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At least 127 records · Page 7
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

Development of a Subchannel Capability for Liquid-Metal Fast Reactors in Pronghorn

This report details the development and demonstration of an entirely new capability in Pronghorn, namely the ability to model liquid-metal fast reactor (LMFR) flow conditions on the engineering scale. We developed two modeling approaches for LMFR that can be used separately or be combined into a hybrid simulation: (1) a modern subchannel capability called Pronghorn-Subchannel for square and hexagonal lattices, and (2) a porous flow capability for LMFR geometries. The report emphasizes the novel aspects of the developed subchannel capability and the interoperability of the subchannel capability, porous flow capability, and multiphysics tools within the multiphysics object oriented simulation environment (MOOSE). Here we demonstrate the ability to: (1) Accurately model subchannel flow in hexagonal lattices; (2) Couple the subchannel flow model to multidimensional finite-element method (FEM) or finite-volume method (FVM) heat conduction models; (3) Model LMFRs using Pronghorn’s porous media FVM approach; (4) Couple porous flow FVM and subchannel models in a single simulation; (5) Explicitly model inter-wrapper flows along with conjugate heat transfer from the intra-element flow; and (6) Demonstrate the numerical robustness of the subchannel algorithm by simulating intra-element flow recirculation in a high-buoyancy, low-flow fuel element.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2023

Accelerated thermal property mapping of TRISO advanced nuclear fuel

TRistructural ISOtropic (TRISO) fuel is a leading-edge nuclear fuel form representing a departure from the more traditional nuclear fuel forms utilized in the reactor fleet of today. Rather than a monolithic fuel pellet of uranium dioxide, integral fuel forms containing TRISO fuel are composed of thousands of microencapsulated uranium-bearing fuel kernels and individually coated with multiple layers of pyrolytic carbon and silicon carbide. These multilayered ceramic coatings serve as an environmental barrier to ensure radioactive and chemically reactive fission products are contained within the reactor fuel elements, but also participate in the transfer of heat generated in the nuclear fuel to the coolant – the primary purpose of a nuclear reactor. Since traditional thermal property measurement techniques, such as laser flash analysis, would be unable to resolve the thermal properties of the individual TRISO coating layers, a simplified frequency-domain thermoreflectance technique has been developed to rapidly map the thermal properties of TRISO particles. Using this technique, the thermal properties of TRISO particles have been mapped from room temperature up to 1000 °C to examine the spatial variation and temperature-dependency of the thermal properties within each layer. Additionally, spatial-domain thermoreflectance was used to examine the anisotropy of the thermal properties for each layer at different locations within a single TRISO particle, and across multiple TRISO particles to assess the intra- and inter-particle uniformity of thermal properties, respectively. To elucidate the underlying causes for the measured variations in thermal properties, scanning electron microscopy and Raman spectroscopy were used to examine variations in microstructure and chemical bonding within the different coating layers. Results from this work are then compared with previous examinations of TRISO fuel particles and microstructurally driven mechanisms for the variations in the measured thermal properties of the different carbonaceous layers are discussed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Carbide-based fuel assembly for thermal propulsion applications

Carbide-based fuel assembly includes outer structural member of ceramic matrix composite material, the interior surface of which is lined in higher temperature regions with an insulation layer of porous refractory ceramic material. Continuous insulation layer extends the length of the fuel assembly or separate insulation layer sections have a thickness increasing step-wise along the length of the fuel assembly from upper (inlet) section towards bottom (outlet) section. A fuel element positioned inward of the insulation layer and between support meshes has a fuel composition including HALEU and the form of a plurality of individual elongated fuel bodies or one or more fuel monolith bodies containing coolant flow channels. Fuel assemblies are distributively arranged in a moderator block, with upper end of the outer structural member attached to an inlet for propellant and lower end of the outer structural member operatively interfaced with a nozzle forming a nuclear thermal propulsion reactor.

