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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 361 records · Page 20
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Spent nuclear fuel dynamic performance under normal conditions of transport

The objective of this project is to evaluate the spent nuclear fuel (SNF) vibration integrity under normal condition of transport. The SNF fatigue strength data has been collected from various pressurized water reactors. The CIRFT (Cyclic Integrated Reversible-Bending Fatigue Tester) has been used to simulate transportation environments. CIRFT is an enabling hot-cell testing technology developed at Oak Ridge National Laboratory (ORNL). The CIRFT consists mainly of a U-frame testing setup and a real-time curvature measurement method. The three-component U-frame setup of the CIRFT has two rigid arms and linkages to a universal testing machine. The curvature of rod bending is obtained through a three-point deflection measurement method. Three linear variable differential transformers (LVDTs) are used and clamped to the side connecting plates of the U-frame to capture the deformation of the rod. CIRFT results have provided insight into the fuel/clad system response to transportation related loads. Because of the non-homogeneous composite structure of the SNF system, finite element analyses (FEA) are needed to translate the global moment-curvature measurement into local stress-strain profiles. The detailed mechanisms of the pellet-pellet and pellet-clad interactions and the stress concentration effects at the pellet-pellet interface cannot be readily obtained directly from a CIRFT system measurement. Therefore, detailed FEA is used to understand the global test response. The major findings of CIRFT test on the high burn-up SNF are as follows: -) SNF system interface bonding plays an important role in SNF vibration performance; -) Fuel structure contributes to the SNF system stiffness; -) There are significant variations in stress and curvature of SNF systems during vibration cycles resulting from segment pellets and clad interaction, and -) SNF failure initiates at the pellet-pellet interface region and appears to be spontaneous.

42 ENGINEERING↗
DOE OSTI2018

Review of V&V Documentation

Consortium for Advance Simulation of Light Water Reactors (CASL) is a Department of Energy Innovation Hub whose mission is the following, "CASL is a collaboration of the nation's leading scientists, institutions, and supercomputers, with an aggressive 10-year mission to confidently predict the performance of existing and next-generation commercial nuclear reactors through comprehensive, science-based modeling and simulation." The CASL program to date has focused on developing the necessary predictive capability. Rightly so, it is characterized by many as a research project. As a matter of intent, the first 6 years of CASL focused on developing and demonstrating the prediction capability of a suite of independent physics codes: MPACT (neutronics), CTH (thermal hydraulics in the core), BISON (fuel performance), and MAMBA (CRUD and boron uptake on fuel rod surfaces) The last 4 years focused on initial attempts to couple the codes and to demonstrate those capabilities through a series of challenge problems aligned to 3 key issues of interest to the nuclear power industry.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

Grid-to-rod fretting wear study of SiC/SiC composite accident-tolerant fuel claddings using an autoclave fretting bench test

Grid-to-rod-fretting (GTRF) in pressurized water reactors (PWRs) is known to cause wear and surface damage on the fuel claddings, potentially leading to radioactive leakage. One of the accident-tolerant fuel (ATF) concepts is to use advanced cladding materials that could withstand higher temperatures. Here, we investigated the wear behavior of candidate silicon carbide (SiC)-based composite claddings with different levels of surface finish in fretting against a commercial ZIRLO alloy grid using a unique bench-scale autoclave GTRF rig. The experiments mimicked the environment in an industrial full-assembly PWR simulator. Fretting tests were conducted with a realistic load (~0.5 N) in deionized water under a pressure of 20–23 bar at 204 °C for 100 h. While the SiC/SiC composite claddings showed significantly higher wear resistance than the commercial ZIRLO alloy cladding as expected, the smoother versions experienced surprisingly higher wear than the much softer counterface, ZIRLO grid. The wear mechanism of the SiC/SiC cladding was attributed to the SiC wear debris that was trapped at the fretting interface causing both 3-body and 2-body (embedded into the grid surface) abrasion of the cladding. Rougher SiC/SiC claddings had less material loss but caused more wear on the ZIRLO grid. Finally, pre-oxidized ZIRLO grid showed better compatibility with the SiC/SiC cladding to protect both the cladding and grid as a result of reduced wear debris trapping.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Thermal and hydraulic performance of volumetrically heated triply periodic minimal surface heaters

