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At least 271 records · Page 15
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Prediction of The TREAT Facility Power and Temperature of Sirius-2c Nuclear Propulsion Fuel Experiment

This paper presents simulation results of the Sirius-2c full power experiment that was performed at the TREAT facility to support experiment modeling and transient analysis of the NASA-sponsored Sirius series of experiments for nuclear thermal propulsion applications. We presents a novel approach to predict the control rod motion and power of the reactor and the temperature of the specimen during transient experiments using the Griffin and Bison packages of the Multiphysics Object Oriented Simulation Environment (MOOSE). First, the model predicts the transient rod motion that results in the desired reactor power shape. Then, the temperature of the specimen is predicted based on the power deposited in the specimen. The prediction results are in good agreement with the measured reactor power and control rod position and over-predicted the specimen temperature which is mainly related to specimen surface emissivities. The reactor reaches a peak power of 80 MW and a specimen temperature of 2500 K and considering the right specimen emissivities the predicted temperature is in an excellent agreement with the measured values. This model can be utilized in helping and supporting the design and optimization a new experiments that are going to be irradiated in the TREAT facility.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2016

Position-Dependency of Fuel Assembly Homogenization in a Pressurized Water Reactor

2-step nuclear reactor core analysis has long been established as the conventional calculational procedure for the pressurized water reactor (PWR) due to its efficiency. This is because the procedure significantly avoids the potentially huge computational burden arising from an explicit geometrical modeling and consideration of over ∼ 500 energy groups. It instead approaches the problem with spatial homogenization and group condensation: a 2-D lattice transport calculation to homogenize the fuel assembly cross-sections and a 3-D whole core nodal diffusion. One notes that the few-group constant generation using deterministic method could be complicated due to its limited flexibilities in geometrical modeling and inherent drawbacks of the multi-group approximation. Furthermore, accuracy of the deterministic few-group constant generation can especially be compromised during the resonance treatment calculation. As such, an alternative lattice calculation using the Monte Carlo (MC) approach, in place of the deterministic fuel assembly homogenization (few-group constant generation), has previously been proposed. During the assembly homogenization, directions and magnitudes of the current at boundaries are unknowns. Therefore, zero net current boundary conditions (infinite lattice) are instead used as the assembly node. The assumption is quite reasonable considering heterogeneities inside the node are more dominant than heterogeneities due to nodal location in the core. In fact, this approximation has successfully been applied to many LWR core analyses to reasonable accuracy. It should nonetheless be noted that the net current zero boundary assumption effectively ignores position-dependency of the fuel assembly homogenization in the general 2-step method. As such, there are a number of well-known researches on the extended assembly homogenization techniques, which introduce correction factors to compensate effects of the assembly environment, such as Generalized Equivalence Theory (GET), Super-homogenization method (SPH), and color set calculation method. Even though the assembly environment effect has long been acknowledged, its effect has not properly been investigated due to the limited computing power in the past and due to the reasonable accuracy of the extended assembly homogenization techniques. Nonetheless, quantification of position-dependency of the assembly homogenization in the core could still potentially offer additional value in terms of calculation accuracy to the 2-step reactor analysis. It is upon this motivation that the paper is prepared. In this research, position-dependent few-group constants were calculated and compared against those of infinite lattices. All calculations were completed using the MC Serpent2 code with ENDF/B7.0 nuclear data library. In this study, position-dependency of the fuel assembly homogenized cross sections in a PWR core has been investigated through comparison of the conventional 2-step reactor analysis with the reference high fidelity Monte Carlo simulation. It was found that the homogenized two-group constants are clearly position-dependent as the discrepancies between infinite lattice and whole-core calculations are quite significant especially in a rodded assembly. As such, the corresponding correction required due to the assembly environment effect is not negligible compared to the correction required due to assembly heterogeneity itself. It is therefore quite clear that position-dependency of the group cross sections should be considered in the 2-step procedure for a more accurate PWR core analysis. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2025

Design and full core fuel performance assessment of high burnup cores for 4-loop PWRs

