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DOE OSTI2025

Temperature sensitivity of the equilibrium neutronics and accident analysis of the HTR-10

Pebble-Bed High-Temperature Gas-cooled Reactors (PB-HTGR) are moderated by the graphite in the fuel pebbles and the graphite reflector surrounding the pebble-bed. Because graphite is by far the most abundant material in PB-HTGRs and the primary moderator, accurate modeling of the graphite material, including density, impurities, and temperatures, is crucial for accurate computational modeling and simulation of these reactors. While main characteristics of the graphite components are often known, the local temperature is less well known and often averaged over all components. Here, this work studies the impact of considering accurate temperature profiles in the graphite material on the generation of a small PB-HTGR model at the state of equilibrium operation and on short-term accident progression. The fuel compositions for the PB-HTGR were determined using a jump-in equilibrium modeling method, the Axial Radial Zone Equilibrium Modeling (AR-ZEM) method. In contrast to previous work, the AR-ZEM method was used considering thermal-hydraulic feedback from the MELCOR code to determine temperatures of the fuel pebbles and the surrounding graphite reflector. The consideration of an axial and radial temperature profile in the core and reflector, as opposed to uniform material temperatures, had an impact of almost 1,300 pcm on the equilibrium core eigenvalue and caused significant differences in the discharged plutonium fuel inventory with up to 4.9% and 11.0% for Pu-239 and Pu-242, respectively. To assess the impact on short-term accident progression, two Anticipated Transient Without SCRAM (ATWS) events, a Pressurized Loss of Forced Coolant (PLOFC) and a Control Rod Withdrawal (CRW) with loss of flow, were simulated with MELCOR. The use of temperature profiles in the equilibrium core models did not reveal a significant impact on the temperature, power, or reactivity responses during the transients. In conclusion, a need for consideration of accurate temperature profiles, in particular for the graphite reflector, was found for the generation of equilibrium PB-HTGRs core models using jump-in methods, but detailed temperature profiles may not be necessary when performing conservative transient analysis.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2019

Investigation of high X-ray conversion efficiency Kr filled gas sources at the National Ignition Facility

In this work, we report on the performance of high x-ray fluence Kr K-shell sources that are being developed for high energy density experiments. These targets are 4.1 mm in diameter 4.4 mm tall hollow epoxy tubes having a 40 μm thick wall holding 1.5 atm of Kr gas. For these shots, the National Ignition Facility laser delivered a nominally constant total energy of ≈750 kJ of 351 nm (3ω) light at the three power levels [e.g., ≈120 (low), ≈145 (medium), and ≈210 TW (high)]. The Kr K-shell (E photon = 8–20 keV) x-ray radiant intensity and radiant energy (kJ/sr) of these sources were found to increase as a function of laser power but began to plateau at the highest laser power. The Kr K-shell radiant energy increased from ≈1 kJ/sr at ≈120 TW to ≈2 kJ/sr at ≈210 TW. Radiation hydrodynamics simulations predict radiant energies to be always higher than these measurements. The increase in K-shell emission is attributed to its strong dependence on the electron temperature. Electron temperature distributions were inferred from measured Heα and Lyα line emission through the use of a genetic algorithm and Scram modeling. The inferred temperatures from the experiment are 20% to 30% higher than those predicted from modeling.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2020

Characterization of an imploding cylindrical plasma for electron transport studies using x-ray emission spectroscopy

We report here on the characterization of the conditions of an imploding cylindrical plasma by time-resolved x-ray emission spectroscopy. Knowledge about this implosion platform can be applied to studies of particle transport for inertial confinement fusion schemes or to astrophysical plasmas. A cylindrical Cl-doped CH foam within a tube of solid CH was irradiated by 36 beams (I total ~5×10 14 W/cm 2 , 1.5ns square pulse, and E total ~ 16.2kJ) of the OMEGA-60 laser to radially compress the CH toward the axis. The analysis of the time-resolved spectra showed that the compression can be described by four distinct phases, each presenting different plasma conditions. First the ablation of the cylinder is dominant; second, the foam is heated and induces a significant jump in emission intensities; third, the temperature and density of the foam reaches a maximum; and finally, the plasma expands. Ranges for the plasma temperature were inferred with the atomic physics code SCRAM (Spectroscopic Collisional-Radiative Atomic Model) and the experimental data have been compared to hydrodynamic simulations performed with the 2D code FLASH, which showed a similar implosion dynamic over time.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2023

