burnup fuels are more susceptible to fuel fragmentation, relocation, and dispersal (FFRD), therefore more data is needed to characterize fuels under key accident scenarios. The Transient Reactor Test Facility (TREAT) located at the Idaho National Laboratory has developed a testing apparatus architecture to test fuels and claddings at prototypic conditions. The Transient Water Irradiation System (TWIST) is the latest iteration of a testing device capable of conducting loss of coolant accidents (LOCAs) in TREAT. The Out-of-Pile Testing and Instrumentation TWIST (OPTI-TWIST) is an electrically heated device that is analogous to TWIST. OPTI-TWIST allows for detailed instrumentation and thermal-hydraulic characterization. TWIST ultimately aims to conduct the most advanced in-situ diagnostics to evaluate FFRD in a prototypic LOCA. Moreover, it will explore the phenomenological bifurcation of a decay-energy heat up driven LOCA and a stored-energy heat up driven LOCA. The instrumentation suite includes conventional thermocouples and pressure transducers in addition to an electro impedance sensor, an acoustic emission sensor, an optical pressure sensor, and an optical pyrometer. Characterizing these instruments in OPTI-TWIST eliminates complications of irradiation effects while preserving extreme thermal-hydraulic conditions. Finally, benchmarking both devices to a thermal-hydraulic code like the Reactor Excursion and Leak Analysis Program (RELAP)5-3D provides a unique opportunity for iteration. Pre-test predictions and post-test interpretations inform the physical designs, operational procedures, test conditions, and instrumentation types and positions.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

Software Requirements for a Point Kinetics Neutronics Code for Simulating TREAT Reactor Transients

This document provides the software requirements for a computer code to perform point kinetics-based simulations of nuclear reactor transients. These requirements address all aspects of code development and use, including input and output parameters and structure, software capabilities, and supporting documentation. This software is intended to replace the existing point kinetics code TREKIN, which is currently used for the analysis of the Transient Reactor Test Facility (TREAT).

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Irradiation of ultrasonic sensors and adhesive couplants for application in light water reactor primary loop piping and components

The Electric Power Research Institute (EPRI) Nuclear Sector and US Department of Energy Light Water Reactor Sustainability Program are committed to engaging in research and development endeavors to address materials aging issues specific to long term operation of light water power reactors. To this effect, EPRI launched an industry initiative to develop nondestructive evaluation systems for online monitoring of existing cracks in light water reactor primary coolant loop piping and components. One of the goals of this initiative is to develop a sensor system (or systems) that can determine nondestructively if cracks are growing or arrested and, in the case of the former, to characterize their growth rates. A missing component of this initiative is an experimental assessment of how sensors and adhesive couplants will perform in service when exposed to chronic energetic neutron radiation, particularly at the primary coolant loop hot and cold leg dissimilar metal welds, which join the primary loop piping to the reactor pressure vessel and reside in the vicinity of the reactor core. The objective of this experimental study was to determine how ultrasonic transducers and adhesive couplants perform when exposed to irradiation in a test reactor to simulate and accelerate in-service exposure. Further, to achieve this objective, the signal stability of piezoelectric transducers and performance of adhesive couplants as a function of accumulated fast neutron fluence were characterized by collecting ultrasonic data in-situ during irradiation. Of particular interest were the ultrasonic signal quality and time decay of the amplitude of acoustic reflections as a function of fast neutron fluence. The results of the study showed that, of the 8 transducer/substrate sample assemblies tested, only 3 generated usable ultrasonic signals through the conclusion of the irradiation campaign. It was found that high temperature epoxy tends to ultrasonically couple the sensors to the substrates better than three types of refractory ceramic cements studied, as is supported by post irradiation examination. The results obtained through this experimental study will be utilized in the achievement of the overall goal of development of a sensor system to perform online monitoring of primary loop components.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2013

Alternatives Analysis for the Resumption of Transient Testing Program

An alternatives analysis was performed for resumption of transient testing. The analysis considered eleven alternatives – including both US international facilities. A screening process was used to identify two viable alternatives from the original eleven. In addition, the alternatives analysis includes a no action alternative as required by the National Environmental Policy Act (NEPA). The alternatives considered in this analysis included: 1. Restart the Transient Reactor Test Facility (TREAT), 2. Modify the Annular Core Research Reactor (ACRR) which includes construction of a new hot cell and installation of a new hodoscope, and 3. No Action.

