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All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Propagating neutronic uncertainties for FFTF LOFWOS Test #13

The safety evaluation conducted for licensing a Sodium-cooled Fast Reactor (SFR) may require modeling transients with best-estimate calculation tools that must first be validated against real-world measurements. To provide the community with a valuable benchmarking opportunity for validating SFR analysis tools and methods, the International Atomic Energy Agency (IAEA) initiated a coordinated research project (CRP) in 2018 for the analysis of the Fast Flux Test Facility (FFTF) Loss of Flow Without Scram (LOFWOS) Test #13. The impact of nuclear data uncertainties on neutronics parameters was previously investigated based on the COMMARA-2.0 covariance matrix. Since the transient simulation results are very sensitive to certain reactivity coefficients, it was decided to employ rigorous uncertainty propagation methods to quantify the impact of nuclear data uncertainties on the best-estimate predication of FFTF LOFWOS Test #13. The DAKOTA code is used to propagate neutronic uncertainties through SAS4A/SASSYS-1 transient simulations, while taking into account spatial and reaction-wise correlations within these uncertainties. This study shows that the remaining discrepancies observed between the Argonne National Laboratory (ANL) best-estimate results and the experimental measurements can be partly explained by the uncertainty associated with Gas Expansion Module (GEM) worth, which contributes the majority of the overall nuclear data uncertainty on the output from the FFTF LOFWOS Test #13 transient simulation. This study also confirmed the importance of including spatial and reaction-wise correlations of nuclear data uncertainties on feedback coefficients in the uncertainty propagation to avoid under-estimating their impact during the transient simulations.

42 ENGINEERING↗
DOE OSTI2020

Benchmark Modeling and Simulation of the FFTF LOFWOS Test #13 Using SAM

The Fast Flux Test Facility (FFTF) was a 400 MW thermal powered, oxide-fueled, liquid sodium cooled test reactor, built to assist development and testing of advanced fuels and materials for fast breeder reactors. In July 1986, a series of unprotected Loss of Flow Without Scram (LOFWOS) transients were performed in FFTF as part of the Passive Safety Testing (PST) program. The LOFWOS Test #13, which was initiated at 50% power and 100% flow with the pump pony motors left off, has been chosen as a benchmark case by IAEA to support collaborative efforts within international partnerships on the validation of simulation tools and models in the area of sodium fast reactor passive safety in an IAEA Coordinated Research Project (CRP), launched in October 2018. The System Analysis Module (SAM) is an advanced and modern system analysis tool under development at Argonne National Laboratory for advanced non-LWR safety analysis. 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. To participate the IAEA CRP and enhance the SAM validation base for advanced reactor transient safety analysis, benchmark simulations of the FFTF LOFWOS Test #13 are performed using the SAM code. In this first phase of the validation effort, the thermal-hydraulic behavior of the reactor system is the focus and the reactor kinetics is not considered in the SAM FFTF model. Instead, the results of Argonne’s neutronics calculations are directly used, including the power shape of the active core region and the power history during the transient. The simulation results of FFTF at steady state agreed well with the measured data from the test. During the transient, reasonably good agreement were also obtained. Future work to improve the model will focus on introducing the reactivity predictions into the model, as well as better understanding or resolving the current discrepancies with the measured data.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Measurement of Void Fraction Distribution in Steam-Water Two-Phase Flow in a 44 Bundle at 2 MPa