Advanced heat transfer surfaces, mainly Triply Periodic Minimal Surfaces (TPMSs) have great potential to increase heat transfer performance over traditional heat transfer surfaces due to their tortuous flow path and higher surface area-volume ratio. The emergence of additive manufacturing of nuclear fuel motivates experimental investigation of such advanced geometries in the context of a nuclear reactor core. Here, this study employed a novel method to explore the performance of gyroid and diamond TPMS geometries. Volumetrically heated heaters were additively manufactured from conductive polymer filament, and airflow was used to simulate the convective heat transfer from the core within a nuclear reactor. The mass flow and volumetric generation rates were varied, local Nusselt number correlations were obtained using embedded thermocouples to measure the solid and fluid temperature, and friction factor correlations were developed using differential pressure measurements. The gyroid TPMS presented a friction factor 1.5 times higher than the diamond TPMS and 90 times that of the rod bundle. TPMS geometries showed Nusselt numbers 8–10 times higher than that of the rod bundle, also maintaining colder internal temperatures, which suggested that they could sustain 250–275 % higher power density than the rod bundle for a given centerline temperature. Hence, having the potential to greatly increase the safety margins and power output of a nuclear reactor core with similar volume.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Coupled Calculations in Helical Steam Generator: Validation on Legacy Data

The NEAMS program aims to develop an integrated multi-physics simulation capability “pellet-to-plant” for the design and analysis of future generations of nuclear power plants. In particular, the Reactor Product Line code suite's multi-resolution hierarchy is being designed to ultimately span the full range of length and time scales present in relevant reactor design and safety analyses, as well as scale from desktop to petaflop computing platforms. Flow-induced vibration (FIV) is widespread problem in energy systems because they rely on fluid movement for energy conversion. Vibrating structures may be damaged as fatigue or wear occurs. Given the importance of reliable components in the nuclear industry, flow-induced vibration has long been a major concern in safety and operation of nuclear reactors. In particular, nuclear fuel rods and steam generators have been known to suffer from flow-induced vibration and related failures. Advanced reactors, such as integral Pressurized Water Reactors (PWRs) considered for Small Modular Reactors (SMR), often rely on innovative component designs to meet cost and safety targets. One component that is the subject of advanced designs is the steam generator, some designs of which forego the usual shell-and-tube architecture in order to fit within the primary vessel. In addition to being more cost- and space-efficient, such steam generators need to be more reliable, since failure of the primary vessel represents a potential loss of coolant and a safety concern. A significant amount of data exists on flow-induced vibration in shell-and-tube heat exchangers, and heuristic methods are available to predict their occurrence based on a set of given assumptions. In contrast, advanced designs have far less data available. Advanced modeling and simulation based on coupled structural and fluid simulations have the potential to predict flow-induced vibration in a variety of designs, reducing the need for expensive experimental programs, especially at the design stage. Over the past five years, the Reactor Product Line has developed the integrated multi-physics code suite SHARP. The goal of developing such a tool is to perform multi-physics neutronics, thermal/fluid, and structural mechanics modeling of the components inside the full reactor core or portions of it with a user-specified fidelity. In particular SHARP contains high-fidelity single-physics codes Diablo for structural mechanics and Nek5000 for fluid mechanics calculations. Both codes are state-of-the-art, highly scalable tools that have been extensively validated. These tools form a strong basis on which to build a flow-induced vibration modeling capability. In this report we discuss one-way coupled calculations performed with Nek5000 and Diablo aimed at simulating available FIV experiments in helical steam generators in the turbulent buffeting regime. In this regime one-way coupling is judged sufficient because the pressure loads do not cause substantial displacements. It is also the most common source of vibration in helical steam generators at the low flows expected in integral PWRs. The legacy data is obtained from two datasets developed at Argonne and B&W.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Demonstration of Fully Coupled Calculations in Helical Steam Generator: Toward Predictions of Fluid Elastic Instability