Increasing the fuel discharge burnup of current light water reactors (LWRs) promises reductions in fuel cycle and/or operations costs. By assuming a constant core power density, the economic gain is enabled by better fuel utilization and/or an increased capacity factor. In this effort to investigate greater than 62 MWd/kgU maximum rod average burnup for 110+ kW/l core power density, two core designs have been developed for a standard 17x17, 193 fuel assemblies pressurized water reactor (PWR). The levelized unit cost methodology is employed to evaluate fuel cycle, operation and maintenance, and capital cost impacts and to examine the economic viability of both core design pathways. Core design and optimization are performed using the commercial STUDSVIK code package. Fuel performance analysis is realized in full core configuration via auditing FRAPCON4.1, FAST1.2, and the high-fidelity code BISON. To provide a realistic assessment, the core design process takes into consideration best practices in current PWR core design. It features acceptable performance in terms of various core design constraints on maximum allowable peaking and boron concentration. Gadolinia (Gd2O3) is used as a burnable poison with a maximum of 9 wt% concentration while feeding 89 or 77 fuel assemblies in a 3-batch refueling scheme. Full core fuel performance simulation, which allows for characterization of relevant fuel temperatures, plenum pressures, stresses, and strains, is performed with respect to two bounding burnup levels. Such performance is potentially licensable for the 18-month high burnup core (<68 MWd/kgU peak pin), while it is more challenging for the 24-month high burnup core design pathway (<75 MWd/kgU peak pin). Maximum rod plenum pressure is identified as the most limiting fuel performance parameter. Here, while the scope of the present study focuses on the steady-state plus overpower conditions, the acceptability of the new discharge burnup has to be further assessed by considering uncertainties and impacts under accident scenarios in the future.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Fiscal Year (FY) 2017 Activities for the Spent Fuel Nondestructive Assay Project

The main focus of research in the NA-241 spent fuel nondestructive assay (NDA) project in FY17 has been completing the fabrication and testing of two prototype instruments for upcoming spent fuel measurements at the Clab interim storage facility in Sweden. One is a passive instrument: Differential Die-away Self Interrogation-Passive Neutron Albedo Reactivity (DDSI), and one is an active instrument: Differential Die-Away-Californium Interrogation with Prompt Neutron (DDA). DDSI was fabricated and tested with fresh fuel at Los Alamos National Laboratory in FY15 and FY16, then shipped to Sweden at the beginning of FY17. Research was performed in FY17 to simplify results from the data acquisition system, which is complex because signals from 56 different 3He detectors must be processed using list mode data. The DDA instrument was fabricated at the end of FY16. New high count rate electronics better suited for a spent fuel environment (i.e., KM-200 preamplifiers) were built specifically for this instrument in FY17, and new Tygon tubing to house electrical cables was purchased and installed. Fresh fuel tests using the DDA instrument with numerous configurations of fuel rods containing depleted uranium (DU), low enriched uranium (LEU), and LEU with burnable poisons (Gd) were successfully performed and compared to simulations.1 Additionally, members of the spent fuel NDA project team travelled to Sweden for a “spent fuel characterization and decay heat” workshop involving simulations of spent fuel and analysis of uncertainties in decay heat calculations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Reactivity-initiated accident simulation to inform transient testing of candidate advanced cladding

This paper leverages three-dimensional reactor-core-simulation capabilities to provide information on appropriate experimental test conditions for candidate advanced cladding materials in a control-rod-ejection event. These test conditions use three-dimensional nodal kinetics simulations of a reactivity-initiated accident (RIA) in a representative state-of-the-art pressurized water reactor with both nuclear-grade iron-chromium-aluminum (FeCrAl) and silicon-carbide (SiC-SiC)-based cladding materials. The effort yields boundary conditions for experimental mechanical tests, specifically peak cladding strain during the power pulse following the rod ejection. The impact of candidate cladding materials on the reactor kinetics behavior of RIA progression versus that of reference Zr cladding is predominantly due to differences in -1) fuel mass/volume/specific power density, -2) spectral effects due to parasitic neutron absorption, -3) control rod worth due to hardened (or softened) spectrum, and -4) initial conditions due to power peaking and neutron transport cross sections in the equilibrium cycle cores resulting from hardened (or softened) spectrum. This study shows minimal impact of SiC-SiC-based cladding configurations on the transient response versus reference Zircaloy cladding. However, the FeCrAl cladding response indicates similar energy deposition, but with significantly shorter pulses of higher magnitude. Therefore, the FeCrAl-based cases have a more rapid fuel thermal expansion rate, and the resultant pellet-cladding interaction occurs more rapidly. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