To contribute to the clarification of the Fukushima Daiichi Accident, JAEA is working on getting instantaneous void fraction distribution data in steam-water two-phase flow in rod bundle geometry under high pressure, high temperature condition, with using Wire Mesh Sensor (WMS) developed at JAEA for high pressure, high temperature condition (Up to 2.8 MPa, 230 deg. C), focusing on the low flow rate condition after the reactor scram. The data is expected to be used in the validation of the detailed two-phase flow codes TPFIT and ACE3D developed at JAEA. The time and space averaged void fraction data is also expected being used in the validation of the drift-flux models implemented in the two fluids codes, such as TRACE code. We are still in the middle of the data analyses and here we only report a part of the results. For the present bundle rod geometry, the void fraction shows easily gathered at the center area surrounded by 4 rods and the void fraction in the sub - channels between two adjacent rods are low. Under the same steam vapor flow rate condition, void faction increase with the decrease of the water flow rate. Under the same water flow rate, the void fraction increases with the increase of the steam vapor flow rate. In recent future, we will perform more experiments under further low water flow rate condition, including stagnation condition, to confirm this changing tendency. In the present experimental range, no effect of pressure on void fraction was observed. A 3-layer WMS, with a vertical distance between the adjacent layers of 1.3 mm, was used to measure the local instantaneous void fraction in this research. By using the correlation coefficient method, we can determine the time delay between the bubble detections from the upper and the lower layer sensors. Then the speed of the bubble, and its length, can be calculated. We are working on the establishment of the databank for the slug-churn bubbles information. (authors)

42 ENGINEERING↗
DOE OSTI2016

Updating a PWR Simulator in Python

A nuclear reactor is a complicated system that is comprised of many systems and components. In order to fully model a reactor, several approximations and correlations must be used for a real-time simulation. In 1984 as part of a doctoral dissertation at Massachusetts Institute of Technology (MIT), a FORTRAN 77 program was developed to simulate a four-loop, Westinghouse pressurized water reactor (PWR). The program was later called PRISM and extended to include a comprehensive set of features accessible in a DOS-based user interface. This present work seeks to update the original PRISM program for cross platform deployment using the Python language. The goal of this present work is to update and improve a teaching tool for both students and reactor operators. Originally, PRISM was created as part of a PhD project in FORTRAN 77, but the code no longer works readily on modern operating systems. Python was chosen to rewrite the program because of the ease of use of the Python modules including Kivy as well as the portability of the program to various operating systems. This project aims to develop desktop and mobile versions of the software, allowing the user flexibility in choice of platform. This simulator is scheduled to be deployed in time for a live demonstration at the 2016 Annual ANS meeting. The project aims to produce a reactor simulator that will showcase the dynamics of a PWR system in real-time under various conditions. Some potential scenarios include startup conditions, scram operations, or failure conditions such as a break in a coolant leg. The user will be able to induce various failure scenarios and the program will simulate how the reactor will behave. The simulation will include a suite of monitoring functions that will allow automatic systems to attempt to adapt to the changing conditions, e.g., control rod movement. This program will provide an educational resource for anyone learning about nuclear reactor kinetics or dynamics of a reactor system. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Sensitivity analysis of xenon reactivity temperature dependency for HTTR LOFC test by using Relap5-3D code - 18365

A high-temperature gas-cooled reactor (HTGR) has superior safety characteristics. A loss of forced cooling (LOFC) test using a high-temperature engineering test reactor (HTTR) was carried out to verify the inherent safety characteristic of HTGR when forced cooling is diminished without reactor scram. In the test, all-gas circulators were tripped with an initial reactor power of 9 MW. The reactor power decreased from 9 MW to 0 MW by negative reactivity feedback in the core. Re-criticality was shown after about 8 h. This study focused on developing a point kinetics method with RELAP5-3D code for an LOFC accident. For future HTGR design and safety analysis, point kinetics method is important because of simple method. One key element of the point kinetics method is the reactivity performance estimation. There is a large temperature difference between the inlet and outlet of the core in HTGR, and the temperature fluctuation range has been large in several accidents. We analyze the temperature dependency of xenon-135 reactivity and show that the temperature dependency of xenon-135 microscopic absorption cross-section affected the re-criticality time of the LOFC test. This finding will be useful for predicting the emerging time and the peak level of reactor power increase after becoming re-critical. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Design and implementation of safety channels for the Idaho State University AGN-201m research reactor control console - 216