The NEAMS program aims to develop an integrated multi-physics simulation capability “pellet-to-plant” for the design and analysis of future generations of nuclear power plants. In particular, the Reactor Product Line code suite's multi-resolution hierarchy is being designed to ultimately span the full range of length and time scales present in relevant reactor design and safety analyses, as well as scale from desktop to petaflop computing platforms. Flow-induced vibration (FIV) is widespread problem in energy systems because they rely on fluid movement for energy conversion. Vibrating structures may be damaged as fatigue or wear occurs. Given the importance of reliable components in the nuclear industry, flow-induced vibration has long been a major concern in safety and operation of nuclear reactors. In particular, nuclear fuel rods and steam generators have been known to suffer from flow-induced vibration and related failures. Advanced reactors, such as integral Pressurized Water Reactors (PWRs) considered for Small Modular Reactors (SMR), often rely on innovative component designs to meet cost and safety targets. One component that is the subject of advanced designs is the steam generator, some designs of which forego the usual shell-and-tube architecture in order to fit within the primary vessel. In addition to being more cost- and space-efficient, such steam generators need to be more reliable, since failure of the primary vessel represents a potential loss of coolant and a safety concern. A significant amount of data exists on flow-induced vibration in shell-and-tube heat exchangers, and heuristic methods are available to predict their occurrence based on a set of given assumptions. In contrast, advanced designs have far less data available. Advanced modeling and simulation based on coupled structural and fluid simulations have the potential to predict flow-induced vibration in a variety of designs, reducing the need for expensive experimental programs, especially at the design stage. Over the past seven years, the Reactor Product Line has developed the integrated multi-physics code suite SHARP. The goal of developing such a tool is to perform multi-physics neutronics, thermal/fluid, and structural mechanics modeling of the components inside the full reactor core or portions of it with a user-specified fidelity. In particular SHARP contains high-fidelity single-physics codes Diablo for structural mechanics and Nek5000 for fluid mechanics calculations. Both codes are state-of-the-art, highly scalable tools that have been extensively validated. These tools form a strong basis on which to build a flow-induced vibration modeling capability. In the last two years a significant effort was devoted to 1) develop and validate a fully coupled capability in SHARP for Fluid-Structure calculations (FSI) and 2) validate the use of SHARP for one-way coupled calculations in helical steam generators on legacy data available as well as novel experiments. In this report we discuss an extension of previous one-way coupled calculations performed with Nek5000 and Diablo aimed at simulating available FIV experiments in helical steam generators in the turbulent buffeting regime. It was demonstrated that in the buffeting regime one-way coupling was judged sufficient because the pressure loads do not cause substantial displacements. However, higher speeds led to a reduction in accuracy and the need for a fully coupled capability for velocity or approaching the critical velocity and fluid elastic instability. In this report we demonstrate such fully coupled capability on the Argonne test case discussed in previous reports.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Containment Analysis Capability of UNF-ST&DARDS

The Used Nuclear Fuel-Storage, Transportation & Disposal Analysis Resource and Data System (UNF-ST&DARDS) methodology to perform automated containment analyses for potential transportation packages based on canister loading map information is described in this paper, and its capability is illustrated with example results. The allowable leakage rate is calculated with the procedures provided in ANSI N14.5-2014 and NUREG/CR-6487, which were adapted for containment analysis of a transportation package containing fuel assemblies with different nuclear characteristics (e.g., initial enrichment, burnup, and cooling time) and clad integrity (intact or damaged). UNF-ST&DARDS applies different source term calculation methodologies for low-burnup fuel (LBF) (i.e., <45 GWd/tonne U) assemblies and high-burnup fuel (HBF) (i.e., ≥45 GWd/tonne U) assemblies. The LBF radionuclide activities are based on actual fuel assembly burnup, initial enrichment, and cooling time. Bounding radionuclide activities based on a fuel pellet burnup value of 65 GWd/tonne U and actual fuel assembly cooling time are used for HBF assemblies. The fraction of failed fuel rods and the release fractions for the contributors to releasable source terms recommended in NUREG-1617 are used in the containment analysis regardless of fuel assembly burnup. However, UNF-ST&DARDS supports different parameter values for LBF and HBF assemblies. Finally, the containment analysis methodology for as-loaded transportation packages is presented in detail, and the UNF-ST&DARDS containment analysis capability is illustrated with results for simulated transportation packages containing spent nuclear fuel canisters in dry storage at selected sites.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Predicting the flow stress of Zircaloy-4 under in-reactor accident conditions - 2016-0075