42 ENGINEERING↗
DOE OSTI2019

Transitional cores and fuel cycle analyses in support of MIT reactor low enriched uranium fuel conversion

The Massachusetts Institute of Technology Reactor (MITR) is a research reactor in Cambridge, Massachusetts designed primarily for experiments using neutron beam and in-core irradiation facilities. At 6 MW, it delivers neutron flux and energy spectrum comparable to power light water reactors (LWRs) in a compact core using highly enriched uranium (HEU) fuel. In the framework of non-proliferation policy, research and test reactors have started a program to convert HEU fuel to low enriched uranium (LEU) fuel. A new type of LEU fuel based on a high density alloy of uranium and molybdenum (U-10Mo) is expected to allow conversion of U.S. high performance research reactors (USHPRRs) like the MITR. The Preliminary Safety Analysis Report (PSAR) for the MITR LEU conversion has been submitted and currently under review by the U.S. Nuclear Regulatory Commission (NRC). A transition core plan, from 22 fresh LEU fuel elements (i.e., beginning-of-life) gradually to 24 of them arranged in an equilibrium configuration, is expected to serve as an appendix chapter in the final SAR. The current study reported here presents the fuel cycle development, which eventually leads to the transition core plan. The results confirm the equilibrium state, where both the shim bank movement (i.e., core reactivity) as well as the content of fissile materials stabilize, can be achieved by fixed-pattern fuel management. Fission density has been evaluated for a number of fully discharged LEU fuel elements, using a best-estimate approach. The fuel cycle calculations also generate power profiles at each core state. A steady-state thermal-hydraulic safety analysis has thus been performed, where onset of nucleate boiling (ONB) is considered as the safety criterion. The results confirm significant margin to ONB at all analyzed transition and equilibrium fuel cycle states.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

NSUF BRR Cask Shipment Summary Report

On September 16, 2021 the Idaho National Laboratory shipped a Battelle Research Reactor (BRR) cask loaded with five Nuclear Science User Facilities (NSUF) experiments and one Accident Tolerant Fuels (ATF) experiment from the Advanced Test Reactor (ATR) to the Hot Fuel Examination Facility (HFEF) at the INL Materials and Fuels Complex (MFC) for post-irradiation examination.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Code Coverage Status of ARC Code-DIF3D

The Argonne Reactor Code (ARC) software system supports users in their fast reactor design goals by providing neutronic, thermal-hydraulic, and structural analysis capabilities. DIF3D plays a pivotal role in the ARC system as the primary homogenized assembly neutronic calculation methodology for fast reactor problems. Over its 40 years history, ARC software usage with DIF3D has been applied to numerous fast and thermal spectrum reactor analysis projects with good to excellent comparison against experiments. With continued improvement of computation resources, many of the geometry modeling capabilities in DIF3D that were primarily used in low order schemes are not really needed anymore. Today, the diffusion and transport capabilities of DIF3D-VARIANT are primarily used in the reactor design process with some scattered usage of DIF3D-FD and DIF3D-Nodal. In recent work, the DIF3D software verification was completed for DIF3D-FD and DIF3D-VARIANT on the geometry options used in the Versatile Test Reactor project. While we can be confident that these capabilities of DIF3D are well used and thus trusted, it does not demonstrate that all possible input options of DIF3D are actually working, but just those that were tested as part of VTR are and that they are correct. Thus, the purpose of the present work is to identify a set of test problems for DIF3D and assess the code coverage of DIF3D for those test problems. The goal is to document what parts of the existing DIF3D code are touched by the set of test problems and which are not. Because the verification work done on DIF3D-VARIANT and DIF3D-FD was focused on the most common uses of DIF3D for fast reactor analysis, the code coverage assessment of those capabilities is the highest priority. This will ensure that nothing is being missed by the existing verification test problems that DIF3D relies upon. The DIF3D-Nodal capability will also be inspected for code coverage as part of this work to further ensure that regular regression testing of DIF3D will trap any likely errors the end user might experience with the DIF3D software. The code coverage analysis of DIF3D was performed with the Code Coverage Tool of the Intel Fortran compiler which requires modifications to the compilation of DIF3D. The detailed coverage tables are given for each submodule of DIF3D separately, and for the submodules which are primarily developed for DIF3D, most of the source files could be at least partially touched. Most of the uncovered parts/files could be easily ignored, because they are either for error message and debugging output or obviously not needed by DIF3D. Out of the entire source codes of DIF3D, only a few uncovered modules deserve further investigation.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Code Coverage Status of the ARC Code DIF3D

