Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “fission power system”

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

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 379 records · Page 21
All sourcesNASA NTRSDOE OSTI
NASA NTRSDate not supplied

Lattice Confinement Fusion-Fast-Fission

Lattice Confinement Fusion (LCF) • Published 2020, Physical Review C 1,2 • Patented and Commercialized 2024, Astral Systems • LCF in compact neutron generator 50x increase in neutron flux, 99% from LCF • Produce medical radioisotopes • LCF Fast-Fission Hybrid Reaction • No Enriched Uranium • Demonstrated with US Navy and GEC • Modeling and Scaling under NIAC and NSF funding • Application • Deep space • Power: Icy worlds • Propulsion: Nuclear Electric Propulsion • Planetary Surface Power: Lunar and Mars • Terrestrial • DoD Operational Energy • Onsite power for Data Centers

nuclear electric propulsion↗
DOE OSTI2023

Evaluation of uncertainty in antineutrino spectra normalization calculations for advanced nuclear reactor monitoring

Antineutrino detection systems have been envisioned as an important aspect of safeguarding the next generation of nuclear reactors, especially considering designs utilizing exotic fuel cycles. Deployment of antineutrino detection systems for safeguarding applications is hindered by the uncertainties associated with the calculations required for antineutrino spectra measurements and predictions. The focus of this work is to assess the impact of system components on antineutrino spectra normalization uncertainties and their significance in reactor power monitoring sensitivity. The dominant limitation in antineutrino detection calculations is typically the uncertainty associated with a cosmogenic background. This limitation becomes more pronounced when signals are weak, although the issue is mitigated in larger reactors due to their stronger source strength. Additionally, antineutrino emission uncertainties vary with the isotopic composition of the reactor fuel. Unconventional fuel cycles, featuring less common fissioning isotopes, such as Pu-240, introduce larger antineutrino yield uncertainties. The findings from this study suggest that future research on safeguard-targeted antineutrino detection should prioritize background mitigation, particularly when background simulation is necessary. Advanced nuclear reactor designs have a major influence on background understanding and successful system implementation.