During loss of coolant accident (LOCA) and reactivity-initiated accident (RIA), nuclear fuel rods experience high heating rates that change the microstructure and properties of zirconium cladding materials. The aim was to determine how different fast heating rates affect the kinetics of transformation and whether its consequences on the material yield stresses in the dual-phase region could be predicted. Phase fraction and texture of cold-rolled Zircaloy-4 were thus measured at fast heating rates similar to those experienced during LOCA and RIA by in-situ high-energy synchrotron x-ray diffraction; the yield stress was also measured at different heating rates to the same dual-phase temperature by electrical resistance method. Crystal plasticity finite element model was used to simulate the flow stresses with the inputs of the measured phase fraction and texture. The model is calibrated using the flow stress at the heating rate of 10 deg. Cs{sup -1} and used to predict the flow stress at the heating rate of 50 deg. Cs{sup -1} at the same temperature. A significant shift of the β transus temperature was identified for fast heating rates in the range of 10-100 deg. Cs{sup -1}. There is great similarity in texture at different heating rates. In the dual-phase temperature region, the a texture is almost identical to the starting texture and the β texture is nearly random. Yield stresses of cold-rolled Zircaloy-4 decrease at faster heating rate at the same temperature of 920 deg. C. A good agreement was found between the experimental and predicted flow stress at different heating rates, indicating this methodology has potential to be used to predict flow behavior in a fast transient regime. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2025

High-Fidelity CFD Assessments of Flow Resistance in a 61-Pin Wire-Wrapped Assembly with Partially Blocked Channels

The examination of thermal-hydraulic behaviors in wire-wrapped rod bundles continues to be an active area of research. The sodium fast reactor, a prominent candidate in next-generation nuclear designs, utilizes a hexagonal configuration of wire-wrapped fuel pins. Here, the potential for channel blockage within this compact arrangement poses a significant safety challenge, spurring a number of recent experimental and computational investigations to evaluate its impact on flow and heat transfer. The present work aims to benchmark the high-fidelity NekRS computational fluid dynamics (CFD) solver in predicting the pressure drops associated with substantial blockages, using available experimental data as a reference. A 61-pin wire-wrapped fuel assembly with two flow blockage configurations has been simulated and investigated at a range of low to moderate Reynolds numbers (487 ≤ Re ≤ 14 600). The NekRS solver demonstrates an exponential reduction of spatial discretization error with increasing polynomial order. The high level of agreement between the numerical results and measured data confirms the accuracy and consistency of the present numerical approach. This benchmark study establishes the capability of NekRS to perform reliable hydrodynamic simulations for sodium fast reactor applications and supports its use in design, licensing, and safety analyses.

CFD Benchmarking↗
DOE OSTI2025

A strategy for automated core design to increase economic viability and minimize fuel fragmentation, relocation, and dispersal susceptibility in high-burnup cores

The nuclear industry aims to increase the cycle length of pressurized water reactors from 18 to 24 months to increase power plant capacity factors and economic viability. These cycle length extensions will inherently require fuel rods to exceed the current peak rod average burnup limit of 62 GWd/MTU. A chief concern of operating beyond the current burnup limit is the fuel fragmentation, relocation, and dispersal (FFRD) phenomenon in which pulverized fuel fragments can axially relocate and escape through a burst in the cladding formed during a loss-of-coolant accident. In this work, we demonstrate an approach for automating core design employing an optimization tool based on a penalty-free, parallel simulated annealing algorithm to produce pressurized water reactor core designs with two different optimization objectives. The two objectives were to produce core designs with (1) mitigated FFRD susceptibility while achieving 24-month cycle lengths (2) maximum cycle length with no regard for the likelihood of FFRD. Batch size was considered in tandem with both cases to maximize economic viability. The PARCS nodal model was the primary reactor physics tool used in the optimizations and used nuclear cross sections calculated with 2D Polaris lattice physics models. Reactor performance and safety characteristics of the optimized cores were verified using high-fidelity Virtual Environment for Reactor Applications models. The core designs produced by the optimization tool are compared with each other and to a high-burnup core design produced and analyzed in previous works to highlight the fuel management strategies that may enhance high-burnup reactor safety and economic viability. The optimized cores satisfied their respective objective functions, producing a maximum cycle length of 720 effective full-power days in one core design and one that may reduce FFRD susceptibility by up to 50% based on the first-order approximation to FFRD risk formulated in this work. The optimized cores met most constraints but exceeded the hot channel factor limit, especially in FFRD cases where fresh fuel carried more power. Furthermore, this highlights the need for future lattice-level optimizations and broader assembly options.