Ionization disequilibrium in K- and L-shell ions

Time-gated Sc K-shell and Ge L-shell spectra are presented from a range of characterized thermodynamic states spanning ion densities of 1019–1020cm−3 and plasma temperatures around 2000 eV. For the higher densities studied and temperatures from 1000 to 3000 eV, the Sc and Ge x-ray emission spectra are consistent with steady-state calculations from the modern atomic kinetics model SCRAM. At the lower ion densities achieved through plasma expansion, however, the model calculations require a higher plasma temperature to reproduce the observed Ge spectrum. We attribute this to ionization disequilibrium of the Sc because the ionization time scales exceed the hydrodynamic timescale when the inferred temperatures diverge.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2025

X-ray absorption spectroscopy measurements of radiatively ionized argon gas

X-ray absorption spectroscopy is a diagnostic tool that can characterize the temperature and ionization state of a plasma. This technique requires experiments to characterize the platform, careful data calibration, and comparison with atomic models to understand the plasma parameters. We performed ionizing radiation wave experiments at the OMEGA Laser Facility that used an ∼ 80 eV X-ray source to heat an Argon (Ar) gas cell at fill pressures of 3 atm. To diagnose the Ar plasma, we used a capsule backlighter offset 10 mm from the gas cell, to produce X-rays that were absorbed by the ionized Ar gas. The absorption analysis was calibrated using the significant line structure in the backlighter spectrum, which served as independent energy fiducials enabling definitive measurements of a 50-eV shift in the energy of the Ar K-edge due to ionization. We compare the measured absorption spectra to two independent atomic models, PrismSPECT and SCRAM, and show that the observed K-edge shift in the heated gas is consistent with ionization up to Ar 4+ and temperatures of 10 eV.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2022

Study of Stark broadening of krypton helium-$\beta$ lines and estimation of electron density and temperature in NIF compressed capsules

Here the National Ignition Facility (NIF) diagnostic instrument manipulator (DIM) - based high resolution (dHIRES) x-ray spectrometer was used to measure the time evolution of the electron density (n e ) and temperature (T e ) in the hot spot of four NIF compressed capsules with 25 ps time resolution during the 'stagnation' phase. The electron density was inferred by comparing the measured Stark broadening of the krypton (Kr) Heβ spectral complex with theoretical calculations that include ion dynamic effects, and the electron temperature was inferred by comparing the measured ratio of the intensity of a dielectronically excited Li-like Kr line to the intensity of the Kr Heβ resonance line with calculations using the spectroscopic collisional radiative atomic model (SCRAM) and CRETIN collisional-radiative models. The inferred, time averaged n e values mainly agree with n e values from neutron diagnostics within uncertainties, but the neutron time-of-flight values of T ion are consistently higher than dHIRES T e values by 200–700 eV. The dHIRES measurements and measurement techniques, method of uncertainty analysis, and discussion of comparisons with measurements from neutron diagnostics are presented.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2016

TRACE/PARCS Analysis of ATWS with Instability for a MELLLA+BWR/5

A TRACE/PARCS model has been developed to analyze anticipated transient without SCRAM (ATWS) events for a boiling water reactor (BWR) operating in the maximum extended load line limit analysis-plus (MELLLA+) expanded operating domain. The MELLLA+ domain expands allowable operation in the power/flow map of a BWR to low flow rates at high power conditions. Such operation exacerbates the likelihood of large amplitude power/flow oscillations during certain ATWS scenarios. The analysis shows that large amplitude power/flow oscillations, both core-wide and out-of-phase, arise following the establishment of natural circulation flow in the reactor pressure vessel (RPV) after the trip of the recirculation pumps and an increase in core inlet subcooling. The analysis also indicates a mechanism by which the fuel may experience heat-up that could result in localized fuel damage. TRACE predicts the heat-up to occur when the cladding surface temperature exceeds the minimum stable film boiling temperature after periodic cycles of dryout and rewet; and the fuel becomes “locked” into a film boiling regime. Further, the analysis demonstrates the effectiveness of the simulated manual operator actions to suppress the instability.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2017