The Argonne Reactor Code (ARC) software system supports users in their fast reactor design goals by providing neutronic, thermal-hydraulic, and structural analysis capabilities. DIF3D plays a pivotal role in the ARC system as the primary homogenized assembly neutronic calculation methodology for fast reactor problems. Over its 40 years history, ARC software usage with DIF3D has been applied to numerous fast and thermal spectrum reactor analysis projects with good to excellent comparison against experiments. With continued improvement of computation resources, many of the geometry modeling capabilities in DIF3D that were primarily used in low order schemes are not really needed anymore. Today, the diffusion and transport capabilities of DIF3D-VARIANT are primarily used in the reactor design process with some scattered usage of DIF3D-FD and DIF3D-Nodal. In recent work, the DIF3D software verification was completed for DIF3D-FD and DIF3D-VARIANT on the geometry options used in the Versatile Test Reactor project. While we can be confident that these capabilities of DIF3D are well used and thus trusted, it does not demonstrate that all possible input options of DIF3D are actually working, but just those that were tested as part of VTR are and that they are correct. Thus, the purpose of the present work is to identify a set of test problems for DIF3D and assess the code coverage of DIF3D for those test problems. The goal is to document what parts of the existing DIF3D code are touched by the set of test problems and which are not. Because the verification work done on DIF3D-VARIANT and DIF3D-FD was focused on the most common uses of DIF3D for fast reactor analysis, the code coverage assessment of those capabilities is the highest priority. This will ensure that nothing is being missed by the existing verification test problems that DIF3D relies upon. The DIF3D-Nodal capability will also be inspected for code coverage as part of this work to further ensure that regular regression testing of DIF3D will trap any likely errors the end user might experience with the DIF3D software. The code coverage analysis of DIF3D was performed with the Code Coverage Tool of the Intel Fortran compiler which requires modifications to the compilation of DIF3D. The detailed coverage tables are given for each submodule of DIF3D separately, and for the submodules which are primarily developed for DIF3D, most of the source files could be at least partially touched. Most of the uncovered parts/files could be easily ignored, because they are either for error message and debugging output or obviously not needed by DIF3D. Out of the entire source codes of DIF3D, only a few uncovered modules deserve further investigation.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Eddy Current Flow Meter Measurements and Modeling Validation

The goal of the eddy current flow meter (ECFM) is to monitor velocities and temperatures of conductive liquid metals. The ECFM can also detect voids in two-phase liquid metal flows. Measurement of these flows can greatly improve the accuracy of the empirical formulas used for fluid dynamics near the reactor core and aid in mapping any non-uniformities in fluid flow such as estimates of temperature and phase. While the eventual goal is to develop the ECFM for the liquid sodium surrounding the Versatile Test Reactor (VTR) core, this report focuses on modeling and experimental validation of ECFM for a moving solid test rod and for liquid mercury in the Target Test Facility (TTF) flow loop to better understand and further develop ECFM measurement capabilities by validating model predictions with benchmark experiments. Eventually the models could be used to design and optimize future ECFMs for specific applications such as monitoring flow within the VTR pool.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Thermal-hydraulic and Fuel Performance Scoping Studies of a Flowing Water Capsule in TREAT