Barringer, Eric A.↗
DOE OSTI2025

Development of a thermal creep model for aluminum alloy 6061 cladding in U-10Mo monolithic fuel plates

Plate-type fuel elements consisting of a high-density, low-enriched uranium (LEU) U–10Mo-based fuel foil encapsulated in an aluminum alloy (AA) cladding are fabricated using the hot isostatic pressing (HIP) technique. During the HIP process, the fuel plate system is heated to 560 °C, then cooled to room temperature. This heat cycle significantly affects the mechanical properties of the aluminum cladding, and experimental investigations have shown that, post-HIP bonding, the mechanical properties of the aluminum cladding transition from those of AA 6061-T6 to something closer to the O temper. More specifically, the ultimate strength of the cladding decreases while its ductility increases, making it challenging to capture the changes in mechanical behavior and material properties. Understanding the residual stresses generated during the HIP process is critical for assessing the fuel plate’s integrity under various temperature, pressure, and irradiation. To simulate the HIP bonding process, the elastic, plastic, and thermal properties of the cladding are assumed to be similar to those of AA 6061-O temper. However, the primary challenge lies in the lack of available data for the creep model of the AA 6061 cladding during this transient process of HIP. The present study focuses on developing a computational model that predicts the creep behavior of the aluminum cladding in the fuel plates during the HIP process, as cladding creep significantly influences the residual stresses generated in U-10Mo fuel plates during HIP fabrication. Furthermore, as HIP bonding occurs at high temperatures that are nearing the melting point of aluminum, the present work considered a temperature-dependent Arrhenius-type creep model. In particular, a hyperbolic sine creep model is employed to estimate the creep properties of the as-fabricated aluminum cladding. In conclusion, the residual stresses predicted in the U-10Mo fuel when using the newly calibrated creep model closely align with the experimental measurements, validating the model’s accuracy.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Test Article and Test Plan for Heat-Pipe Gap-Conductance Testing

The SPHERE Facility at INL is established to test phenomena related to heat pipes in nuclear applications. The current capability of the SPHERE testbed is operation of liquid alkali-metal-cooled heat pipes in a variety of atmospheres. The controlled atmosphere can be used to prevent oxidation of system parts and also to affect the thermal conductivity of gases for any interfacial gaps within the heat-pipe assembly. Currently available modes are vacuum, air, nitrogen, helium, and argon. Controlled mixtures of helium and argon are available to vary the thermal conductivity of the gas. The primary test article will be based on the existing 7-hole block design used for shakedown testing of the SPHERE Facility. The SPHERE test stand will be used with test articles to establish the conductivity between fuel elements as well as between the heat pipe and other layers of the fuel-rod assembly. Existing and new test blocks will be used and modified with additional instrumentation to establish the thermal conductivity between the gap of the block wall and the heat pipes.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Rapid depletion analysis of flowing-pebble reactor systems at equilibrium using SCALE

Several high-temperature gas-cooled reactor concepts (and more recently, salt-cooled designs such as the fluoride salt-cooled high-temperature reactor) feature core designs employing continuously circulating fuel pebbles. These reactor designs permit both continuous online refueling of fuel elements as well as higher overall achievable discharge burnups. However, rapid calculation of time-dependent fuel isotopic inventories proves challenging for this class of dynamic systems with current analysis tools. While iterative approaches employing coupled neutron transport have been developed to solve this issue, rapid depletion analysis techniques are needed to calculate time-dependent inventories for individual pebbles and batches (and thus the construction of full- core inventory at equilibrium). We propose a depletion analysis strategy for this type of system for cores at equilibrium. Drawing upon previous neutronic analysis of the PBMR-400 equilibrium core, we demonstrate the viability of developing collapsed one-group cross section libraries suitable for performing rapid depletion analyses with SCALE. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2026

Impact of Molten Gallium on the Microstructure and Corrosion Behavior of Aluminum and Uranium-Aluminum Alloys for Used Nuclear Fuel Reprocessing