The purpose of the work described in this paper is to design, implement, and install new safety channels for the control console of the Idaho State University (ISU) AGN-201m Research Reactor. These channels will replace the current vacuum-tube driven system, which has been in operation for over 50 years at the ISU nuclear engineering laboratory. Due to the aging issues associated with this technology an entirely new control console is needed. There are 3 safety channels which serve as the monitors and safety systems for adequately shutting down the reactor should any limits be broken. These 3 safety channels work harmoniously with other systems within the new control console to ensure appropriate operation of the reactor. The safety channel replacements are composed of three complementary signal processing systems whose circuits monitor the operating parameters of the reactor. The key reactor information is shown on several independent meters corresponding to each of the three safety channels. Finally, the reactor trip points are set to correspond with the standards promulgated by the Nuclear Regulatory Commission (NRC) to ensure safe operation of the ISU AGN-201m Research Reactor. The interconnections between each of these systems ensure that if any Channel fails or otherwise breaks a limit the entire system will initiate a scram and the reactor will become adequately sub-critical. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Interdiffusion Behavior of U{sub 3}Si{sub 2} and FeCrAl via Diffusion Couple Studies

Uranium Silicide (U{sub 3}Si{sub 2}) is in use as a dispersion plate fuel in test reactors, but has not been fully tested in light water reactor (LWR) conditions. U{sub 3}Si{sub 2} pellets have been fabricated for use in commercial LWRs as part of an accident tolerant fuel system U{sub 3}Si{sub 2} has a higher uranium loading density as well as a higher thermal conductivity relative to UO{sub 2}. FeCrAl alloy is a candidate cladding material to replace the industry standard, Zircaloy. FeCrAl has outstanding high-temperature mechanical strength and oxidation resistance under accident conditions. The U{sub 3}Si{sub 2}-FeCrAl system is expected to provide more than double the amount of time from a Scram until failure criteria for the reactor is met, as well as producing roughly half the amount of H{sub 2} in the event of a loss of coolant accident (LOCA). Uranium and iron have well know eutectics that may affect/enhance diffusion of the U{sub 3}Si{sub 2}/FeCrAl fuel system. The interaction of iron with molten uranium has been investigated In the U-Fe system, UFe{sub 2}, which melts congruently at 1235 deg. C, forms a eutectic with iron that melts at 1080 deg. C and has a composition of U-77at% Fe. UFe{sub 2} also forms a eutectic with uranium that melts at 725 deg. C and has a composition of U-34at% Fe. The U-34at% Fe eutectic extended the temperature range down to 740 deg. C. Regular operating temperatures for U{sub 3}Si{sub 2} fuel is expected to be around 300 - 400 deg. C. The eutectics of iron and uranium would nominally only be of concern in accident scenarios.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Preliminary Analysis of FFTF Loss-of-Flow Tests with SAS4A/SASSYS-1

The Fast Flux Test Facility (FFTF) was the most recent Sodium-cooled Fast Reactor (SFR) built in the United States. The Passive Safety Test (PST) program was carried out on FFTF between 1984 and 1986 to demonstrate the capability of passive reactor shutdown and decay heat removal in response to protected and unprotected transients. In order to recover the information and knowledge that has been gained from FFTF, a research project was initiated focusing on the simulation of Unprotected Loss of Flow Without Scram (LOFWOS) tests similar to the knowledge preservation activities conducted for EBR-II by Argonne National Laboratory in earlier years. These knowledge preservation activities are intended to support the validation of computational tools and methods in the area of SFR passive safety. Comparisons with experimental data and other safety codes provide opportunities to improve SFR design and safety analysis codes and methods.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Quantification of functional impact classification on the current U.S. nuclear fleet