The restart of the Transient Reactor Test (TREAT) facility has provided a much needed capability for integral transient testing of nuclear fuel. This testing is necessary to qualify new fuel concepts such as those developed under the Accident Tolerant Fuel (ATF) program, increasing the burnup limit of light water reactor (LWR) fuels, or a variety of other programs under the Advanced Fuels Campaign (AFC). The ATF campaign has been the main driver behind the development and implementation of a variety of capsules for TREAT experiments. The Separate-Effect Test Holder (SETH) was a dry capsule that enabled testing of ATF concepts under reactivity-initiated accident (RIA) heating conditions. Following SETH, the Static Environment Rodlet Transient Test Apparatus (SERTTA) was developed to enable RIA testing in a static water environment. In efforts to support the need for future Loss-of-Coolant Accident (LOCA) tests, the Transient Water Irradiation System for TREAT (TWIST) capsule has been developed that allows for water to drain from around the fuel rod and lower the pressure to simulate LOCA conditions. All these capsules that have been developed for LWR fuel testing all lack the capability for forced convective cooling which in some applications may limit their ability to test under prototypic conditions. A design and modeling effort has been started to modify the TWIST capsule by adding a flow tube and impeller that can create flowing coolant conditions for the fuel rod. RELAP5-3D and BISON models are being used to study the differences between RIA, LOCA, Anticipated Operational Occurrences (AOO), LWR power cycling, and other transient scenarios under flowing conditions capable in the flowing capsule and the current stagnant water capsules (SERTTA and TWIST). The scoping study will provide guidance on the needed capabilities for the flowing capsule and the limitations of the currently developed capsule for LWR testing in TREAT.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Post-irradiation examination of the Sirius-1 nuclear thermal propulsion fuel test

Nuclear Thermal Propulsion (NTP) systems hold promise in reducing transit times for exploration of the solar system by both crewed and uncrewed missions. NTP systems currently under investigation include a once-through high-temperature gas-cooled fission reactor to provide thermal energy to heat the coolant, which also serves as the propellant. The fuel system consists of angular UN fuel particles dispersed in a matrix of W/Re, creating a ceramic and metallic composite or cermet, which has been irradiated in Idaho National Laboratory's (INL) Transient Reactor Test Facility (TREAT). This enabled evaluation of these materials under representative nuclear heating rates (~95 K/s) and peak surface temperatures (~2527 K). Further, these surface conditions were indicative of achievement of the target peak internal temperature of approximately 2850 K. These tests named Sirius-1 (UN–W/Re), have been irradiated and this paper will present post-irradiation examination results. The Sirius-1 test produced minor, stable cracks in the fuel specimen and spalling of surface material. Volatilized uranium ‘soot’ was found deposited on the inner wall of the irradiation capsule indicating loss of some fissile material from the fuel specimen. Spalling from the surfaces was also noted upon visual inspection. Uranium diffusion from the fuel particles resulted in the formation of U/Re phases and the production of a laminar microstructure on the edge of the fuel system. Overall, the specimen was stable and remained within a coolable geometry upon conclusion of multiple thermal cycles to prototypical operating temperatures.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Development of Conceptual Lead Cartridge Design to Perform Irradiation Experiments in VTR

This paper seeks to introduce the latest design of the Extended Length Test Assembly–Cartridge Lead (ELTA-CL) with associated thermal-hydraulic (TH) assessment and related experiment activities to support the critical component development performed by the ELTA-CL team (Los Alamos National Laboratory, Westinghouse Electric Company, and the University of New Mexico). The goal of the ELTA-CL program is to develop and validate an experimental capability to perform irradiation experiments in the Versatile Test Reactor (VTR) addressing Lead Fast Reactor (LFR) technology gaps, in support of the commercial development of advanced lead-cooled fast reactor concepts. Through a design maturation process and parametric study, a conceptual design is proposed to meet the requirements for material and corrosion testing. Thermal-hydraulic characteristics for the conceptual design at desired operating conditions are assessed with systems-level (one-dimensional) and computational fluid dynamics (three-dimensional) simulations. Along with the conceptual design work, experimental activities for the development of critical components such as the pump and flowmeter are undertaken. From both the modeling study and the experimental results, the design requirements of the Phase 1 ELTA-CL (e.g., 500°C and 2 m/s) are achievable with the current conceptual design. Additional design improvements and safety assessments at both steady-state and transient conditions for the final ELTA-CL design will be pursued.