Test reactors around the world utilize highly enriched uranium fuel to achieve high neutron fluxes for materials testing. Once spent, the remaining uranium is a valuable resource for subsequent fuel fabrication. However, some of these test reactor cores consist of curved plate-type fuel elements, fabricated using aluminum alloy 6061 (AA6061) cladding to encapsulate a uranium-aluminum alloy (UAlx) fuel matrix. These assemblies require non-standard reprocessing approaches for uranium recovery, as aluminum dissolves readily in acidic solutions, generating large volumes of waste and complicating downstream chemical separations. In this work, we investigate a novel chemical decladding strategy based on the interaction between AA6061/UAlx and molten gallium (Ga). Ga is known to induce severe degradation of aluminum metal through liquid metal embrittlement (LME), even at relatively low Ga concentrations. By penetrating the aluminum crystal lattice, Ga disrupts grain cohesion and facilitates fracture or dissolution of the aluminum matrix. Thermodynamic analysis of the Al–Ga binary phase diagram suggests that Ga may offer a viable pathway to selectively weaken or dissolve the AA6061 cladding, and potentially the aluminum component of the UAlx fuel matrix within. To this end, parametric experiments were performed at 50 °C and 100 °C across a range of Al–Ga atomic fractions. At lower Al fractions, the AA6061 was completely molten after 2 hours of exposure to the Ga metal. In contrast, samples with higher Al fractions (0.9 Al, 0.1 Ga) contained residual solids after 2 hours, which were characterized by microstructural examination using electron backscatter diffraction (EBSD) and transmission electron microscopy (TEM). These Al-Ga compositions were also evaluated using FactSage thermodynamic modeling to further elucidate the relationship between phase diagram behavior and LME.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

In-situ microstructure observation of oxidized SiC layer in surrogate TRISO fuel particles under krypton ion irradiation

In accidental scenarios of high temperature gas-cooled reactors, both oxidation of and irradiation to the SiC layer in tri-structural-isotropic (TRISO) fuel particles can change the microstructure and integrity of the fuel elements. Here, in the present study, microstructure and defect evolution in the oxidized SiC layer of surrogate TRISO fuel particles under 1.2 MeV krypton ion irradiation was observed by in-situ transmission electron microscopy. The SiC layers oxidized in water vapor at 1200 °C were irradiated azt room temperature and 800 °C and at damage levels of 0.28–11.2 dpa, respectively. SiC and SiO 2 were found to still be in their crystal structures at the damage level of 11.2 dpa at 800 °C, while SiC was observed to have been amorphized at only 0.56 dpa irradiation at room temperature. The defect number density at 800 °C was an order of magnitude lower than that in the sample irradiated at room temperature. Also, crystalline SiO 2 had higher radiation resistance compared to SiC. A defect reaction rate theory was utilized to understand the fundamental defect evolution process and irradiation resistance difference.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Model MC&A for Pebble Bed Reactors (Technical Direction No. 5 Task 2.6 Letter Report)

In preparation for non-light water reactor (non-LWR) activities, US Nuclear Regulatory Commission (NRC) staff are advancing risk-informed and performance-based licensing approaches and addressing key policy issues. One non-LWR reactor concept is a pebble bed reactor (PBR). This reactor design uses spherical fuel elements (pebbles) that are continually added to and removed from the reactor core. The free movement of the fuel in this design presents new challenges for material control and accounting (MC&A) programs. Therefore, an assessment of MC&A program features and measures for a PBR was performed to help NRC staff develop associated MC&A regulations or regulatory guides. The current regulatory framework for non-LWR fuel cycles excludes support for licensing reviews for MC&A programs for PBRs. Licensing reviews of an MC&A program for PBRs can be facilitated by (1) a model MC&A program for a PBR based on identification and assessment of MC&A program features and recommended measures for a reference PBR and (2) a methodology for assessing MC&A performance that can help assess different MC&A program features and measures. This report supports the NRC’s non-LWR Vision and Strategy Near-Term Implementation Action Plans.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
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 OSTI2024