Cycle length↗
DOE OSTI2017

SuPer-Homogenization (SPH) Corrected Cross Section Generation for High Temperature Reactor

The deterministic full core simulators require homogenized group constants covering the operating and transient conditions over the entire lifetime. Traditionally, the homogenized group constants are generated using lattice physics code over an assembly or block in the case of prismatic high temperature reactors (HTR). For the case of strong absorbers that causes strong local depressions on the flux profile require special techniques during homogenization over a large volume. Fuel blocks with burnable poisons or control rod blocks are example of such cases. Over past several decades, there have been a tremendous number of studies performed for improving the accuracy of full-core calculations through the homogenization procedure. However, those studies were mostly performed for light water reactor (LWR) analyses, thus, may not be directly applicable to advanced thermal reactors such as HTRs. This report presents the application of SuPer-Homogenization correction method to a hypothetical HTR core.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Predicting the flow stress of Zircaloy-4 under in-reactor accident conditions - 2016-0075

During loss-of-coolant accident (LOCA) and reactivity-initiated accident (RIA), nuclear fuel rods experience high heating rates that change the microstructure and properties of zirconium cladding materials. We sought to determine how different fast heating rates affect the kinetics of transformation and whether their consequences on the material yield stresses in the dual-phase region could be predicted. Phase fraction and texture of cold-rolled Zircaloy-4 were thus measured at fast heating rates similar to those experienced during LOCA and RIA by in situ high-energy synchrotron X-ray diffraction; the yield stress was also measured at different heating rates to the same dual-phase temperature by the electrical resistance method. A crystal plasticity finite-element model was used to simulate the flow stresses with the inputs of the measured phase fraction and texture. The model was calibrated using the flow stress at the heating rate of 10 deg. C s{sup -1} and used to predict the flow stress at the heating rate of 50 deg. C s{sup -1} at the same temperature. A significant shift of the β transus temperature was identified for fast heating rates in the range of 10 to 100 deg. C s{sup -1}. There was great similarity in texture at different heating rates. In the dual-phase temperature region, the α texture was almost identical to the starting texture, and the b texture was nearly random. Yield stresses of cold-rolled Zircaloy-4 decreased at a faster heating rate at the same temperature of 920 deg. C. A good agreement was found between the experimental and predicted flow stress at different heating rates, indicating this methodology can potentially be used to predict flow behavior in a fast transient regime. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2024

Leveraging Optimal Sparse Sensor Placement to Aggregate a Network of Digital Twins for Nuclear Subsystems

Nuclear power plants (NPPs) require continuous monitoring of various systems, structures, and components to ensure safe and efficient operations. The critical safety testing of new fuel compositions and the analysis of the effects of power transients on core temperatures can be achieved through modeling and simulations. They capture the dynamics of the physical phenomenon associated with failure modes and facilitate the creation of digital twins (DTs). Accurate reconstruction of fields of interest (e.g., temperature, pressure, velocity) from sensor measurements is crucial to establish a two-way communication between physical experiments and models. Sensor placement is highly constrained in most nuclear subsystems due to challenging operating conditions and inherent spatial limitations. This study develops optimized data-driven sensor placements for full-field reconstruction within reactor and steam generator subsystems of NPPs. Optimized constrained sensors reconstruct field of interest within a tri-structural isotropic (TRISO) fuel irradiation experiment, a lumped parameter model of a nuclear fuel test rod and a steam generator. The optimization procedure leverages reduced-order models of flow physics to provide a highly accurate full-field reconstruction of responses of interest, noise-induced uncertainty quantification and physically feasible sensor locations. Accurate sensor-based reconstructions establish a foundation for the digital twinning of subsystems, culminating in a comprehensive DT aggregate of an NPP.

42 ENGINEERING↗
DOE OSTI2017

Improved Safety Injection Flow Map Associated with Target RCS Depressurization to Maintain Core Coolability of OPR1000