Physics↗
DOE OSTI2021

AGR 5/6/7 Data Qualification Report for ATR Cycles 162B through 168A

This report provides the qualification status of experimental data for the Advanced Gas Reactor (AGR) 5/6/7 fuel irradiation. AGR-5/6/7 was conducted in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) in support of development and qualification of tri-structural isotropic (TRISO) low-enriched fuel for use in high temperature gas-cooled reactors. The objectives of the AGR-5/6/7 experiments are to: (i) irradiate reference-design fuel particles to support fuel qualification, (ii) establish operating margins for the fuel beyond normal operating conditions, and (iii) provide irradiated-fuel performance data and irradiated-fuel samples for post-irradiation examination (PIE) and safety testing. The test train contains five separate capsules that were independently controlled and monitored. Each capsule contains multiple 12.51-mm-long compacts filled with low enriched uranium carbide/oxide (UCO) TRISO fuel particles. The primary objective of the AGR-5/6 test (Capsules 1, 2, 4, and 5) is to verify successful performance of the reference-design fuel under normal operating conditions. The AGR-7 test (Capsule 3) was designed to explore fuel performance at higher temperatures to demonstrate the capability of the fuel to withstand conditions beyond normal operating conditions in support of plant design and licensing. AGR 5/6/7 will also provide irradiated-fuel performance data on fission-gas release from failed particles during irradiation. The AGR-5/6/7 capsules were irradiated in the ATR northeast flux trap location. The experiment began on February 16, 2018 and ended on July 22, 2020, spanning nine ATR cycles over two and a half years. Thus, the AGR-5/6/7 fuel compacts were irradiated for a total of 360.9 effective full power days. The AGR 5/6/7 experiment was able to remain in the reactor core during all three Powered Axial Locator Mechanism (PALM) cycles (163A, 165A, and 167A) without overheating its fuel compacts. This report includes irradiation monitoring data from nine ATR Cycles: 162B, 163A, 164A, 164B, 165A, 166A, 166B, 167A, and 168A, as stored in the Nuclear Data Management and Analysis System (NDMAS). During irradiation, data records consisted of instantaneous measurements recorded every minute and provided by text files automatically every 2 hours. The AGR 5/6/7 data streams addressed in this report include thermocouple (TC) temperatures, sweep gas data (flow rates [capsule inlet, outlet, and downstream at detector], pressure, and moisture content), and Fission Product Monitoring System (FPMS) data (release rates and release to birth rate ratios [R/Bs]) for each of the five capsules. A total of 94,989,908 TC temperature and sweep gas data records were received and processed by NDMAS for AGR 5/6/7 irradiation. Of these records, 41,593,387 (or 43.7% of the total) met data collection and accuracy requirements and are labeled as Qualified. A total of 57,746,693 TC temperature readings were captured from 54 installed TCs. Among them, 10,034,676 TC temperature records (only 17.4%) were Qualified and 47,701,371 TC temperatures (or 82.6%) are Failed due to 48 TC failures (63.5%) and due to missing values (19.1%). To assess performance of the operational TCs, analysis of daily correlations between TCs found no evidence of virtual junction failure for any TCs. Analyses on control charts of TC temperature differences revealed trending in TC readings for TC2, 4, 5, and 13 in Capsule 3, but there is no conclusive indication of TC drift failure that caused those trends. Therefore, TC control charts are not used to disqualify TC data, but only for users’ consideration. For sweep gas flow rates, a total of 31,519,747 gas flow records (84.4%) are Qualified for use for AGR-5/6/7 experiment; 5,723,468 gas flow records (15.4%) are Failed due mostly to missing values; and 74,641 high sweep gas flow rates (0.2 %) are Trend. A large number of Failed missing TC temperature and gas flow values were caused by an error in the data output script that outputted a ‘NULL’ value when values were unchanged. This problem was fixed during the outage of Cycle 166B, which led to a substantially decreased number of missing values during the last three cycles. Nonetheless, a large amount of non-missing data remained because of the high data acquisition frequency (1-minute) and still provided sufficient data to effectively monitor the experiment as designed. For FPMS data, NDMAS received and processed fission product release and R/B data for nine ATR cycles, when ATR core reached full power during AGR 5/6/7 irradiation. These data consist of 110,388 release rate records and 110,388 R/B records for the twelve radionuclides (Kr 85m, Kr 87, Kr 88, Kr 89, Kr 90, Xe 131m, Xe 133, Xe 135, Xe 135m, Xe 137, Xe 138, and Xe 139) for each of the five capsules. Equivalent numbers of uncertainty records associated the release rates and R/B values were provided. To date, qualification status of the FPMS data stored in the NDMAS dat

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Quantitative intramolecular fission in oligoacenes, materials, and methods of use thereof

The present invention provides soluble, stable singlet fission (SF) compounds, compositions, materials, methods of their use, and methods for their preparation that provide efficient intramolecular singlet fission (iSF) and multiple excitons. The SF compound may be a dimer, an oligomer, or a polymer of polyoligoacenes, where for example, the compound achieves a triplet yield reaching about 200% per absorbed photon. In this system, SF does not depend on intermolecular inter-actions. Instead, SF is an intrinsic property of the molecule and therefore occurs independent of intermolecular interactions. Singlet fission has the potential to significantly improve the photocurrent in single junction solar cells and thus raise the Shockley-Queisser power conversion efficiency limit from about 33% to about 46% or greater. Quantitative SF yield at room temperature has only been observed in crystalline solids or aggregates of higher acenes.

36 MATERIALS SCIENCE↗
DOE OSTI2023

The value of fusion energy to a decarbonized United States electric grid

Fusion could be a part of future decarbonized electricity systems, but it will need to compete with other technologies. In particular, pulsed tokamak plants have a unique operational mode, and evaluating which characteristics make them economically competitive can help select between design pathways. Using a capacity expansion and operations model, here we determined cost thresholds for pulsed tokamaks to reach a range of penetration levels in a future decarbonized US Eastern Interconnection. The required capital cost to reach a fusion capacity of 100 GW varied from $2,700 to $7,500 kW -1 , and the equilibrium penetration increases rapidly with decreasing cost. The value per unit power capacity depends on the variable operational cost and on the cost of its competition, particularly fission, much more than on the pulse cycle parameters. These findings can therefore provide initial cost targets for fusion more generally in the United States.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Materials Assessment for FLiBe Fusion Blankets

This evaluation was prepared at the request of DOE-FES and expands on an earlier evaluation prepared by Wilson, Wiffen and Keiser in 2015 (attached as Appendices A and B). Additional useful information for materials in fission molten salt reactor systems that use FLiBe can be found in a prior NRC report [Busby 2019]. The purpose of this report is to review the considerations that will determine the fusion blanket structural materials that will be suitable for use in a FLiBe breeder-coolant fusion power plant (FPP). Identifying a fusion-relevant structural material that is sufficiently compatible with FLiBe is a key issue to be addressed. The constraints and trade-offs made in structural materials selection include the usual power generation requirements such as mechanical properties, fabricability and durability, which assumes reasonable compatibility with the working fluid. For an FPP, additional constraints include neutron economy (breeding ratio), resistance to irradiation damage and low activation (i.e. waste disposal) characteristics.