Criticality Analysis of FSV Spent Nuclear Fuel in the DOE Standard Canister

The U.S. Department of Energy (DOE) is responsible for managing over 300 types of spent nuclear fuel (SNF). To manage this large variety of fuel types, DOE plans to employ standardized canisters for the transportation, long-term storage, and eventual disposal of SNF. Idaho National Laboratory is currently supporting DOE’s SNF Packaging Demonstration Project, in which Fort Saint Vrain (FSV) fuel assemblies will be loaded into a DOE Standard Canister. This paper presents criticality calculations demonstrating that all four or five FSV fuel assemblies loaded into the DOE Standard Canister will remain subcritical in any expected normal or credible abnormal conditions. Previous criticality analyses were performed for one FSV fuel assembly and 12 Peach Bottom Core 2 fuel elements loaded into a DOE Standard Canister. This paper covers the criticality analysis performed for loading both four and five FSV fuel assemblies into a DOE Standard Canister. Various intact and degraded mode configurations were modeled in conducting the criticality calculations. This analysis encompassed three different configurations: (1) a single DOE Standard Canister loaded into a concrete storage overpack, (2) seven DOE Standard Canisters loaded into a concrete storage overpack, and (3) nine DOE Standard Canisters loaded into a concrete storage overpack. The overpack dimensions were varied for each of the three configurations, and transport, storage, and disposal scenarios were analyzed for each configuration. For transport scenarios, a pair of degradation cases was analyzed. In the first case, the fuel compacts became degraded and were removed from the fuel block, then deposited at the bottom of a horizontally placed canister, thereby simulating a drop event. The canister was considered to remain intact. In the second case, the spacing between horizontally placed canisters in a nine-canister overpack was reduced such that the canisters were piled on top of each other, simulating a drop event. For this case, no degradation of the canister internals or fuel was considered. For storage scenarios, the water moderator location in the system was varied to enable identification of the most reactive configurations. Dry and wet conditions were analyzed for the fuel materials, canister, and overpack. For disposal scenarios, two degradation cases were analyzed. In the first, the stainless-steel internals of the canister degraded to either hematite or goethite under both dry and wet conditions. In the second case, degraded FSV fuel formed a uranium-water slurry that filled the coolant/void holes. None of the cases presented exceeded the application specific upper subcritical limit.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

Advanced Characterization of Fuel-cladding Chemical Interaction between U-10Zr Fuel and HT9 Cladding Tested in Fast Flux Test Facility

Fuel cladding chemical interaction (FCCI) can greatly accelerate the cladding failure. However, due to the limited space in a fuel cladding assembly, it is historically challenging to gain an mechanistical understanding of the formation mechanism of FCCI and its influence on fuel and cladding performance. With the imminent need to qualify U-10Zr based metallic fuel cladded by HT-9 for advanced reactors demonstration project, it is of vital importance to use advanced characterization method to study FCCI in a unprecedent detailed manner and gain better mechanism understanding of FCCI. Mechanistic Fuel Failure (MFF) series of prototypic fuel elements irradiated in FFTF [1, 2] provides the best samples to study FCCI since the MFF-series assemblies had an axial fuel height the same as proposed length by industry partners. Jason et al. [3] has performed preliminary post-irradiation examination on a MFF fuel pin sample, which was extracted from the HT9 cladded U-10at.%Zr MFF-3 pin MFF-3 pin (#193045) at an axial location of X/L = 0.98. This sample has a peak burnup of 5.7 at.% and peak inner cladding temperature (PICT) of around 615 °C during in-core testing. Scanning electron microscope examination has identified visible FCCI region on more than half of the HT9 circumference [3]. The most striking feature are grain boundary attacking by apparently lanthanides (Lns) rich phase. However, SEM cannot provide accurate assessment of phase and concentration of grain boundary phases and prevented a better understanding of the formation mechanism of such attach. This study, by pairing transmission electron microscope (TEM) characterization and atom probe tomography (APT) techniques with in-situ micro-tensile testing in scanning electron microscope (SEM), aims at gaining in-depth understanding on the formed FCCI region. The identified FCCI region roughly consists of multilayers as illustrated in Figure 1 (c). The main findings are: (1) layer-B shows observable lanthanides (Lns) infiltration along grain boundaries and mechanical softening due to FCCI- and irradiation-induced microstructural and microchemistry changes, particularly the recovery of martensitic lath structure and dissolution of pre-existing M23C6 together with the formation of coarsened Laves phases, (Fe, Cr)2(Mo, W); (2) layer-C is Fe depleted but Lns significantly enriched, becoming very brittle; (3) layer-D is mainly composed of UFe2 and Lns; (4) three FCCI-induced intermetallic U-Fe-Zr phases, ? (Fe0.5Zr0.32U0.18), e (Fe0.3Zr0.4U0.3), ? (Fe0.06Zr0.23U0.71), were identified near layer-E; (5) the ? (Fe0.5Zr0.32U0.18) phase was characterized to be a face centered cubic (FCC) crystal structure. These results will help to better understanding the governing mechanism of FCCI and facilitating the development of theoretical model for assessing the performance of metallic fuel and cladding integrity.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Milestone 1.2.11: H 2 Production from Surrogate Non-Native Corrosion Plumes on Aluminum 6061-T6 Fuel Cladding Surrogates

