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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

Expansion of the Fuel Motion Monitoring System

The Idaho National Laboratory (INL) Transient Reactor Test Facility (TREAT) Fuel Motion Monitoring System (FMMS) is a diagnostic tool used to visualize the movement and location of fuel within the nuclear reactor during transient experiments. TREAT is designed to test the behavior of nuclear fuel under accident conditions, such as a rapid power increase. The TREAT FMMS, also referred to as the “hodoscope” because of the system’s massive steel collimator component, is located on the North beam port of the reactor. While the FMMS system is capable of incorporating a total of 360 sensors, initial refurbishment efforts resulted in a array of 96 fast neutron detectors. This initial array was configured to provide a narrow (2 channels) full-length view of experiment vessels with an extended 4-channel–wide region in the center of the array capable of providing a full view of all initially planned experiments. Planning and efforts to expand the FMMS array began immediately following the restart of TREAT. An additional 96 detectors were evaluated, characterized, and prepared for installation at TREAT during the summer of 2018. Unfortunately, funding reductions forced the project towards a 5-year hiatus. In October of 2023 funding from the Department of Energy’s Nuclear Science User Facilities enabled the expansion to recommence. This report serves to document the activities performed to expand the FMMS detector array to 192 channels in order to provide a broader view of the larger and more sophisticated test capsules currently planned for future irradiation in the TREAT reactor. This expansion also included doubling the data acquisition capability, addressing the new heat load on the system, and synchronizing the time for all digital components.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Developing the Irradiation Plan for U-10Mo LEU Fuel Qualification for the Advanced Test Reactor

The Department of Energy (DOE), National Nuclear Security Administration’s Office of Material Management and Minimization (M3) mission is to convert, remove, and dispose of vulnerable nuclear material located at civilian sites worldwide. As part of its mission, M3’s Office of Convert works around the world to convert research reactors and isotope production facilities to non-weapon-usable nuclear material, both domestically and abroad. In support of this, the Office of Convert is working with the Idaho National Laboratory (INL) to develop and qualify new fuels and technologies to support conversion efforts domestically and abroad. The INL also supports the Office of Convert by working on converting the Advanced Test Reactor (ATR) and the Advanced Test Reactor Critical (ATRC) from using High Enriched Uranium (HEU) fuel to Low Enriched Uranium (LEU) fuel. To ensure the viability of ATR LEU fuel, the fuel system should meet a similar functional and operational envelope as the current HEU fuel. A series of irradiation tests of the U-10Mo LEU fuel system is required to support the qualification of the fuel system for use in the ATR and several other U.S. High-Performance Research Reactors (USHPRRs). These irradiation tests include: • Mini-Plate tests (MP-1, MP-2, and MP-ATR), • Full-Size Plate tests (FSP-1, and FSP-ATR), and • Element Tests (ET-1, ET-2, and ET-3). The MP tests will use multiple small-scale samples that will provide sufficient data across a wide variety of irradiation conditions to qualify the fuel. The FSP tests will use a small number of full-sized fuel plates to confirm the MP test results are representative of the prototypic plate geometry performance. The MP and FSP irradiation tests occur in ATR test locations. The ETs will use full-sized fuel elements located in ATR core driver positions to explore the full operating ranges of the ATR and other USHPRRs and to confirm that the MP and FSP test results are representative of the prototypic element geometry performance. The established testing functional and operational requirements of MP-1, MP-2, FSP-1, and ET 1 achieve the base fuel qualification needs of the USHPRRs, while the high-level purposes of MP-ATR, FSP ATR, ET 2, and ET-3 are focused on demonstrating U-10Mo fuel for ATR use.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Advanced Reactors Development in USA

Small and micro-reactor advanced reactor development in the USA. Advanced Reactor programs, nuclear renaissance, and INL as a test bed facility. These slides are for presentations in general on the past, INL and nuclear engineering worldwide, GEN IV reactor designs, SMRs and micro Rx designs, multiple commercial designs, and what INL as a test bed means and who uses this to their advantage.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Corrosion testing needs and considerations for additively manufactured materials in nuclear reactors

Additive manufacturing (AM) technologies have developed rapidly in recent years, creating new opportunities and challenges for the nuclear industry; however, adoption requires that their corrosion performance be evaluated. Here, we discuss known reactor-specific corrosion issues for multiple reactor types and engineering concerns such as regulations and standards. A review of corrosion studies conducted on select AM alloys informs a discussion on key bulk and surface factors likely to impact corrosion behaviors. Recommendations to assess corrosion performance for AM materials are provided, including management of the unique nature of as-built AM surfaces and the inherent process variability that occurs for AM components.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Using the sound of nuclear energy