Sodium-Cooled Fast Reactor Reference Plant Model

This report details the progress of Idaho National Laboratory (INL) in creating a reference plant multiphysics model for the Advanced Burner Test Reactor (ABTR). This model was developed under Task 13 of the U.S. Nuclear Regulatory Commission project “Development and Modeling Support for Advanced Non-Light Water Reactors,” and is an extension of the reference plant model developed in Task 4b, which was improved upon in the following ways. (1) The discrete ordinates method was used in lieu of the super-homogenization (SPH)-corrected diffusion approximation in order to better capture the anisotropic scattering contribution and the neutron leakage change due to thermal expansion. (2) The novel neutronic spatial discretization approach, termed the ring-heterogeneous (RH) approximation, was conceptualized and introduced to capture the differential expansion of the materials in the core. This new technique proved capable of preserving fission rates and maintaining the eigenvalue within 2.5% and 266 pcm with 9 neutron energy groups, respectively. Separating the different materials in the core enables the differential expansion of materials to be explicitly accounted for, eliminating the need for problem-specific cross-section functionalization techniques. (3) The SAM model for the core and system thermal-hydraulics analysis was updated to include 61 channels instead of just four representative ones. This enables users to obtain improved spatial resolution for sodium temperature and density scalar fields. (4) All the mesh files were created via the Multiphysics Object-Oriented Simulation Environment (MOOSE) Reactor module, eliminating all reliance on external tools for mesh creation. (5) Finally, the fuel axial expansion now leverages the HT9 and UPuZr material properties that have been validated against experimental data. The reference plant model was used to perform a full-core unprotected loss of flow (ULOF) transient calculation, including neutronics, thermal and mechanical feedback mechanisms. Future work will be devoted to further enhancements of the model. Potential improvements to the model include the addition of the control rod driveline expansion feedback and the upgrading of the support plate model so as to explicitly include 3D effects. Additionally, a Nuclear Energy Advanced Modeling and Simulation funded parallel effort has completely automated the creation of the ring-heterogeneous (RH) mesh from the fully heterogeneous (FH) geometry, thus maximizing user friendliness for the sodium fast reactor sodium-cooled fast reactor (SFR) workflow and will be incorporated in future work.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

An Overview of Material Models in the BISON Fuel Performance Code

BISON is a modern finite element-based nuclear fuel performance code that has been under development at Idaho National Laboratory (INL) since 2009. The code is applicable to both steady and transient fuel behavior and supports the use of one-, two-, and three-dimensional meshes. BISON is designed to be a multi-fuels code. Although primarily applied to Light Water Reactor (LWR) fuel to date, the code has been used to analyze TRISO-coated particle fuel and metal fuel in rod and plate form, design and interpret fuel irradiation experiments and investigate novel fuel concepts. Nuclear fuel operates in an extreme environment that induces complex multiphysics phenomena occurring over distances ranging from inter-atomic spacing to meters, and times scales ranging from microseconds to years. To simulate this behavior requires a wide variety of material models which are often complex and nonlinear. Nearly all of the material models currently available in BISON are empirical, using correlations from existing fuel performance codes or material property libraries. Although these models perform adequately under many conditions, they have well-understood limitations. One obvious weakness is that the models are only valid in the domain of the associated experimental data, and extrapolating beyond this domain can result in substantial errors. A further weakness is that these models are often correlated in terms of imprecise quantities, such as burnup. Materials can, for example, reach a given burnup under very different circumstances, leading to very different material behavior. To overcome these weaknesses, efforts are underway with BISON to develop models based on a multi-scale approach, where material behavior is described with a series of interdependent microstructure based models. This paper provides an overview of the material models available in BISON. To keep the scope manageable, only LWR fuel will be considered. The first section focuses on UO{sub 2} fuel, identifying the models generally required and briefly describing those that have been implemented in the code. The second section repeats this approach for Zircaloy cladding. A final section provides examples of areas where multi-scale material modeling has been applied to BISON. BISON is a modern 2D and 3D nuclear fuel performance code that is applicable to multiple fuel forms. A wide variety of complex and nonlinear material models are required to simulate fuel behavior. Focusing on UO{sub 2} fuel and Zircaloy cladding (LWR fuel), the set of empirical models currently implemented in BISON have been reviewed. To overcome some of the deficiencies associated with empirical models, multi-scale material models are being developed. Two multi-scale models that have been used in BISON were reviewed. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2023