Thick, localized, “non-native” corrosion plumes have been identified on Advanced Test Reactor fuel elements, raising concern on their impact on the radiolytic formation of molecular hydrogen gas (H 2 ) from aluminum-clad spent nuclear fuel (ASNF) under proposed extended (> 50 years) dry storage conditions. Here, we report our findings on H 2 generation from the gamma irradiation (up to 52 MGy) of surrogate “non-native” corrosion plume coupons: ambient-temperature-corroded (~350 days in water) aluminum alloy 6061 (AA6061) coupons in helium gas environments with ~0% added relative humidity. Additionally, we provide a comparison of proposed ASNF drying techniques— vacuum drying only, vacuum drying + 100 °C for 4 hr, and vacuum drying + 220 °C for 4 hr—on the yield of H 2 from these surrogate systems. The presented data indicates that similar amounts of H 2 (~2 × 10–3 µmol J–1) are formed from gamma irradiated AA6061 coupons corroded under different temperature regimes, i.e., ambient/350 days vs. 90 C/30 days. These findings validate current, complimentary modelling predictions based on high-temperature-corrosion irradiation data only. Further, the application of a heat treatment procedure (100 and 220 °C), in conjunction with vacuum drying, accelerated the rate at which a steady-state H 2 yield was attained, in comparison to vacuum only, due to the removal of H 2 precursors in the form of adsorbed waters. Interestingly, within the confidence limits of our measurements, negligible difference in total H 2 yield was found between the two investigated heat treatment procedures.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2022

Fast ignition inertial fusion energy using laser-driven ion beams

Ion fast ignition (IFI), or fusion fast ignition initiated by a laser-driven ion beam, is a promising path to high-gain inertial fusion energy (IFE). In IFI, cold, dense deuterium-tritium (DT) fuel is first assembled using lasers or pulsed power drivers. Then, a high-power ion beam is focused onto a small volume within the fuel (the hot spot), heating the fuel rapidly to conditions where fusion ignition takes place. Fusion burn in this hot spot propagates to the fuel surrounding the hot spot, leading to burnup of a significant fraction of this fuel and the possibility of high gain (G~100), as needed for inertial fusion energy. IFI uses separate drivers for the two basic elements, fuel compression and ignition, allowing maximum control and optimization of each. On the other hand, conventional laser fusion uses multiple beams of the same driver to compress the fuel and shock-heat its very center to ignite a burn wave. Despite impressive progress in conventional laser fusion, the precise spatial symmetry, temporal pulse shaping and timing required for high gain and IFE remain a serious unmet challenge.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2020

Codisposal Waste Package Loading Options for DOE SNF and HLW

The U.S. Department of Energy (DOE) is responsible for managing spent nuclear fuel (SNF) that is currently in, or will in the future come into, its possession. DOE must continue to safely store that SNF, transport it to an interim storage site or a repository, and dispose of it. These fuels come from a wide range of reactor types, such as light- and heavy-water-moderated reactors, graphite-moderated reactors, and breeder reactors, with various cladding materials and enrichments. Many of these reactors, now decommissioned, had unique design features, such as core configuration, fuel element and assembly geometry, moderator and coolant materials, operational characteristics, and neutron spatial and spectral properties resulting in a large diversity of reactor and fuel designs. Because of the wide variety and condition of SNF, a robust canister termed the DOE Standardized SNF Canister was proposed that would confine radionuclides and preclude moderator. The DOE Standard SNF Canister had four variations consisting of a 10- or 15-foot length and an 18- or 24-inch diameter. For ultimate disposal in the Yucca Mountain Repository these canisters were to be grouped with 24-inch diameter HLW canisters in a codisposal waste package. The 18-inch DOE Standardized SNF Canister could be placed in the middle of five HLW canisters. The 24-inch DOE Standardized SNF Canister would take the place of one the five HLW canister on the outer ring in the codisposal waste package. No DOE Standardized SNF Canisters have been loaded, and many fuel types can be loaded in multiple canisters. Depending on the selected diameter of the DOE Standardized SNF Canister for a particular fuel, the number of potential codisposal waste packages could vary. In addition, different types of DOE SNF may be able to be combined in a single canister. This could significantly change the number of DOE SNF canisters that are loaded and result in a decrease of the number of total potential codisposal waste packages. Ultimately, if too many HLW canisters need disposition, then a reduction in the number of DOE Standardized SNF Canisters will change the final total of codisposal waste packages very little. Likewise, if the ratio of DOE Standardized SNF Canisters to HLW canister produced is too high, then DOE Standardized SNF Canisters will not have enough HLW canisters to surround them in the codisposal waste package. This paper examines the results of differing loading strategies for a few DOE SNFs. It also compares the ranges of DOE Standardized SNF Canisters that may be produced to the ranges of HLW canisters that may be produced in order to determine potential scenarios.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Assessment of Near-Term Fuel Screening and Qualification Needs for Nuclear Thermal Propulsion Systems