The generation of sound by heat has been documented as an “acoustical curiosity” since a Buddhist monk reported the loud tone generated by a ceremonial rice-cooker in his diary, in 1568. Over the last four decades, significant progress has been made in understanding “thermoacoustic processes,” enabling the design of thermoacoustic engines and refrigerators. Motivated by the Fukushima nuclear reactor disaster, we have developed and tested a thermoacoustic engine that exploits the energy-rich conditions in the core of a nuclear reactor to provide core condition information to the operators without a need for external electrical power. The heat engine is self-powered and can wirelessly transmit the temperature and reactor power level by generation of a pure tone which can be detected outside the reactor. We report here the first use of a fission-powered thermoacoustic engine capable of serving as a performance and safety sensor in the core of a research reactor and present data from the hydrophones in the coolant (far from the core) and an accelerometer attached to a structure outside the reactor. These measurements confirmed that the frequency of the sound produced indicates the reactor’s coolant temperature and that the amplitude (above an onset threshold) is related to the reactor’s operating power level. Furthermore, these signals can be detected even in the presence of substantial background noise generated by the reactor’s fluid pumps.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2022

Experiment design for the neutron irradiation of $\mathrm{PM-HIP}$ alloys for nuclear reactors

Here, this article describes the design of an Advanced Test Reactor (ATR) drop-in neutron irradiation experiment aiming to directly compare the performance of nuclear structural alloys fabricated by powder metallurgy with hot isostatic pressing (PM-HIP) against conventional casting or forging. There is growing interest in PM-HIP alloys for nuclear applications because of their microstructural uniformity, superior mechanical properties, and reduced dependence on welding and machining, compared to cast/forged alloys. Nuclear code-qualification of PM-HIP alloys requires neutron irradiation testing to demonstrate performance under relevant conditions. In this experiment, six nuclear structural alloys were irradiated: Ni-based alloys 625 and 690, Grade 91 ferritic steel, SA508 pressure vessel steel, and 304L and 316L austenitic stainless steels. The experiment is assembled into seven capsules in four test trains and irradiated in three ATR inboard A positions. Both the PM-HIP and cast/forged versions of each alloy were irradiated under nearly identical conditions for comparative purposes, to target doses of 1 ± 0.2 and 3 ± 0.2 dpa at temperatures of 300 ± 50 °C and 400 ± 50 °C. A thorough description of the experiment design and thermal, structural and neutronic analyses performed to ensure the targeted irradiation conditions are met is provided. Specimens were configured as small disks, compact tension specimens and tensile bars to facilitate post-irradiation examination (PIE) that will include mechanical testing, microstructure characterization, and fracture toughness testing. Given the considerations for ASTM standardized mechanical testing, comparative fluence and temperature across specimen pairs, and comprehensive PIE planning herein, this work serves as a template for future nuclear materials qualification experiment designs.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

Non-Nuclear MicroReactor Test Bed and Article

The goal of the project is to provide proof of concept using sodium heats pipes as an efficient way to transfer heat from to rods to an integrated Brayton Cycle. A single simulated heat pipe experiment was performed and the results of the experiment are discussed. The results are also compared to the ABAQUS model that was made for the simulated experiment. The second portion of the poster discusses the design and challenges related to the test bed and the instrumentation required for the project. This brushes over the general requirements and the application for the data.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

Combined Assay Systems for the Characterisation of Historical ILW at of the Dounreay Shaft and Silo-18140