Initiating Event Rates at U.S. Nuclear Power Plants: 1988–2016

Analyzing initiating event rates is important because it indicates performance among plants and also provides inputs to several U.S. Nuclear Regulatory Commission (NRC) risk-informed regulatory activities. This report presents an analysis of initiating event frequencies at U.S. commercial nuclear power plants from calendar year 1988 through 2016 as reported in licensee event reports. Engineers with nuclear power plant experience reviewed each event report since the last update to this report for the presence of valid scrams or reactor trips at power. To be included in the study, an event had to meet all of the following criteria: (1) the event includes an unplanned reactor trip (not a scheduled reactor trip on the daily operations schedule), (2) the sequence of events started when reactor is critical and at or above the point of adding heat, (3) the event occurred at a U.S. commercial nuclear power plant (excluding Fort St. Vrain and LaCrosse), and (4) the event was reported by a licensee event report. Sixteen initiating event groupings are trended and displayed. For some of the categories as relevant initiators are plotted separately for boiling water reactors and pressurized water reactors. P values are given for the possible presence of a trend over the most recent 10 years. Statistically significant decreasing trends were identified for BWR loss of heat sink and general transients.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2018

BOP-301 and BOP-302R: Test Definitions and Analyses

Following the conclusion of the International Atomic Energy Agency (IAEA) coordinated research project (CRP) on analysis of two EBR-II tests, Argonne has continued its EBR-II analysis with additional tests from the Shutdown Heat Removal Test (SHRT) series. The tests selected for this follow-up analysis are BOP-301 and BOP-302R, two unprotected loss of heat of sink tests in which the intermediate sodium pump tripped without scramming the control rods or tripping the primary pumps. The IAEA CRP focused on the SHRT-17 and SHRT-45R tests, the most severe protected and unprotected loss of flow tests performed during the SHRT series. With the primary pumps tripping, flow rates decreased to levels where flowmeter uncertainty was very high. Accurately capturing the progressions of the two tests despite this uncertainty was one of the key modeling and simulation challenges during the CRP. BOP-301 and BOP-302R were selected because the primary sodium pumps did not trip for those tests. Primary flow rate measurements are known with greater certainty. Power and temperature discrepancies can be explored without concern that flow predictions match test measurements but not the actual flow rates from the tests. These two tests were performed during the same test window as SHRT-45R and had the same core load configuration. Therefore, core models created for analysis of SHRT-45R were easily adapted for analysis of BOP-301 and BOP-302R. Another reason for analyzing BOP-301 and BOP-302R is to analyze tests in which the core inlet temperature changed significantly, which did not occur during the loss of flow tests. The transient progression of the unprotected SHRT-45R test was driven by a decreasing core flow rate while the core inlet temperature remained relatively constant. For the BOP tests, the transient progression was driven by an increasing core inlet temperature while the core flow rate was stable. This report defines the BOP-301 and BOP-302R tests and provides a selection of measured test data. Initial predictions from SAS4A/SASSYS-1 simulations of these tests are then presented. Following these simulations, additional analysis was performed with the Dakota uncertainty quantification and optimization toolkit to determine which parameters have the greater impact on the transient results and the optimal values of the most important parameters.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2019