36 MATERIALS SCIENCE↗
NASA NTRS1976

Parametric analyses of planned flowing uranium hexafluoride critical experiments

Analytical investigations were conducted to determine preliminary design and operating characteristics of flowing uranium hexafluoride (UF6) gaseous nuclear reactor experiments in which a hybrid core configuration comprised of UF6 gas and a region of solid fuel will be employed. The investigations are part of a planned program to perform a series of experiments of increasing performance, culminating in an approximately 5 MW fissioning uranium plasma experiment. A preliminary design is described for an argon buffer gas confined, UF6 flow loop system for future use in flowing critical experiments. Initial calculations to estimate the operating characteristics of the gaseous fissioning UF6 in a confined flow test at a pressure of 4 atm, indicate temperature increases of approximately 100 and 1000 K in the UF6 may be obtained for total test power levels of 100 kW and 1 MW for test times of 320 and 32 sec, respectively.

Rodgers, R. J.↗
NASA NTRS1993

Advanced nuclear thermal propulsion concepts

In 1989, a Presidential directive created the Space Exploration Initiative (SEI) which had a goal of placing mankind on Mars in the early 21st century. The SEI was effectively terminated in 1992 with the election of a new administration. Although the initiative did not exist long enough to allow substantial technology development, it did provide a venue, for the first time in 20 years, to comprehensively evaluate advanced propulsion concepts which could enable fast, manned transits to Mars. As part of the SEI based investigations, scientists from NASA, DoE National Laboratories, universities, and industry met regularly and proceeded to examine a variety of innovative ideas. Most of the effort was directed toward developing a solid-core, nuclear thermal rocket and examining a high-power nuclear electric propulsion system. In addition, however, an Innovative Concepts committee was formed and charged with evaluating concepts that offered a much higher performance but were less technologically mature. The committee considered several concepts and eventually recommended that further work be performed in the areas of gas core fission rockets, inertial confinement fusion systems, antimatter based rockets, and gas core fission electric systems. Following the committee's recommendations, some computational modeling work has been performed at Los Alamos in certain of these areas and critical issues have been identified.

Howe, Steven D.↗
DOE OSTI2024

Flow Sensor Test Article (F-STAr) - Water Testing and Commissioning Report for FY2024

Argonne National Laboratory’s Mechanism Engineering Test Loop (METL) facility is developing new experimental testing capabilities with the Flow Sensor Test Article (F-STAr). F-STAr’s mission is to provide high flowrate, sodium submersible testing capabilities to support the development of sodium cooled nuclear fission reactors. Figure 1 shows a model of F-STAr (left) and a photo of the partially assembled test article (right) with several of the main components labeled. The system currently includes a large pump with desired flowrates of up to 100 GPM; an immersion heater with a maximum power of 5 kW; an immersion cooler with an estimated maximum cooling power of 2 kW; and an experimental test section to support sub-test articles. Initially, F-STAr will be setup for testing flow sensors, specifically Eddy Current Flow Sensors (ECFS). However, the system can be reconfigured to support a wide range of testing needs. For example, F-STAr could be setup to complete testing of hydraulic components, heat exchangers, hydrodynamic bearings, seals, and more. This report will provide an update on the development of F-STAr with a focus on the water testing and commissioning activities during Fiscal Year 2024 (FY24). First, the report will focus on describing the pump development. Specifically, updates will be provided on troubleshooting the pump’s excessive vibrational issues. In addition to updates on the pump, pressure-flowrate performance testing in water will be described. Furthermore, updates will be provided on a pump shaft bushing development as well as rigging procedures. Future work and paths forward for F-STAr will be described. Finally, a short summary on the ECFS development will be described at the end of the report.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
NASA NTRS2011

Design of a Mechanical NaK Pump for Fission Space Power

Alkali liquid metal cooled fission reactor concepts are under development for spaceflight power requirements. One such concept utilizes a sodium-potassium eutectic (NaK) as the primary loop working fluid, which has specific pumping requirements. Traditionally, electromagnetic linear induction pumps have been used to provide the required flow and pressure head conditions for NaK systems but they can be limited in performance, efficiency, and number of available vendors. The objective of the project was to develop a mechanical NaK centrifugal pump that takes advantages of technology advances not available in previous liquid metal mechanical pump designs. This paper details the design, build, and performance test of a mechanical NaK pump developed at NASA Marshall Space Flight Center. The pump was designed to meet reactor cooling requirements using commercially available components modified for high temperature NaK service.