High-fidelity multiphysics load following and accidental transient modeling of microreactors using NEAMS tools: Application of NEAMS codes to perform multiphysics modeling analyses of micro-reactor concepts

The feasibility of modeling microreactors using high-fidelity models with the Nuclear Energy Advanced Modeling and Simulation (NEAMS) tools is investigated in this report. Three overarching questions guided this research: can NEAMS tools readily be applied for high-fidelity multiphysics modeling of different types of transients in microreactor designs; how accurate are the results obtained; and are improvements needed in accuracy or user experience of NEAMS tools, especially considering newly developed capabilities? This work builds upon FY-2022 work, and two microreactor concepts considering heat pipe (HP-MR) and gas-cooled (GC-MR) technologies were further analyzed using high-fidelity multiphysics simulations. The NEAMS tools considered and coupled within the MultiApp environment are Griffin for neutronics, BISON for thermo-mechanics, Sockeye for heat pipe modeling (in HP-MR), SAM for 1D Fluid – 3D solid modeling of coolant channels and system modeling of balance of plant components (in GC-MR), and the SWIFT code for hydrogen redistribution in hydride moderator. The Heat Pipe MicroReactor (HP-MR) concept was further analyzed in FY-2023 to demonstrate the stochastic TRISO failure modeling capability in BISON to check operational limits of the TRISO fuel. A new full-core Gas-Cooled MicroReactor (GC-MR) model was developed based on the initial assembly-model used in Y-2022 and used for steady-state and accidental depressurization transient simulations. Accuracy of the simulations performed was assessed through 1) verification analyses completed on the different physics with code-to-code comparison, and 2) validation of the multiphysics simulations based on modeling of the Kilopower Reactor Using Stirling Technology (KRUSTY) experiment. In FY-2023, the mesh and model of KRUSTY was updated to closely match publicly available data, and the neutronic model was verified and validated against experimental control rod worth measurements. The multiphysics model of KRUSTY was developed and used for steady-state analysis and for modeling reactivity insertion transient. The calculated power increase and stabilization agrees well with experimental data following adjustment in fuel thermal expansion coefficient. As an important component of this project, the ANL team gathered experience with a wide range of NEAMS tools: the MOOSE Mesh System, Griffin, BISON, SWIFT, Sockeye, SAM, Workbench, and the MOOSE MultiApp System, and provided assessment of new capabilities. Noteworthy are the user assessment of the “vapor-only” flow model in Sockeye and development of a multiphysics startup transient in HP-MR unit cell for use as tutorial in Sockeye. The full-core GC-MR model was used for assessment of SAM for balance of plant modeling and for demonstrating the SWIFT code capability for hydrogen redistribution modeling in multiphysics transient analyses. In this process, several bugs/issues were identified and reported to developers. Finally, the assembly GC-MR model developed in FY-2022 coupling Griffin, BISON and SAM through flow blockage and rod ejection transients was published to the National Reactor Innovation Center (NRIC) Virtual Test Bed (VTB). The Heat Pipe MicroReactor (HP-MR) concept high-fidelity multiphysics coupling of Griffin/BISON/Sockeye in load-following and heat pipe failure transients was also published on the VTB. Those submissions are enabling thorough review of these models as well as wide distribution to industry, regulator, and university users. In this analysis, several new research questions were uncovered, and follow-up analyses are recommended to further improve some models, consider additional transients, and continue development of VTB models.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