Reactivity Feedback Modeling in SAM

System Analysis Module (SAM) is being developed at Argonne National Laboratory as a modern system-level modeling and simulation tool for advanced non-light water reactor safety analyses. It utilizes the object-oriented application framework MOOSE to leverage the modern software environment and advanced numerical methods. The capabilities of SAM are being extended to enable the transient modeling, analysis, and design of various advanced nuclear reactor systems. This report presents the development of Point-Kinetics modeling and reactivity feedback mechanisms in SAM. It also presents a new capability for predicting the thermal expansion in various structural components by coupling SAM with an external thermomechanics module (Tensor Mechanics) from the MOOSE framework. The analysis of the transient behavior of a nuclear reactor requires the coupled simulation of reactor kinetics and thermal-hydraulics of the reactor core, especially for those unprotected transients where the reactor scram system may not function properly. The point kinetics model has been widely used for reactor safety analysis due to its simplicity to capture the transient behavior of the reactor. Various reactivity feedback models have been developed and integrated with the Point-Kinetics module, including fuel axial expansion, core radial expansion, fuel Doppler, and coolant density reactivity. The reactivity feedback models in SAM are similar to the respective models used in SAS4A/SASSYS-1. This report first presents the brief theory of the Point-Kinetics module and reactivity feedback models. A number of verification tests have been performed where the code simulations are compared to the analytical model results. The reactivity feedback due to the thermal deformation, such as the fuel axial expansion and core radial expansion, is important for SFR transient analysis. Simplified thermal expansion models for the fuel pin and reactor constrain system (e.g. grid plate) are developed and verified in SAM. Additionally, a coupling interface is developed to couple SAM with external thermomechanical analysis modules for more accurate predictions of the thermal expansion of different components during the transients. The current coupling interface has been tested with the Tensor Mechanics module from MOOSE. These point kinetics and reactivity feedback modeling capabilities have also been demonstrated by simulating the early stage of the unprotected loss-of-flow (ULOF) accident in the Advanced Burner Test Reactor (ABTR). Both the stand-alone SAM and coupled SAM and Tensor Mechanics simulations are performed. It is confirmed that the major physics phenomena in the heat transport system of the ABTR reactor are captured by SAM, and the point kinetics model, reactivity feedback models, and the coupling schemes are working well as expected.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2020

Report Summarizing Boiling Transition (Dryout) Testing of FeCrAl Cladding

Nuclear fuel and cladding in a light water reactor must maintain integrity under all normal operational and anticipated operational occurrences. In a Boiling Water Reactor (BWR) some Anticipated Operational Occurrences (AOOs) and Anticipated Transients Without SCRAM (ATWS) events have been known to result in cladding damage due to dryout with conventional Zircaloy-2 claddings. This work presents an assessment of commercially-fabricated, accident tolerant iron-chromium-aluminum (FeCrAl) alloy C26M cladding in such events with regard to the ballooning failure mode for unirradiated cladding In the severe accident test station (SATS), an experiment has been developed to gauge accident tolerance of fuel cladding materials under simulated cyclic dryout conditions. Thermal cycling from 300°C to 650°C was implemented on C26M and Zircaloy-2 tubes. The C26M tube did not fail after 54 cycles. The Zircaloy- 2 tube run under the same parameters as the C26M tube failed during the fourth cycle, consistent with expectations from the literature. Due to thermal gradients present in the furnace, both tubes experienced a maximum temperature at ~780°C. Due to differences in tube geometry, the hoop stress of the C26M tube was significantly higher than that of the Zircaloy-2 tube, 76 MPa at the maximum peak temperature compared to 55 MPa at burst, respectively. The cyclic dryout experiments show an increased performance of C26M over Zircaloy-2 under simulated conditions, an indication of C26M possessing superior high temperature oxidation resistance, mechanical properties and potentially resistance to fatigue.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2020

Multiphase representation of hotspot formation and growth

Multiphase reacting flow approach has been adopted for modeling of hotspot formation and growth on sub-zonal scale. Mechanical response of the high explosive (HE) is modeled by the Visco-Plastic SCRAM model which includes plastic work, crack growth, and frictional heating along the cracks. In this work a sub-zonal regions with high temperature, hotspots, are modeled as a new phase. Formation and growth of the hotspot phase from the cold phase is modeled as a phase change. Each phase has its own temperature and pressure. Chemical reactions also progress in both phases. The growth of hotspots are determined by the solution of the heat conduction equation with appropriate heat sources due to friction and plastic heating. Sample calculations using the BABBO code show successful detonation calculations with hotspot formation and growth.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2022