Mireles, Omar R.↗
NASA NTRS2018

Human Mars Surface Mission Nuclear Power Considerations

A key decision facing Mars mission designers is how to power a crewed surface field station. Unlike the solar-powered Mars Exploration Rovers (MER) that could retreat to a very low power state during a Martian dust storm, human Mars surface missions are estimated to need at least 15 kilowatts of electrical (kWe) power simply to maintain critical life support and spacecraft functions. 'Hotel' loads alone for a pressurized crew rover approach two kWe; driving requires another five kWe-well beyond what the Curiosity rover’s Radioisotope Power System (RPS) was designed to deliver. Full operation of a four-crew Mars field station is estimated at about 40 kWe. Clearly, a crewed Mars field station will require a substantial and reliable power source, beyond the scale of robotic mission experience. This paper explores the applications for both fission and RPS nuclear options for Mars.

Rucker, Michelle A.↗
DOE OSTI2022

Application of SCALE to Molten Salt Fueled Reactor Physics in Support of Severe Accident Analyses

As part of a US Nuclear Regulatory Commission–sponsored project to assess the modeling and simulation capabilities for accident progression, source term, and consequence analysis for advanced reactor technologies with SCALE and MELCOR, SCALE was used for the modeling and simulation of a molten salt-fueled reactor (MSR). SCALE capabilities for the modeling of MSR physics were demonstrated based on the Molten Salt Reactor Experiment (MSRE). Of primary interest were the determination of the system’s nuclide inventory, as well as the inventories in the various regions of the loop, considering that the fuel is continuously pumped through the system. This report contains discussions on the following: 1. Determination of the system-average fuel salt inventory considering fission gas removal in the off-gas system and noble metal removal through plating out at the heat exchanger using recent enhancements in SCALE’s depletion sequence TRITON, 2. Assessment of the nuclide spatial distribution throughout the loop using SCALE’s depletion solver ORIGEN, 3. Calculation of the core’s power profile, flux profile, temperature reactivity coefficients, and xenon reactivity using full-core calculations with SCALE’s Monte Carlo code KENO-VI. The results obtained with SCALE were post-processed to provide the MELCOR team with the core inventory and decay heat of the system, as well as the inventory and decay heat of individual regions in the loop, a zone-wise power profile, temperature feedback coefficients, and the xenon worth.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
NASA NTRS2017

Orbital Space Solar Power Option for a Lunar Village

One of the most significant challenges to the implementation of a continuously manned lunar base is power. During the lunar day (14 Earth days), it is conceptually simple to deploy solar arrays to generate the estimated 35 kilowatts of continuous power required. However, generating this level of power during the lunar night (also 14 Earth days) has been an extremely difficult problem to solve. Conventional solutions range from the requirement that the base be located at the lunar south pole so as to take advantage of the continuous sunshine available there to developing a space-qualified nuclear reactor and power plant to generate the needed energy. There is a third option: Use the soon-to-be-available Space Launch System to place a space based solar power station in lunar orbit that would beam the needed energy to the lunar base. Several detailed studies have been performed by NASA, universities and others looking at the lunar south pole for locating the base. The results are encouraging: by taking advantage of the moon's orbital tilt, large solar arrays can be deployed there to track the sun continuously and generate the power needed to sustain the base. The problem with this approach is inherent to its design: it will only work at the lunar south pole. There is no other site on the Moon with geometry favorable to generating continuous solar power. NASA has also considered the development of a compact fission reactor and power plant to generate the needed power, allowing the base to be sited anywhere on the Moon. The problem with this approach is that there are no space fission reactors available, none are being planned and the cost of developing one is prohibitively expensive. Using an orbiting space based solar power station to generate electrical power and beam it to a base sited anywhere on the moon should therefore be considered. The technology to collect sunlight, generate greater than the estimated 35 kilowatts of power, and beam it to the surface using microwaves is available today. The problem with this concept in the past would have been the mass and packaging volume (for launch) required to put such a system in place in lunar orbit. This problem is potentially solved with the advent of the Space Launch System (SLS). The SLS, with its 70 mT launch capacity, it more than capable of placing such a system into lunar orbit in a single launch. This paper will examine the potential use of an SLS-launched, space solar power system in lunar orbit as the primary power source for a first-generation, continuously-occupied lunar base and compare it with the other power generation and storage options previously considered.