In 1955 the Dounreay nuclear power development establishment was created on the northern coast of Scotland primarily to pursue the UK Government policy of developing Fast Breeder Reactor (FBR) technology. The site operated then by the United Kingdom Atomic Energy Authority (UKAEA) saw the development, test and operation of 3 nuclear reactors, two of them FBRs, alongside fabrication and reprocessing facilities for the FBR fuel elements. From the late 1950's until 1977 Intermediate Level radioactive Waste (ILW) generated on site was disposed of into a 65 m deep, 4.6 m diameter shaft. In January 1977, an explosion thought to be caused by hydrogen generated from sodium and potassium wastes reacting with water occurred in the upper part of the shaft, thus stopping any further waste disposal. A second ILW storage facility, the ILW-Silo, started operation in 1971 to deal with the increasing volume of intermediate level waste generated on site in the mid-1960's and continued receiving waste until 1998. The decommissioning of the Shaft and Silo facilities is a challenging engineering task due to the complexity and potential hazards linked with the retrieval and handling of historical waste. Following recovery it is intended to shred the majority of the waste and load it into 200-l assay drums. Sludges would be separated and solidified within a special annulus drum to make them suitable for assay. To support the decommissioning strategies put forward by Dounreay Site Restoration Limited (DSRL), Mirion Technologies Canberra UK Ltd have designed and completed the manufacture of a pair of combined non-destructive waste assay (NDA) systems. The two identical automated ILW counters are intended to provide a means to characterise the waste materials by performing active and passive neutron measurement as well as segmented high resolution gamma spectrometry (HRGS) of 200-l sacrificial waste drums. The active neutron technique is based on the Differential Die Away (DDA) technique using a pulsed Deuterium-Tritium neutron generator to interrogate fissile material present in the waste. The active neutron measurements support, as well as the passive neutron coincidence measurements, a matrix correction capability by use of an Add-a-Source mechanism. The publication will describe in broad lines the approach proposed by DSRL in the processing of the waste retrieved from both the Shaft and Silo facilities. It will also present a description of the overall design of the NDA counters, illustrating how the measurements to be performed will allow DSRL to safely characterise the retrieved waste materials. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

The Advanced Test Reactor

The Advanced Test Reactor (ATR) is currently the highest-powered research reactor in the world with a maximum thermal power of 250 MW. The ATR began operation in 1967 to support fuels and materials irradiation testing programs in the United States and current plans (2019) are to operate the ATR for two more decades. Since 2007, the ATR has been designated a National Scientific User Facility (NSUF) to provide more access for non-government researchers to use the ATR for academic irradiation research. This article is a brief description of the ATR history, construction, and experiment configurations.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Preliminary Evaluation of Driver Fuel Plates and Reduced Sodium Temperature to Increase Fast Test Reactor Experimental Neutron Irradiation Capabilities

All sodium-cooled fast test reactors operated so far have used essentially the same fuel geometrical ar-rangement, namely cylindrical fuel pins assembled into a triangular array in hexagonal fuel assemblies. Fuel compositions and dimensions may vary but the overall geometrical pattern remains identical. Fur-thermore, most sodium-cooled reactors have operated, and still operate, with core-average inlet and outlet sodium temperatures of, respectively, 350-400°C and 500-550°C. The present preliminary study revisits the standard fuel assembly geometry as well as operating tempera-tures to evaluate potential impact on fast test reactor neutron irradiation capabilities. The reference fuel is U-20Pu-10Zr. Pin (reference) and plate geometries are considered together with two sets of core-average sodium inlet/outlet temperatures: 350/500°C (reference) and 200/400°C. More prototypic environments could be obtained inside ad-hoc irradiation testing vehicles that are independent from the reactor primary coolant. Everything else being the same, fuel temperature is significantly lower in plates than in pins. Therefore, plates could in principle accommodate higher plutonium content, thus, reducing the need for uranium enrichment or allowing the use of lower quality plutonium (U-xPu-10Zr thermal conductivity and solidus temperature decrease as x increases, hence, necessitating additional thermal margins). For the reference pin configuration, lowering the inlet/outlet sodium temperatures from the reference 350/500°C down to 200/400°C provides additional thermal margins that can be used to increase the peak fast flux from about 4.5 × 10 15 n/cm 2 -s to 6 ×10 15 n/cm 2 -s for the same core power of 300 MW. Assuming 300 Equivalent Fuel Power Days (EFPD) of operation per calendar year, a (steel) test article could accu-mulate up to 75 dpa/year. The use of fuel plates provides even more thermal margins which may allow the peak fast flux to reach values as high as 8 × 10 15 n/cm 2 -s for both sets of inlet/outlet temperatures. In this environment, a test article could accumulate up to 95 dpa/year assuming 300 EFPD/year, hence greatly accelerating irradiation testing. Allowing a fast test reactor to operate over a wide range of inlet/outlet temperatures could add significant flexibility to its neutron irradiation capabilities. Steel present in the driver fuel assemblies is not expected to accumulate more than 90 displacements per atom (dpa) over its lifetime. Hence, steel alloys that are not suited for very high dpa but can operate over a wide range of temperatures (from low to high) would be appropriate for this application. Finally, because pin and plate assemblies fit on the same grid plate, a fast test reactor could in principle start-up with standard fuel pin assemblies and, later on, move to a plate-type Mark-II fuel to further in-crease its neutron irradiation capabilities while potentially reducing the level of uranium enrichment needed.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Molten Salt in the Advanced Test Reactor [Slides]