Currently, nuclear power generates only a small fraction of the world's electricity, though it does have the potential to meet the needs of the world as long as its production can compete with the alternative energy base load producers. In U.S the commercial nuclear fleet comprises of Light Water Reactors (LWRs), consisting of Pressurized Water Reactors (PWRs) and Boiling Water Reactors (BWRs). Many plants experience multiple events each year, which lead to reactor trips or scrams. This subsequently leads to plants being off the grid, which is not an economically favorable situation in the current energy markets. This study helps identify functional impact (FI) event categories for both PWR's and BWR's. Understanding that not all events and their causes can be anticipated, current industry best practices implement lessons learned to provide challenges and issues identification. In turn these are studied to achieve improved operational practices, and design modifications. This leads to overall enhanced resiliency. FI groups are the categories of initiating events (IE). IEs are unplanned events that occur while a nuclear plant is in critical operation, and requires that the plant be shut down to achieve a stable state. When a risk cannot be eliminated, IEs are identified and studied to reduce the risk and contribute to the identification of mitigations, safer designs and operational practices. Breaking IE events into groups' results in the classification of FI groups. A FI group is a risk-significant event category that could impact the ability of a NPP to remove decay heat from the reactor. Inability to remove decay heat from the reactor has the potential to lead to core damage. For an event to be considered a FI, it must be associated with a manual or automatic reactor trip, irrespective of the order of events. In addition, the plant must be at or above the point of adding heat and the event must happen before or shortly after a reactor trip. One or more FIs may be identified in any individual reactor trip sequence. The relationship between FI sequence, and/or frequency per a plant event is outside the scope of this paper. For example, the loss of offsite power (LOOP) can occur from eternal events (i.e. weather, human, wildlife etc.). This results in an event that requires a transfer of power to an emergency source, such as a generator, which produces a reactor trip and then a closure of main steam isolation valves (MSIVS). The FI applicable to this reactor trip sequence would be LOOP and closure of at least one MSIV in each main steam line. The results presented herein are focused on the analysis and evaluation of FI event categories. Only FI categories that can be found in IE studies, Initialing Events Spreadsheet and are recorded in a historical review of Licensee Event Reports (LERs) from 1988 through 2013. The results of this analysis can support plant decisions, design, procedures, and identify weakness.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

A simulation of I 2 S-LWR selected transients

The Integral Inherently Safe (I 2 S-LWR) reactor was designed to improve the safety performance of PWR types reactors, and several design basis accidents were eliminated using innovative approaches to the I 2 S-LWR plant layout. However, some accidents such as station blackout (SBO), main steam line break (MSLB) type accident – the secondary hot leg break (SHLB) – were deemed statistically relevant, and the work performed here was to simulate these events using the system thermal–hydraulics code RELAP5 coupled to the 3D spatial kinetics code PARCS. The safety response of the reactor to a station blackout is predicted to be acceptable at both beginning of cycle (BOC) and end of cycle EOC core conditions with either the UO 2 or U 3 Si 2 fuel options considered here. In the analysis of the SHLB performed here, an unprotected (no SCRAM) condition was assumed to highlight the I 2 S-LWR capabilities. Results indicate that with the UO 2 core, the fuel temperatures rise notably, but no fuel melting was predicted and departure from nucleate boiling ratio (DNBR) limits were not exceeded. Similarly, the U 3 Si 2 fuel performed well, but with lower fuel temperatures and decreased DNBR. Following the break of the steam line in the I 2 S-LWR, shutdown is achieved after about two minutes via negative reactivity from loss of primary flow. In all cases considered, the I 2 S-LWR appears to accomplish the goals of increased safety performance. Further design improvements and other fuel options may marginally improve on the observed safety performance. Additionally, higher fidelity analysis with the next generation subchannel, Computational Fluid Dynamics (CFD), or full core transport may be useful in refining the prediction of the peak fuel and cladding temperatures, as well limiting DNBR during the accidents.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2022

Assessment of SCALE and MELCOR for a generic pebble bed fluoride high-temperature reactor