SAS4A/SASSYS-1 Validation with the FFTF LOFWOS Tests #10-12

The FFTF liquid sodium cooled test reactor was built to assist the development and testing of advanced fuels and material for fast reactors. The goals of the FFTF Passive Safety Testing program included confirming the safety margins of FFTF as a liquid metal reactor, providing data for computer code validation, and demonstrating the inherent and passive safety benefits of its specific design features. The program included thirteen unprotected loss of flow without scram tests. The results of modeling and simulation of LOFWOS Tests #10, #11, and #12 are presented in this report to support validation of Argonne’s SAS4A/SASSYS-1 fast reactor safety analysis code. These three tests did not use the primary loop pony motors and thus transitioned to natural circulation flow rates from varying initial power levels. This validation activity leverages modeling and simulation efforts that originated under an International Atomic Energy Agency Coordinated Research Project (CRP) for benchmark analysis of the FFTF LOFWOS Test #13. The SAS4A/SASSYS-1 model originally developed for the CRP has been extended to include LOFWOS Tests #10-12 and modernized to utilize new code features and modeling practices. Predicted results from SAS4A/SASSYS-1 were compared to measured test data and differences were investigated. Overall, it was concluded that, for LOFWOS Tests #10-12, there was good agreement between the flow, power, and temperature predictions and the measured data.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2022

Benchmark Specification for FFTF LOFWOS Test #13

The Fast Flux Test Facility (FFTF) at the Hanford site in Washington was designed by the Westinghouse Electric Corporation for the U.S. Department of Energy. FFTF was a 400 MW thermal, oxide-fueled, liquid sodium cooled test reactor, built to assist development and testing of advanced fuels and materials for fast breeder reactors. After reaching criticality in 1980, FFTF operated until 1992, providing the U.S. Department of Energy (DOE) with the means to test fuels, materials, and other components in a fast neutron flux environment. In July 1986, a series of unprotected transients (with the plant protection system intentionally disabled) were performed in FFTF as part of the passive safety demonstration program. Among these were thirteen loss of flow without scram (LOFWOS) tests. The goals of this program included confirming the liquid metal reactor safety margins, providing data for computer code validation, and demonstrating the inherent and passive safety benefits of specific design features. The test defined in this benchmark is LOFWOS Test #13, which was initiated at 50% power and 100% flow with the pump pony motors turned off. This benchmark specification is intended to support collaborative efforts within international partnerships on the validation of simulation tools and models in the area of Sodium-cooled Fast Reactor (SFR) safety. Validated tools and models are needed to evaluate SFR inherent safety characteristics and assess the impact of passive design features in response to accident initiators. Comparisons with experimental data and the results of safety analyses from other groups create unique opportunities to improve predictive capabilities of computational codes and methods for SFR modeling and simulation. The conditions of the LOFWOS test along with the feedback from FFTF’s limited free bow core restraint system and the novel passive reactivity control Gas-Expansion Modules (GEMs) pose a very challenging and uniquely valuable benchmark exercise.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2022

SAS4A/SASSYS-1 Validation with EBR-II Tests Performed During the SHRT Testing Program

The EBR-II sodium-cooled fast reactor supported a wide range of fast reactor research activities. The final fifteen years of operations at EBR-II were used for experiments and tests to demonstrate the importance of passive safety in liquid metal reactors and the capability of such systems to provide strong passive responses during off-normal conditions. The Shutdown Heat Removal Test program at EBR-II provided test data supporting the validation of computer codes for design, licensing, and operation of LMRs, among other objectives. The program included nearly sixty tests, most of which were protected and unprotected loss of flow and loss of heat sink tests run at various initial powers and flow rates. This report documents the results of modeling and simulation of tests from the Shutdown Heat Removal Test program to support validation of Argonne’s SAS4A/SASSYS-1 fast reactor safety analysis code. SHRT-17 was a protected loss of flow test where a loss of electrical power to all sodium coolant pumps was simulated to demonstrate the effectiveness of natural circulation cooling characteristics. SHRT-45R was an unprotected loss of flow test where the control rod scram function of the plant protection system was disabled to demonstrate the effectiveness of EBR-II’s passive reactivity feedbacks. BOP-301 and BOP-302R were unprotected loss of heat sink tests that demonstrated how reactivity feedback effects driven by changes to the core inlet temperature can shut down the fission process. This validation activity leverages modeling and simulation efforts that originated under an International Atomic Energy Agency Coordinated Research Project for benchmark analysis of the EBR-II Shutdown Heat Removal Tests. The SAS4A/SASSYS-1 models originally developed for the CRP have been modernized to utilize new code features and modeling practices. Differences between the predicted and measured test data were explored and the causes of these differences were identified as being due to modeling approximations, uncertainty in transient component behavior, and instrumentation uncertainties. Overall, it was concluded that for SHRT-17, SHRT-45R, BOP-301, and BOP-302R there was good agreement between the flow, power, and temperature predictions and the measured data.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2023