Nuclear thermal propulsion (NTP) is an in-space propulsion technology capable of both high specific impulse (850–1000 s) and thrust (44–1112 kN), which can help reduce trip times for crewed missions beyond low Earth orbit. NTP technology has been demonstrated during historic programs. Over 20 ground test reactor experiments were performed, which demonstrated the prototypic reactor operations, during the Nuclear Engine for Rocket Vehicle Application (NERVA)/Rover program (1955–1972). Although historical programs have shown that NTP is a viable in-space propulsion technology, developing NTP in modern programs is contingent on the development and qualification of ultrahigh-temperature nuclear fuel technologies that can withstand engine operating conditions. In historical NTP development programs such as NERVA/Rover, prototypic reactor/engine schemes were ground tested to assess the overall system feasibility and to qualify the reactor fuel forms for eventual flight systems. Although this approach is effective to verify fuel performance under prototypic conditions, relying solely on full-scale NTP reactor tests as the pathway for verifying or qualifying fuel is inefficient and cost prohibitive today. Additionally, modern nuclear licensing requirements state that before test reactor approval, reactor components and fuel elements should be qualified via non-nuclear (out-of-pile) and nuclear (in-pile) testing under representative operating conditions. Using this methodology, fuel matures as production scale fabrication methods are established, and as produced fuel performance is demonstrated. Here, this paper provides an overview of historical approaches to NTP fuel performance maturation, including fuel screening and qualification needs, and provides insight for establishing an efficient testing paradigm that can be implemented to rapidly and affordably develop NTP fuel forms for eventual qualification.

42 ENGINEERING↗
DOE OSTI2023

Predicting Safety Rod Reactivity Insertion in the Advanced Test Reactor

The Advanced Test Reactor (ATR), and complimentary zero-power ATR Critical (ATRC) reactor, located at Idaho National Labs (INL), are undergoing conversion from Highly Enriched Uranium (HEU) to Low Enriched Uranium (LEU). Both have a variety of testing locations that can receive large variations in flux due to its unique serpentine design, consisting of five lobes surrounding nine flux traps (see Figure 1). Initial criticality and power distribution throughout the core are controlled by core-external outer shim control cylinders (OSCCs). Distinct test loops allow for testing at specific temperatures, pressures, and irradiation conditions such as flux and fission density. The ATR is one of the key nuclear engineering research and testing facilities within the DOE National Laboratory Complex, and the ATRC supports its operation [1]. Currently, the Office of Material Management and Minimization (M3) within the National Nuclear Security Administration of the DOE is working to convert the remaining research reactors, including the ATR, from 93% HEU fuel to 19.75% LEU fuel (LEU) to support non-proliferation [2]. Extensive materials testing at INL and internationally has demonstrated that a high-density uranium molybdenum (U 10Mo) alloy can meet the performance requirements of the remaining high powered research reactors. The current LEU fuel element design is named the LOWE element. However, there are many technical challenges to address before the conversion to LEU can be successful, including the accurate characterization of the reactor core physics with LEU fuel. To ensure safe operation of the ATR, reactor engineers prepare a CSAP (Core Safety Assurance Package) before each cycle. The purpose of the CSAP is to verify the reactor performance calculation used to determine if the selected fuel loading meets operational, experimental, and safety criteria. Many of the criteria in the CSAP are limits on reactivity insertion in various accident scenarios.

42 ENGINEERING↗
DOE OSTI2023

IER 296 Experiment-Thermal Analysis and IRSN Response [Slides]

This set of slides summarizes progress, highlights, and issues related to the LANL design engineering effort for the IER 296 critical experiment. FEA simulation results are presented for the fuel elements on an insulated surface. Preliminary responses for the meeting agenda are documented.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