This paper presents the development and application of SCALE and MELCOR models for a fluoride salt-cooled high-temperature reactor (FHR) based on publicly available specifications. SCALE version 6.3beta15 was used to generate power distributions and decay heat curves, and MELCOR version 2.2.18019 was used to calculate the thermal hydraulic response of an assumed FHR primary system. Here, an approach was developed for determining the equilibrium state of the core using depletion of a core slice model and blending of fuel compositions at different burnups to provide three-dimensional fuel composition in the core. Results compared between a core with entirely fresh fuel and one with an equilibrium fuel composition revealed that the equilibrium core led to lower steady-state temperatures but slower cooldown during a loss of flow accident (LOFA). A sensitivity study was then conducted to explore the transient response of our FHR system to variations in thermal hydraulic parameters of the system using the equilibrium core. The inlet temperature, graphite thermal conductivity, and SCRAM time all provided significant control over peak fuel and coolant temperatures. Uncertainties in radionuclide decay data in SCALE were used to perform a decay heat sensitivity study, and we found that uncertainties in decay heat led to negligible impact on peak temperatures during the course of the transient. In all cases evaluated, the observed peak fuel temperatures remained approximately 700 K below anticipated failure limits for the LOFA.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

Benchmarking FFTF LOFWOS Test# 13 using SAM code: Baseline model development and uncertainty quantification

The development and deployment of advanced reactors, such as the sodium-cooled fast reactor (SFR), relies on sophisticated modeling tools to ensure the safety of the design under various transients. The predictive capability of these advanced modeling tools requires validation to garner trust in supporting the licensing of the advanced reactors. For this reason, the International Atomic Energy Agency (IAEA) initiated a coordinated research project (CRP) in 2018 for the analysis of the Fast Flux Test Facility (FFTF) Loss of Flow Without Scram (LOFWOS) Test #13.In this study, we present and discuss the benchmarking efforts of the modern system code SAM on the FFTF LOFWOS Test #13. Further, the SAM baseline model was developed according to the benchmark specification, which included a detailed core model with reactivity feedback. Generally, good agreement was observed between the baseline results and benchmark measurements; however, discrepancies persisted, particularly in predicted fuel assembly coolant outlet temperatures. Utilizing the baseline model, uncertainty quantification (UQ) and sensitivity analysis (SA) were conducted with the assistance of various statistical learning and machine learning methods, including kernel density estimation, Gaussian processes, and Sobol indices. Following the baseline model prediction and UQ and SA results, we discuss the reasons for the simulation discrepancies and propose further improvements to the model. This benchmarking effort adheres to the best-estimate plus uncertainty approach and can serve as a valuable example for supporting risk-informed licensing of advanced reactors.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Impact of U-10Mo HALEU fuel element tolerances on the Massachusetts Institute of Technology reactor safety and operational performance – Thermal hydraulics

The U.S. is coordinating efforts for the conversion of six U.S. High Performance Research Reactors (USHPRR) including one critical facility from highly enriched uranium (HEU) to low-enriched uranium (LEU). In order to continue the mission of these reactors, including the Massachusetts of Institute of Technology Reactor (MITR), and achieve similar performance, high-assay low-enriched uranium (HALEU) with a high-density metallic alloy of uranium with 10 wt% molybdenum (U-10Mo) is being evaluated. The impact of the fabrication specification and tolerances was assessed following the preliminary design of the MITR LEU fuel elements using the U-10Mo monolithic alloy. This research focuses on the analysis of fabrication specification impact on thermal hydraulics (TH) characteristic of the MITR LEU core as a function of the variation of the relevant fuel specification parameters (e.g., coolant channel gap thickness, fuel plate thickness, etc.). The analyses are performed based on an all-fresh LEU fuel conversion plan identified in a preliminary safety analysis report submitted to the Nuclear Regulatory Commission. The reactor power margin to the onset of nucleate boiling (ONB) is assessed under the limiting safety system settings (LSSS), where a scram occurs, to ensure there is sufficient margin to the reactor safety limit, which is defined by the onset of flow instability that occurs after the ONB. The best estimate plus uncertainty approach is employed to analyze this TH characteristic, which yields realistic results while maintaining adequate conservatism, utilizing a statistical uncertainty propagation method with the STAT7 code. The TH characteristic is analyzed as a function of the variability of the specification parameters resulting from the fabrication process. The main findings of this study show that the MITR core can meet the TH safety and operational requirements at the all-LEU initial core startup (cycle 1), selected transition cycles (most reactive cycle and most limiting cycle: cycle 3 and 5, respectively) and equilibrium (cycle 14) cores under all limiting fabrication parameter combinations considered. In addition, the analyses show that the dependency of the core power margin to ONB on those specification parameters that have the most direct impact on TH performance is non-linear but monotonically decreasing within the specification tolerances. The third order polynomial fit curves are reported in detail for selected limiting cases and can serve as a powerful tool for future MITR fuel management in cases such as when HALEU supply is established that may allow additional cycle length or other operational benefits.