SAS4A/SASSYS-1 Commercial Grade Dedication Example Report for a Generic Sodium Pool-Type Fast Reactor Application

In the U.S., a key component of the commercialization of advanced reactors is completion of a license application, which must ultimately be approved by the Nuclear Regulatory Commission (NRC). The approval of the license application by the NRC is contingent on satisfactory demonstration of the design basis and the response of the advanced reactor design to transient and accident scenarios using accepted codes and methods. This report describes the qualification and dedication requirements that the advanced reactor safety analysis system software SAS4A/SASSYS-1 are expected to need to fulfill to be used for sodium-cooled pool-type fast reactor licensing. The qualification and dedication requirements are identified through performance critical characteristics and evaluation model acceptance criteria representative of the advanced reactor design considered for licensing. This document captures, additionally, the verification process developed to demonstrate that the software fulfills the qualification and dedication requirements for a generic sodium-cooled pool-type fast reactor as part of the commercial grade dedication process. Like most software that has primarily existed in the research and development space, the most significant challenge facing SAS4A/SASSYS-1 for use in a licensing framework is the availability of a documentation basis describing the code pedigree. SAS4A/SASSYS-1 has been used for licensing of the fast flux test facility (FFTF) and the JOYO sodium-cooled fast reactor in Japan, as well as the design of the CRBR Plant. However, the historical verification and validation (V&V) activities supporting SAS4A/SASSYS-1 development do not align with modern software quality assurance (SQA) and V&V requirements. Two approaches to use of SAS4A/SASSYS-1 in a commercial licensing framework have been identified: commercial-grade dedication (CGD) and software qualification. The methods and requirements prescribed in the ASME NQA-1-2008/2009 Standard and Regulatory Guide 1.203 on the evaluation model development and assessment process (EMDAP) have been used as guidance to define the CGD and qualification processes, respectively. A qualification and dedication requirements matrix has been developed which utilizes fundamental software verification. In this process, software verification is defined as a software quality process aimed at defining software requirement specifications, developing software design documentation, and performing and documenting acceptance testing of the code against requirements. A key element of software qualification and dedication includes determination of software acceptance with respect to critical characteristics relevant to the functional requirements of the software. To assist with identification of cross-cutting transient phenomena and functional requirements, domestic SFR vendor designs have been reviewed to identify a reference SFR design. For this report, the reference design is defined as a pool-type reactor with metal alloy fuel, a liquid-metal intermediate heat transport system, and passive decay heat rejection systems. Given this reference, a series of high-level cross-cutting phenomena was identified for a general class of single-fault undercooling or reactivity insertion transients that scopes the design basis space, with the goal of assisting with prioritization of documentation development efforts for key transient models in SAS: 1) Reactivity feedback response prior to scram; 2) System-wide thermal inertia; 3) Transition in natural circulation flow regime in heat removal systems; 4) Decay heat generation; 5) Steady-state fuel characterization; 5) Clad/fuel behavior at elevated temperatures; 6) Point kinetics and decay heat; 7) Pump coastdown behavior; 8) Core flow redistribution in loss of forced convection; 9) Pool stratification. As a demonstration of CGD of SAS4A/SASSYS-1 for a sodium pool reactor, a software qualification and dedication gap analysis as it relates to code documentation has been performed. This effort leverages the framework established as part of the SAS4A/SASSYS-1 SQA Program. This CGD demonstration provides a framework that vendors can build upon to demonstrate the applicability of the SAS4A/SASSYS-1 software for licensing a sodium-cooled pool-type fast reactor.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2023