In Korea, Severe Accident Management Guidance (SAMG) for Optimized Power Reactor (OPR1000) has been developed through the research results related to the WOG SAMG, in which the operation phase shifts from the Emergency Operation Procedure (EOP) to the SAMG when the Core Exit Temperature (CET) reaches 923 K. Among 7 mitigation strategies envisioned, the third management strategy, injecting coolant into Reactor Coolant System (RCS) by depressurizing the RCS is one of the key strategies to fortify the in-vessel retention capability. As an aid to such mitigation strategy, a safety injection (SI) flow map has been developed by Lee et al. for predicting required flow rate associated with target RCS depressurization to recover in-vessel coolability. The mechanistic SI flow map used the CET increase rate and core water level decrease rate obtained from MELCOR code simulation. It should be noted that during the severe accident, the total heat generated is sum of the decay heat and the oxidation heat. Thus, it is necessary to consider additional heat sources by oxidation reaction between cladding, structure, control rod, and hot steam prior to the injection of coolant. It was believed that the SI model estimated required flow rate rather conservatively by considering the expected amount decay heat and the accumulated heat in the core based on the RCS pressure. However, it was confirmed that the SI model developed by Lee et al. could not represent the total heat removal especially since start of oxidation. Under transient conditions, the heat sources accumulated in the heat structures (fuel, cladding, supporting structures) should be embarked in the model for accurate contribution of the total heat source to the required flow rate. Therefore, the objective of this study is to improve the previous SI flow map predicting the injection flow rate associated with target RCS depressurization to satisfy the in-vessel coolability during severe accidents. Additional terms were developed to reflect accurate heat accumulating and included in the existing model. Resulting SI flow map includes the target RCS depressurization, which enables discharge of the required flow rate by utilizing the flow rate-pressure curve with operation of two high pressure safety injections (HPSIs). Also recent MELCOR simulation result of a postulated steam generator tube rupture (SGTR) accident verified the accuracy of the improved map.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Results of TRIPOLI-4 version 10 on the TRIPOLI-4 criticality validation suite, thermal spectra

This paper presents the results of the latest version (version 10) of TRIPOLI-4 on its criticality validation suite, thermal spectra part. The criticality validation suite is made of 148 selected ICSBEP benchmarks with various enrichments, fissile materials and spectra. The ICSBEP benchmarks computational models are prepared by the TRIPOLI-4 team and are run in the same conditions for each benchmark: initial source, number of neutrons per cycle, number of inactive and active cycles, k{sub eff} estimator and nuclear data. Comparisons are made with experimental k{sub eff}. On one case (PST-009), we have noted a discrepancy between the computational results and the experiments. This discrepancy is also found with other codes, which might indicate a possible issue with the benchmark. For all the other considered cases, uranium and plutonium solutions and fuel-rods-arrays, the agreement with the experiments is very good.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Fuel Cycle Performance of Thermal Spectrum Small Modular Reactors

Small modular reactors may offer potential benefits, such as enhanced operational flexibility. However, it is vital to understand the holistic impact of small modular reactors on the nuclear fuel cycle and fuel cycle performance. The focus of this paper is on the fuel cycle impacts of light water small modular reactors in a once-through fuel cycle with low-enriched uranium fuel. A key objective of this paper is to describe preliminary reactor core physics and fuel cycle analyses conducted in support of the U.S. Department of Energy Office of Nuclear Energy Fuel Cycle Options Campaign. Challenges with small modular reactors include: increased neutron leakage, fewer assemblies in the core (and therefore fewer degrees of freedom in the core design), complex enrichment and burnable absorber loadings, full power operation with inserted control rods, the potential for frequent load-following operation, and shortened core height. Each of these will impact the achievable discharge burn-up in the reactor and the fuel cycle performance. This paper summarizes the results of an expert elicitation focused on developing a list of the factors relevant to small modular reactor fuel, core, and operation that will impact fuel cycle performance. Preliminary scoping analyses were performed using a regulatory-grade reactor core simulator. The hypothetical light water small modular reactor considered in these preliminary scoping studies is a cartridge type one-batch core with 4.9% enrichment. Some core parameters, such as the size of the reactor and general assembly layout, are similar to an example small modular reactor concept from industry. The high-level issues identified and preliminary scoping calculations in this paper are intended to inform on potential fuel cycle impacts of one-batch thermal spectrum SMRs. In particular, this paper highlights the impact of increased neutron leakage and reduced number of batches on the achievable burn-up of the reactor. Fuel cycle performance metrics for a small modular reactor are compared to a conventional three-batch light water reactor in the following areas: nuclear waste management, environmental impact, and resource utilization. Metrics performance for a small modular reactor are degraded for mass of spent nuclear fuel and high level waste disposed, mass of depleted uranium disposed, land use per energy generated, and carbon emission per energy generated

Worrall, Andrew↗
DOE OSTI2024

Model Validation and Uncertainty Quantification on the KRUSTY Microreactor Design Using GRIFFIN Neutron Transport Code [Poster]