Conversion↗
DOE OSTI2020

Burst behavior of nuclear grade FeCrAl and Zircaloy-2 fuel cladding under simulated cyclic dryout conditions

A novel experiment to simulate cyclic dryout in boiling water reactors has been developed to better understand the performance of nuclear grade FeCrAl cladding in a BWR during dryout conditions caused by an Anticipated Operational Occurrence or Anticipated Transient Without SCRAM - both of which are Design Basis Accidents. Internally pressurized C26 M FeCrAl alloy cladding and Zircaloy-2 cladding were subjected to rapid 300°-650 °C thermal cycling in a steam environment; actual maximum temperatures were found to vary between materials but were always above 650 °C. In the range of 32–55 MPa hoop stress, Zircaloy-2 cladding burst within 1–16 cycles (about 100 s of dryout duration above 600 °C), while at 76 MPa hoop stress, C26 M cladding remained virtually undeformed after completing 54 cycles (over 1000 s of dryout duration above 600 °C). Higher temperature 300°-700 °C and 300°C–800 °C cycling experiments had to be performed to induce C26 M burst – failure occurred after 20 cycles in the former and during the first cycle in the latter. Zircaloy-2 and C26 M failure criteria were used to generate hoop stress specific dryout lifetimes. Overall, the simulated cyclic dryout experiments show that nuclear grade C26 M cladding has significantly enhanced survivability under dryout conditions relative to Zircaloy-2.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

BISON analysis of FeCrAl and Zircaloy cladding deformation during simulated BWR cyclic dryout conditions

The BISON fuel performance analysis code has been used to study FeCrAl and Zircaloy cladding performance to better understand cladding deformation behavior in boiling water reactors during conditions caused by anticipated operational occurrences (AOOs) or by anticipated transients without scram (ATWS). BISON models of novel separate effects tests simulating ATWS events predicted mechanical responses to thermal cycling boundary conditions. Here, we assessed FeCrAl and Zircaloy deformation and burst models to determine those best suited for cyclic dryout event analysis. In addition, we compared diametral strain profiles of the post-experiment claddings to simulation results. The capability of the BISON code for modeling BWR AOO or ATWS events is promising.

36 MATERIALS SCIENCE↗
DOE OSTI2018

Safety analysis of the FASt TEst reactor (FASTER) preconceptual design

Here, the FASt TEst Reactor (FASTER) plant is a sodium-cooled fast spectrum test reactor design that was developed as part of an advanced demonstration and test reactor study, and provides high levels of fast and thermal neutron flux for scientific research and development. The 120 MWe FASTER plant has been designed with extended testing capabilities in mind, while keeping the design as simple as possible in order to make it efficient and cost attractive. The main function of the reactor is to provide neutrons for irradiation testing utilizing existing proven technologies to ensure a high technology readiness level. A preliminary safety analysis was performed of the FASTER design using the fast reactor safety analysis code SAS4A/SASSYS-1 to assess the reactor's safety performance during beyond design basis accidents. Three types of transient scenarios were analyzed for the FASTER design: unprotected transient overpower, unprotected loss of heat sink, and unprotected station blackout. These transients were simulated at beginning of equilibrium cycle and end of equilibrium cycle conditions. For each transient, failures of both the primary and secondary reactor scram systems were assumed. Because of the favorable reactivity feedbacks, relatively low peak fuel, cladding, and coolant temperatures were obtained during the transients and the reactor vessel temperature remained well below the limit for ensuring structural stability.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
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↗