ATF Transient Prescription Tests in TREAT

EXECUTIVE SUMMARY The Transient Reactor Test facility (TREAT) was constructed in the late 1950’s and provided thousands of transient irradiations before being placed in standby mode in 1994. In 2017, it resumed operation in order to reclaim its crucial role in nuclear-heated safety research. TREAT is a transient reactor where inherently-safe core physics, a nimble transient rod drive system, and a philosophy of continual facility improvement work together to enable flexible power maneuvers for fuel safety research. While the facility’s flexible design and multi-mission nature saw historic experiments for numerous transient types, TREAT was not specifically adapted to optimize or demonstrate its full performance capability for Light Water Reactor (LWR) postulated accidents including Hot Zero Power Reactivity Initiated Accidents (HZP-RIA) and Loss of Coolant Accidents (LOCA). The physics of HZP-RIA drive very brief pulses with fairly large energy excursions. TREAT is currently capable of approaching HZP-RIA pulse widths through transients initiated on a large step insertion of reactivity and terminated shortly thereafter by rapid re-insertion of the transient rods (“clipping”). LOCAs are postulated to occur quite differently, starting from normal power operations followed by a long period of lower power decay heat. TREAT is able to simulate LOCA events with an initial flattop transient for tens of seconds, followed by a sizeable power reduction, to a longer segment where specimen fission heating represents internal decay heat generation. Two reactor physics testing campaigns were undertaken to address the HZP-RIA type transients, neither of which had been performed in any great measure historically to address data gaps in the timing, repeatability of clipping and effect of dynamic transient rod motions on core-to-specimen fission power coupling. The first campaign, referred to as the Accident Tolerant Fuel (ATF) transient prescription tests, employed dosimeters in the core to characterize the effect of rod motions on specimen power coupling for clipped, unclipped, and low-level steady state operations each with approximately the same moderate core energy release. This campaign was accompanied by efforts to measure dosimetry and compare results to analytic predictions. The other HZP-RIA campaign, referred to as Narrow Pulse Width (NPW) transient prescriptions, employed higher step insertions with increased core energy release to demonstrate TREAT’s capability in achieving the narrowest-possible pulse width. Additionally, a series of LOCA transient prescriptions were also performed to demonstrate how well TREAT could perform long-duration shaped transients. Both NPW and LOCA transient prescriptions did not include dosimeters but used existing plant power instrumentation to produce transient power data. All three transient prescription campaigns were successfully carried out in TREAT. The ATF-3 transient prescriptions showed that TREAT can perform clipped transients with remarkable repeatability. Measurement and modeling both showed that the power coupling in small, fissile dosimeters does not change by more than 8% between clipped, unclipped, and steady state runs while uncertainty sources in both the modeling and measurement methods were identified and found to be on the order of 20% in total. Overall, the ATF-3 transient prescription tests were successful in determining repeatability and consistency of dosimeter power coupling, developing processes for modeling and measuring this crucial parameter, and demonstrating the need for continued efforts in transient prescription work at TREAT in support of future experiment design and post test data interpretation. The NPW and LOCA transient prescriptions were also successfully performed to exhibit TREAT’s capability for these important LWR accident categories. The NPW transients showed pulse widths of ~90 ms, which were the narrowest ever achieved in TREAT to date and relevant for the upper end range of HZP-RIA simulation. This effort provided crucial data for calibrating kinetic models to be used in enhanced pulse-narrowing strategies in upcoming irradiations. Similarly, The LOCA transient prescriptions showed that TREAT can simulate the needed nuclear heating and event time scales for LOCA fuel safety research and provided essential data for future experiment design.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Initial Assessment of CTF for Time-at-Temperature Applications