The Advanced Test Reactor (ATR) is one of the world’s most important resources for fuel qualification. Prior to this project, ATR could not support molten salt irradiation experiments. This work expands ATR’s capability to support fuel qualification for molten salt reactors.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
NASA NTRSDate not supplied

Stirling Power Conversion for Lunar Applications

NASA is developing Stirling power convertors for use in nuclear power systems that would provide electricity to users on the lunar surface. NASA’s Radioisotope Power Systems (RPS) Program is maturing advanced dynamic convertor technologies that would increase the efficiency beyond the heritage systems for missions to the Moon and other solar system bodies of interest. Development efforts have focus on Stirling and Brayton designs that meet reliability and robustness requirements for operation in the lunar environment. GRC is also maturing low power convertors in the range of 1-40 watts to power micro-spacecraft or a network of distributed probes and include basic demonstration of in-house designs for convertors and controllers. In addition to RPS applications, conceptual fission systems have been demonstrated using Stirling convertors in multiple electrically heated tests and one test heated by a nuclear reactor. The Stirling convertors used in these fission system concepts include 100 We convertors to enable a 1 kWe system, 1 kWe convertors to enable a 10 kWe system, and 6 kWe convertors to enable a 40 kWe system. The maturation of Stirling power convertors for nuclear applications includes verification of performance in relevant environments and validation of the design with a focus on high reliability and robustness. Over 1 million hours of testing has been accumulated on dozens of convertors since 1999 in the Stirling Research Laboratory at the NASA Glenn Research Center located in Cleveland Ohio, where the longest running convertor is the work record holder at 137,000 hours, or 15.5 years. These maturation efforts have been instrumental in increasing the technology readiness level of Stirling technologies for space applications.

S. Wilson↗
DOE OSTI2022

Coupling coarse-mesh CFD with fine-mesh CFD for modeling molten-salt reactors in the Virtual Test Bed (VTB)

The Nuclear Energy Advanced Modeling and Simulation (NEAMS) program aims at developing a simulation tool kit to accelerate the development and deployment of nuclear power technologies. NEAMS multiphysics tools have been designed to provide numerical simulation support for the design and licensing of GEN IV reactors. Pronghorn is NEAMS's coarse-mesh computational fluid dynamics (CFD) tool, which is designed to run 3D core transients in GEN IV reactors at a reduced computational cost. To increase their accuracy, coarse-mesh CFD simulations require calibrated closure coefficients. One way of computing these coefficients is via the Nek5000, NEAMS's high-fidelity CFD tool. This article discusses our current research lines in informing Pronghorn closure coefficients via Nek5000 to enable multiphysics simulations of the core cavity of the molten-salt fast reactor. We present an application in which Pronghorn mixing length turbulent viscosity has been calibrated from Nek5000 simulations. The resulting Pronghorn thermal-hydraulics model is then coupled to Griffin, the NEAMS neutron transport solver, to solve for the steady-state configuration of the molten-salt fast reactor. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Coupling coarse mesh CFD with fine mesh CFD for modeling for Molten Salt Reactors in the Virtual Test Bed VTB

The Nuclear Energy Advanced Modeling and Simulation (NEAMS) program aims at developing a simulation tool kit to accelerate the development and deployment of nuclear power technologies. NEAMS multiphysics tools have been designed to provide numerical simulation support for the design and licensing of GEN IV reactors. Pronghorn is NEAMS’s coarse-mesh computational fluid dynamics (CFD) tool, which is designed to run 3D core transients in GEN IV reactors at a reduced computational cost. To increase their accuracy, coarse-mesh CFD simulations require calibrated closure coefficients. One way of computing these coefficients is via the Nek5000, NEAMS’s high-fidelity CFD tool. This article discusses our current research lines in informing Pronghorn closure coefficients via Nek5000 to enable multiphysics simulations of the core cavity of the molten-salt fast reactor. We present an application in which Pronghorn mixing length turbulent viscosity has been calibrated from Nek5000 simulations. The resulting Pronghorn thermal-hydraulics model is then coupled to Griffin, the NEAMS neutron transport solver, to solve for the steady-state configuration of the molten-salt fast reactor.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