42 ENGINEERING↗
DOE OSTI2020

Transition Core Analysis for HEU to LEU Fuel Conversion at the University of Missouri Research Reactor

A new type of low-enriched uranium (LEU) fuel based on an alloy of uranium and molybdenum is expected to allow the conversion of U.S. domestic high-performance research and test reactors requiring high density fuel from highly enriched uranium (HEU) to LEU. The University of Missouri Research Reactor (MURR (R)) has undergone design and performance calculations for conversion to this LEU fuel. Presented in this paper is the analysis of a crucial step in the conversion process: the sequence of MURR transition cores from all fresh to equilibrium burnup LEU operations. During the initial conversion from HEU to LEU fuel, MURR will operate atypically due to the lack of burned LEU elements. Given the constraints of MURR operation and experiments, a proposed transition scheme minimizes the time MURR operates atypically compared to the prototypic cycles currently run with HEU fuel and moves quickly to the same sort of equilibrium cycles for the LEU fuel.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Development of Metallic Fuels for Actinide Transmutation

Research and development activities on metallic fuels are focused on their potential use for actinide transmutation in future sodium fast reactors. As part of this application, there is also a need for a near zero-loss fabrication process and a desire to demonstrate a multifold increase in burnup potential. The incorporation of Am and Np into the traditional U-20Pu-10Zr metallic fuel alloy was demonstrated in the US during the Integral Fast Reactor Program of the 1980’s and early 1990’s. However, the conventional counter gravity injection casting method performed under vacuum, previously used to fabricate these metallic fuel alloys, was not optimized for mitigating loss of the volatile Am constituent in the casting charge; as a result, approximately 40% of the Am casting charge failed to be incorporated into the as-cast fuel alloys. Fabrication development efforts of the past few years have pursued an optimized bottom-pour casting method to increase utilization of the melted charge to near 100%, and a differential pressure casting approach, performed under an argon overpressure, has been demonstrated to result in essentially no loss of Am due to volatilization during fabrication. In short, a path toward zero-loss fabrication of metallic fuels including minor actinides has been shown to be feasible. Irradiation testing of advanced metallic fuel alloys in the Advanced Test Reactor (ATR) has been underway since 2003. Testing in the ATR is performed inside of cadmium-shrouded positions to remove >99% of the thermal flux incident on the test fuels, resulting in an epi-thermal driven fuel test that is free from gross flux depression and producing an essentially prototypic radial temperature profile inside the fuel rodlets. To date, three irradiation test series (AFC-1,2,3) have been completed. Over 20 different metallic fuel alloys have been tested to burnups as high as 30% with constituent compositions of Pu up to 30%, Am up to 12%, Np up to 10%, and Zr between 10 and 60%. In general, the performance of all of these substantially disparate metallic fuel alloys has been observed to be excellent, and their irradiation behaviors are generally consistent with historic norms for metallic fuels without minor actinide additions and having lower Pu or Zr contents. Future work is being undertaken with a view toward increasing the burnup potential of metallic fuels even more. Design innovations under investigation include: 1) lowering the fuel smear density in order to accommodate more swelling, 2) annular fuel geometry to eliminate the need for a sodium bond, 3) minor alloy additions to stabilize lanthanide fission products inside the fuel and prevent their transport to the cladding where they can participate in fuel-cladding chemical interaction (FCCI), and 4) coatings/liners on the cladding inner surface to mitigate FCCI and enable higher temperature operation. This paper will present the current state of development of metallic fuels for actinide transmutation in the US. Highlights will include recent results from metallic fuel casting experiments, experiments to identify alloy additions to immobilize lanthanide fission products, and postirradiation examinations of annular metallic fuels at low burnup.