The US nuclear industry is interested in improving the economics of their fleet of light-water reactors (LWRs) by uprating US plants. One option being considered is to regain lost margin from overly conservative fuel safety limits. The current limit requires avoidance of critical heat flux (CHF) and prevents further operation of fuel that experiences a dry-out in boiling water reactors (BWRs) or departure from nucleate boiling (DNB) in pressurized water reactors (PWRs); however, it has been shown that temporary, mild dry-out of the fuel does not necessarily increase the risk of fuel failure during its normal anticipated operating life. Such mild dry-out or DNB events may occur during a plant anticipated operational occurrence (AOO), such as a locked rotor in a PWR or a pump trip in a BWR. The time-at-temperature (TAT) approach to regulating fuel operation aims to demonstrate that the fuel rod’s integrity is not challenged during such a mild transient that leads to CHF in which the fuel operates at an elevated temperature for a brief period of time. However, implementing this approach will require extensive fuel material experimental data, as well as supporting modeling and simulation (M&S) predictions, to ensure that the predicted fuel response during AOOs, with all applicable uncertainty considered, will not threaten the safety of the fuel during the transient or the remainder of its anticipated lifecycle. To address this need, a comprehensive effort is being proposed that includes generating cladding material data under TAT conditions, assessment of available code capabilities for TAT conditions, development of new mechanistic models, and demonstration of the M&S capabilities for AOOs of interest. This will require a joint effort between the Nuclear Energy Advanced Modeling and Simulation (NEAMS) and Advanced Fuels Campaign (AFC) programs, as well as close collaboration with nuclear industry stakeholders. The outcome of this collaboration will result in development and assessment of capabilities that can be used by the nuclear industry to support qualification of a TAT-based fuel failure criteria safety limit. This report focuses on the thermal hydraulics (T/H) modeling capabilities and summarizes currently available data for validating the T/H subchannel code CTF for TAT conditions, as well as preliminary assessment results of the code. The initial assessment also resulted in implementation of an alternative post-CHF heat transfer package, which has been shown to significantly improve accuracy. This report is not a final assessment and does not consider all available validation data; it is intended that a future assessment will more fully validate the code for this application.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Development of pebble-based extruded carbon rods for extreme plasma heat flux environments

This work presents first experiments toward the development of continuously renewable (extrudable) pebble-based carbon rods for use as plasma-facing components in extreme steady-state plasma flux environments. The primary envisioned application of this work is a first wall that can survive long-term in future magnetic fusion power reactors while also improving recovery of the reactor fuel (tritium and deuterium atoms). Bench tests applying extreme steady-state front-surface heat loads of up to 50 MW/m 2 are presented. Continuous pebble rod front-surface recession and intact pebble recovery are successfully demonstrated, at a rate of order 0.2 cm/s. Numerical simulations of the pebble rod front-surface recession are able to match observations reasonably well, indicating that the recession mechanism can be understood as occurring due to pebble thermal expansion and resulting shock and cracking of the inter-pebble binder. Tests of the pebble rod extrusion demonstrate that friction between the rods and the stainless-steel extrusion channel is tolerably low (<50 N for the expected channel length) over a wide range of temperatures. Front-surface outgassing rates below 1000 Torr L/s/m 2 are achieved, believed to be sufficiently low for use in magnetic fusion reactors. In conclusion, initial parametric scans over pebble rod size and binder fraction to vary front-surface recession rates are presented.

36 MATERIALS SCIENCE↗
DOE OSTI2017

Sensitivity Analysis of Lbloca In Response To 10 Cfr50.46c New Rulemaking

The Nuclear Regulatory Commission (NRC) is proposing a new rulemaking on emergency core system/loss-of-coolant accident (LOCA) performance analysis. In the proposed rulemaking, designated as 10 CFR 50.46c, the US NRC puts forward an equivalent cladding oxidation criterion as a function of cladding pre-transient hydrogen content. The proposed rulemaking imposes more restrictive and burnupdependent cladding embrittlement criteria; consequently more fuel rods need to be analyzed under LOCA conditions to maintain the safety margin, in contrast to the current practice with which only one hot rod needs to be analyzed. New multi-physics analysis methods are required to provide a thorough characterization of the reactor core in order to identify the locations of the limiting rods and quantify safety margins under LOCA conditions. The U.S. DOE’s Light Water Reactor Sustainability (LWRS) Program has initiated a project to develop multi-physics analytical capabilities, called LOTUS, to support the industry in the transition to the proposed rule. An approach to uncertainty quantification and sensitivity analysis with LOTUS was developed. A typical four-loop PWR plant model was developed and simulated by RELAP5-3D with inputs generated from core design and fuel performance analyses, and uncertainty quantification and sensitivity analysis were performed with seventeen uncertain input parameters. The maximum equivalent cladding reacted ratio (ECRR) and peak clad temperature ratio (PCTR) were selected as the figures of merit (FOM). Pearson, Spearman, partial correlation coefficients, and Sobol indices were considered for all of the figures of merit in the sensitivity analysis.