SAS4A/SASSYS-1 Validation with EBR-II Tests Performed During the SHRT Testing Program (Rev. 2)

The EBR-II sodium-cooled fast reactor supported a wide range of fast reactor research activities. The final fifteen years of operations at EBR-II were used for experiments and tests to demonstrate the importance of passive safety in liquid metal reactors and the capability of such systems to provide strong passive responses during off-normal conditions. The Shutdown Heat Removal Test program at EBR-II provided test data supporting the validation of computer codes for design, licensing, and operation of LMRs, among other objectives. The program included nearly sixty tests, most of which were protected and unprotected loss of flow and loss of heat sink tests run at various initial powers and flow rates. This report documents the results of modeling and simulation of tests from the Shutdown Heat Removal Test program to support validation of Argonne’s SAS4A/SASSYS-1 fast reactor safety analysis code. SHRT-17 and SHRT-6 were protected loss of flow tests where a loss of electrical power to all sodium coolant pumps was simulated to demonstrate the effectiveness of natural circulation cooling characteristics. SHRT-45R, SHRT-39, and SHRT-43R were unprotected loss of flow tests where the control rod scram function of the plant protection system was disabled to demonstrate the effectiveness of EBR-II’s passive reactivity feedbacks. BOP-301 and BOP-302R were unprotected loss of heat sink tests that demonstrated how reactivity feedback effects driven by changes to the core inlet temperature can shut down the fission process. PICT-5/6 demonstrated how core power could be increased and decreased via changes in the intermediate sodium and feedwater flow rates. This validation activity leverages modeling and simulation efforts that originated under an International Atomic Energy Agency Coordinated Research Project for benchmark analysis of the EBR-II Shutdown Heat Removal Tests. The SAS4A/SASSYS-1 models originally developed for the CRP have been modernized to utilize new code features and modeling practices. Differences between the predicted and measured test data were explored and the causes of these differences were identified as being due to modeling approximations, uncertainty in transient component behavior, and instrumentation uncertainties. Overall, it was concluded that for all tests simulated there was good agreement between the flow, power, and temperature predictions and the measured data.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2024

An Autonomous Critical Data Extrapolator for the AGN-201m

Nuclear nonproliferation serves as a key goal, being undertaken by the International Atomic Energy Agency (IAEA). To recognize proliferation there are two pathways that states, who intend to use nuclear material for malicious purposes can take, diversion can misuse. Diversion is when fissile nuclear material is declared to the IAEA for non-weapon purposes, but then covertly removed. If the source of nuclear material, that is not declared and not fissionable, is placed inside the reactor core to create fissile material used to create weapons then the state is using the second pathway of proliferation, misuse. With the emerging development in areas of simulation and machine learning the creation of virtual models of reactor systems, digital twins, serve as a potential method to identify proliferation through detecting anomalous behavior in the reactor. A digital twin for a physical nuclear reactor has never been developed, as digital twins serve as an emerging technology. To investigate the process for development and use of a digital twin for a nuclear reactor Idaho State University’s AGN-201m serves as the nuclear reactor used for development of this digital twin. A data acquisition system has been installed to the reactor system allowing for the transfer of collected data from a reactor operation to Idaho National Laboratory’s Deeplynx data warehouse. When utilizing data to train reactor physics and machine learning models, a significant challenge encountered is the initial state of the data. Nuclear proliferation will have the capacity to be detected when the reactor immediately starts up, nor will it occur after the reactor shuts down. Generally, it will be detected when the reactor is operating at some desired power over a sufficient period for that specific reactor design. For the AGN-201m this will be when the reactor is critical (generally 1 mW or above) for a timespan that is within or less than the range of a regular business day. Datasets sent to Deeplynx have had to be manually cut to when the reactor is critical based on plots of power levels. This method is inefficient and laborious, especially when using multiple datasets at once to train a model. To provide a more streamlined approach an automated critical data extrapolator is developed, with capabilities of recognizing when the reactor operation first reaches criticality, and when the reactor undergoes a SCRAM and is shutdown.

99 GENERAL AND MISCELLANEOUS↗