actinide transmutation↗
DOE OSTI2013

Hot Isostatic Press Can Optimization for Aluminum Cladding of U-10Mo Reactor Fuel Plates: FY12 Final Report and FY13 Update

Currently, the proposed processing path for low enriched uranium – 10 wt. pct. molybdenum alloy (LEU-10Mo) monolithic fuel plates for high power research and test reactors includes hot isostatic pressing (HIP) to bond the aluminum cladding that encapsulates the fuel foil. Initial HIP experiments were performed at Idaho National Laboratory (INL) on approximately ¼ scale “mini” fuel plate samples using a HIP can design intended for these smaller experimental trials. These experiments showed that, with the addition of a co-rolled zirconium diffusion barrier on the LEU-10Mo alloy fuel foil, the HIP bonding process is a viable method for producing monolithic fuel plates. Further experimental trials at Los Alamos National Laboratory (LANL) effectively scaled-up the “mini” can design to produce full-size fuel prototypic plates. This report summarizes current efforts at LANL to produce a HIP can design that is further optimized for higher volume production runs. The production-optimized HIP can design goals were determined by LANL and Babcock & Wilcox (B&W) to include maintaining or improving the quality of the fuel plates produced with the baseline scaled-up mini can design, while minimizing material usage, improving dimensional stability, easing assembly and disassembly, eliminating machining, and significantly reducing welding. The initial small-scale experiments described in this report show that a formed-can approach can achieve the goals described above. Future work includes scaling the formed-can approach to full-size fuel plates, and current progress toward this goal is also summarized here.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Shutdown-induced tensile stress in monolithic miniplates as a possible cause of plate pillowing at very high burnup

Post-irradiation examination of Reduced Enrichment for Research and Test Reactors (RERTR)-12 miniplates showed that in-reactor pillowing occurred in at least 4 plates, rendering performance of these plates unacceptable. To address in-reactor failures, efforts are underway to define the mechanisms responsible for in-reactor pillowing, and to suggest improvements to the fuel plate design and operational conditions. To achieve these objectives, the mechanical response of monolithic fuel to fission and thermally-induced stresses was modeled using a commercial finite element analysis code. Calculations of stresses and deformations in monolithic miniplates during irradiation and after the shutdown revealed that the tensile stress generated in the fuel increased from 2 MPa to 100 MPa at shutdown. The increase in tensile stress at shutdown possibly explains in-reactor pillowing of several RERTR-12 miniplates irradiated to the peak local burnup of up to 1.11x1022 fissions/cm3 . This paper presents the modeling approach and calculation results, and compares results with post-irradiation examinations and mechanical testing of irradiated fuel. The implications for the safe use of the monolithic fuel in research reactors are discussed, including the influence of fuel burnup and power on the magnitude of the shutdown-induced tensile stress.

rertr↗
DOE OSTI2014

The effect of birthrate granularity on the release-to-birth ratio for the AGR-1 in-core experiment

The AGR-1 Advanced Gas Reactor (AGR) tristructural-isotropic-particle fuel experiment underwent 13 irradiation intervals from December 2006 until November 2009 within the Idaho National Laboratory Advanced Test Reactor in support of the Next Generation Nuclear Power Plant program. During this multi-year experiment, release-to-birth rate ratios were computed at the end of each operating interval to provide information about fuel performance. Fission products released during irradiation were tracked daily by the Fission Product Monitoring System using 8-h measurements. Birth rate calculated by MCNP with ORIGEN for as-run conditions were computed at the end of each irradiation interval. Each time step in MCNP provided neutron flux, reaction rates and AGR-1 compact composition, which were used to determine birth rate using ORIGEN. The initial birth-rate data, consisting of four values for each irradiation interval at the beginning, end, and two intermediate times, were interpolated to obtain values for each 8-h activity. The problem with this method is that any daily changes in heat rates or perturbations, such as shim control movement or core/lobe power fluctuations, would not be reflected in the interpolated data and a true picture of the system would not be presented. At the conclusion of the AGR-1 experiment, great efforts were put forth to compute daily birthrates, which were reprocessed with the 8-h release activity. The results of this study are presented in this paper.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