10CFR50.46c↗
DOE OSTI2016

Research in Support of TREAT Kinetics Calculations Using Rattlesnake/BISON Coupling Within MAMMOTH

This paper summarizes key aspects of research in evaluation of modeling needs for TREAT transient simulation. Using measured data from transient experiments performed circa 1959 in a small, test core used for startup testing, MAMMOTH multi-physics simulations are per-formed to simulate the core transient behavior. MAMMOTH uses internal libraries from Rattlesnake for time-dependent transport calculations, BISON for fuel performance calcula-tions and the MOOSE multi-physics framework to perform strongly coupled calculations to properly characterize the three-dimensional temperature-driven kinetics performance of the core during rapid power excursions. Early calculations were able to match the general shape of measured transient data but under-predicted peak power. Closer examination of the phys-ics of the core and uncertainty associated with rod positioning led to calculations in which the best estimate of the actual measured reactor period was modeled. This correction resulted in outstanding agreement between measurements and coupled multi-physics simulation results.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

Selection and Ranking of Experiments from the Halden Database in support of Multiscale Model Validation

Validating fuel performance codes, such as BISON, requires an extensive amount of experiments covering a wide range of operating conditions and fuel types. Within the light-water reactor (LWR) space, there have been several international experimental programs that have contributed to the wealth of available experimental data available for use. One of those international programs, the Halden Reactor Project (HRP), began in 1958 and utilized the Halden Boiling Water Reactor to conduct many highly instrumented experiments until the reactor closed in 2018. Idaho National Laboratory, through the U.S. Department of Energy, has utilized several Halden experiments to perform the initial validation of the BISON code based upon their inclusion in international modeling and simulation benchmarks. Recently, the HRP has provided member organizations a complete copy of all available data, reports, and presentations since the HRP began. This report provides an initial exploration of the data available in the database for use in validating the multiscale models under development in the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program for LWR applications. Ranking tables that identify potential validation cases are provided for the high priority models of interest. It was found that some of the recommended high priority experiments correspond to additional rods in existing assemblies already available in the BISON validation suite. It is expected that several of these cases will be incorporated into future NEAMS milestones in the fuels technical area for increased validation of BISON.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

MODELING A SPENT NUCLEAR FUEL CASK SEISMIC TEST

The US Department of Energy Spent Fuel and Waste Science and Technology (SFWST) program is planning to conduct a series of full-scale shake table tests to simulate hypothetical earthquake conditions and record the response of surrogate spent nuclear fuel (SNF) assemblies in a canister dry storage system mockup. The shake table motions will represent a range of hypothetical earthquake conditions at hypothetical locations in the continental US to generally define the range of mechanical loads that SNF can be expected to experience during extended dry storage periods. The earthquake conditions will represent seismic hazards in the 2,000-to-20,000-year return period range. The test will use instrumented pressurized water reactor fuel assemblies with surrogate mass inside the fuel rods instead of radioactive fuel pellets. Dummy assemblies with similar mass and dimensions to actual fuel assemblies will occupy the rest of the fuel assembly locations in a SNF canister that holds 32 fuel assemblies. The canister will be located inside a mockup vertical concrete cask. Instrumentation will record the motion of the major components of the complex dynamic system, and strain gauges will be used to record the cladding strain at select locations. Preparations for the test require modeling predictions to identify the range of response of the system and to help select specific earthquake cases to be simulated on the shake table from a large set of potential cases. This paper describes the pretest nonlinear finite element modeling efforts that have been completed to date, including cask system level modeling and fuel assembly modeling in LS-DYNA. The cask system level models are critical for anticipating sliding or tipping of an unanchored cask during the test. The fuel assembly model is needed to estimate the range of cladding strain response and fuel assembly structural response to be expected during the test and cladding strain measurements will be key metrics for model validation and the development of modeling best practices after the test is concluded.