Argonne National Laboratory (ANL) and INL have developed a GRIFFIN steady state neutronics model for the multiphysics simulations of the Kilopower Reactor Using Sterling TechnologY (KRUSTY) microreactor in the Multiphysics Object Oriented Simulation Environment (MOOSE). The reliability of such deterministic neutronics models can be validated by comparing with computations from Monte Carlo codes (e.g. MCNP, SERPENT, OpenMC, Shift, etc). Furthermore, potential modeling/design improvements can be identified by incorporating uncertainty quantification (UQ), which can be performed by MOOSE’s Stochastic Tools Module (STM). KRUSTY is a prototype for a 5-kW thermal nuclear-powered space reactor. Its primary components consist of nuclear fuel, heat pipes, a control rod, a reflector, and the shielding. The fuel consists of 3 stacked U-7.65Mo cylinders with a hole in the center for the control rod. 8 liquid sodium heat pipes transfer fission energy from the solid fuel block to the Sterling power conversion system where the energy is extracted, and the cooled sodium flows back to the core via capillary action . The movable Boron Carbide control rod regulates the neutron population during startup or when a reactor temperature boost is needed . The beryllium oxide reflector is in 3 places in the reactor; it surrounds the core axially, it lies beneath the core on a platen, and it is present in the shim. The axial and lower reflectors rest on an adjustable stainless-steel platen that moves upward to cover the fuel and help the reactor reach criticality. Lastly, radial stainless steel surrounds the core offering protection from radiation exposure .

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Determination of He and D permeability of neutron-irradiated SiC tubes to examine the potential for release due to micro-cracking

Driven by the need to enlarge the safety margins of light water reactors in both design-basis and beyond-design-basis accident scenarios, the research and development of accident-tolerant fuel (ATF) has become an importance topic in the nuclear engineering and materials community. Continuous SiC fiber-reinforced SiC matrix ceramic composites are under consideration as a replacement for traditional zirconium alloy cladding owing to their high-temperature stability, chemical inertness, and exceptional irradiation resistance. Among the key technical feasibility issues, potential failure of the fission product containment due to probabilistic penetrating cracking has been identified as one of the two most critical feasibility issues, together with the radiolysisassisted hydrothermal corrosion of SiC. The experimental capability to evaluate the hermeticity of SiC-based claddings is an urgent need. In this report, we present the development of a comprehensive permeation testing station established in the Low Activation Materials Development and Analysis laboratory at Oak Ridge National Laboratory. Preliminary results for the hermeticity evaluation of un-irradiated monolithic SiC tubes, uncoated and coated SiC/SiC composite tubes, and neutron-irradiated monolithic SiC tubes at room temperature are exhibited. The results indicate that this new permeation testing station is capable of evaluating the hermeticity of SiC-based tubes by determining the helium and deuterium permeation flux as a function of gas pressure at a high resolution of 8.07 x 10 -12 atm-cc/s for helium and 2.83 x 10 -12 atm-cc/s for deuterium, respectively. The detection limit of this system is sufficient to evaluate the maximum allowable helium leakage rate of lab-scale tubular samples, which is linearly extrapolated from the evaluation standard used for a commercial as-manufactured light water reactor fuel rod at room temperature. The un-irradiated monolithic SiC tube is hermetic, as is manifested by the un-detectable deuterium permeation flux at various feeding gas pressures. A large helium leakage rate was detected for the uncoated SiC/SiC composite tube exposed to atmosphere, indicating it is inherently not hermetic. The hermeticity of coated SiC/SiC composite tubes is strongly dependent on the coating materials and the preparation of the substrate SiC/SiC composite samples. To simulate the practical application environment, monolithic CVD SiC tubes were exposed to neutron irradiation at the High Flux Isotope Reactor under high heat flux from the internal surface to the external surface. Although finite element analysis and resonant ultrasound spectroscopy measurement indicated that the combined neutron irradiation and high heat flux gave rise to a high probability of cracking within the sample, the hermeticity evaluation of the tested sample still exhibited gas tightness, emphasizing that SiC cracking is inherently a statistical phenomenon. The developed permeation testing station is capable of measuring the gas permeation flux in the range of interest with full confidence based on the presented results. It is considered a critical pre- /post-irradiation examination technique to characterize SiC-based cladding materials in asreceived and irradiated states to aid the research and development of ATF.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