Klymyshyn, Nicholas A.↗
DOE OSTI2016

Multi-Physics Simulation of The TRIGA Reactor

In nuclear engineering, the description of phenomena occurring in the core involves a coupling between neutronic and thermal hydraulics through the so called reactivity or thermal feedback. In water cooled reactor, thermal feedback and temperature coefficients of reactivity are important, when temperature changes, microscopic cross sections change as well impacting directly the neutron flux and so the thermal power. Other feedback, such as the variation of the macroscopic cross section due to a modification of the density can also have a significant impact on the core performance of the reactor. Those reasons were naturally leading in the development of coupling method in order to predict correctly the behaviour of the core in normal operation and accidental situation. This work describes the coupling scheme developed for the description of the TRIGA mark II reactor at JSI. The demonstration case presented here, the reactor is operating in steady state configuration at several power levels. The aim of our calculation was, for a given thermal power of the reactor, to numerically reproduce fuel temperature value monitored during reactor operation. The expected impact of the control rod on the neutron flux distribution was also verified. The predictions of the coupling scheme seem to be correct according to the few temperature measurements available and the expected neutron flux behaviour expecting from a control rod motion. We are considering precise measurement of the fuel temperature at more positions but also in the coolant channels between the fuel elements to complete the experimental validation phase. An upgrade of the coupling scheme into the transient mode is also considered to observe self-stabilisation of the power level, after a control rod perturbation, due to the negative temperature reactivity coefficient of the fuel. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

BISON Theory Manual The Equations behind Nuclear Fuel Analysis

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

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Application of the SHARP Toolkit to Sodium-Cooled Fast Reactor Challenge Problems

The Simulation-based High-efficiency Advanced Reactor Prototyping (SHARP) toolkit is under development by the Nuclear Energy Advanced Modeling and Simulation (NEAMS) Campaign of the U.S. Department of Energy, Office of Nuclear Energy. To better understand and exploit the benefits of advanced modeling simulations, the NEAMS Campaign initiated the “Sodium-Cooled Fast Reactor (SFR) Challenge Problems” task, which include the assessment of hot channel factors (HCFs) and the demonstration of zooming capability using the SHARP toolkit. If both challenge problems are resolved through advanced modeling and simulation using the SHARP toolkit, the economic competitiveness of a SFR can be significantly improved. The efforts in the first year of this project focused on the development of computational models, meshes, and coupling procedures for multi-physics calculations using the neutronics (PROTEUS) and thermal-hydraulic (Nek5000) components of the SHARP toolkit, as well as demonstration of the HCF calculation capability for the 100 MWe Advanced Fast Reactor (AFR-100) design. Testing the feasibility of the SHARP zooming capability is planned in FY 2018. The HCFs developed for the earlier SFRs (FFTF, CRBR, and EBR-II) were reviewed, and a subset of these were identified as potential candidates for reduction or elimination through high-fidelity simulations. A one-way offline coupling method was used to evaluate the HCFs where the neutronics solver PROTEUS computes the power profile based on an assumed temperature, and the computational fluid dynamics solver Nek5000 evaluates the peak temperatures using the neutronics power profile. If the initial temperature profile used in the neutronics calculation is reasonably accurate, the one-way offline method is valid because the neutronics power profile has weak dependence on small temperature variation. In order to get more precise results, the proper temperature profile for initial neutronics calculations was obtained from the STAR-CCM+ calculations. The HCFs of the peak temperatures at cladding outer surface, cladding inner wall surface, and fuel centerline induced by cladding manufacturing tolerance and uncertainties on the cladding, coolant, and fuel properties were evaluated for the AFR-100. Some assessment on the effect of wire wrap configuration and size of the bundle shows that it is practical to use the 7-pin bare rod bundle to calculate the HCFs. The resulting HCFs obtained from the high-fidelity SHARP calculations are generally smaller than those developed for the earlier SFRs because the most uncertainties involved in the modeling and simulations were disappeared. For completeness, additional investigations are planned in FY 2018, which will use random sampling techniques.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