Johnson, Les↗
NASA NTRS2005

Multiple Restart Testing of a Stainless Steel Sodium Heat Pipe Module

A heat pipe cooled reactor is one of several candidate reactor cores being considered for space power and propulsion systems to support future space exploration activities. Long life heat pipe modules. with designs verified through a combination of theoretical analysis and experimental evaluations. would be necessary to establish the viability of this option. A hardware-based program was initiated to begin experimental testing of components to verify compliance of proposed designs. To this end, a number of stainless steel/sodium heat pipe modules have been designed and fabricated to support experimental testing of a Safe Affordable Fission Engine (SAFE) project, a 100-kWt core design pursued jointly by the Marshall Space Flight Center and the Los Alamos National Laboratory. One of the SAFE heat pipe modules was successfully subjected to over 200 restarts. examining the behavior of multiple passive freeze/thaw operations. Typical operation included a 1-hour startup to an average evaporator temperature of 1000 K followed by a 15 minute hold at temperature. Nominal maximum input power during the hold period was 1.9 kW. Between heating cycles the module was cooled to less than 325 K, returning the sodium to a frozen state in preparation fop the next startup cycle.

Martin, James↗
NASA NTRS1988

Nuclear propulsion: A vital technology for the exploration of Mars and the planets beyond

The physics and technology issues and performance potential of various direct thrust fission and fusion propulsion concepts are examined. Next to chemical propulsion the solid core fission thermal rocket (SCR) is the olny other concept to be experimentally tested at the power (approx 1.5 to 5.0 GW) and thrust levels (approx 0.33 to 1.11 MN) required for manned Mars missions. With a specific impulse of approx 850 s, the SCR can perform various near-Earth, cislunar and interplanetary missions with lower mass and cost requirements than its chemical counterpart. The gas core fission thermal rocket, with a specific power and impulse of approx 50 kW/kg and 5000 s offers the potential for quick courier trips to Mars (of about 80 days) or longer duration exploration cargo missions (lasting about 280 days) with starting masses of about 1000 m tons. Convenient transportation to the outer Solar System will require the development of magnetic and inertial fusion rockets (IFRs). Possessing specific powers and impulses of approx 100 kW/kg and 200-300 kilosecs, IFRs will usher in the era of the true Solar System class speceship. Even Pluto will be accessible with roundtrip times of less than 2 years and starting masses of about 1500 m tons.

Borowski, Stanley K.↗
NASA NTRS1989

Nuclear propulsion - A vital technology for the exploration of Mars and the planets beyond

The physics and technology issues and performance potential of various direct thrust fission and fusion propulsion concepts are examined. Next to chemical propulsion the solid core fission thermal rocket (SCR) is the only other concept to be experimentally tested at the power (approx 1.5 to 5.0 GW) and thrust levels (approx 0.33 to 1.11 MN) required for manned Mars missions. With a specific impulse of approx 850 s, the SCR can perform various near-earth, cislunar and interplanetary missions with lower mass and cost requirements than its chemical counterpart. The gas core fission thermal rocket, with a specific power and impulse of approx 50 kW/kg and 5000 s offers the potential for quick courier trips to Mars (of about 80 days) or longer duration exploration cargo missions (lasting about 280 days) with starting masses of about 1000 m tons. Convenient transportation to the outer Solar System will require the development of magnetic and inertial fusion rockets (IFRs). Possessing specific powers and impulses of approx 100 kW/kg and 200-300 kilosecs, IFRs will usher in the era of the true Solar System class spaceship. Even Pluto will be accessible with roundtrip times of less than 2 years and starting masses of about 1500 m tons.

Borowski, Stanley K.↗
NASA NTRS2004

Main Power Distribution Unit for the Jupiter Icy Moons Orbiter (JIMO)

Around the year 2011, the Jupiter Icy Moons Orbiter (JIMO) will be launched and on its way to orbit three of Jupiter s planet-sized moons. The mission goals for the JIMO project revolve heavily around gathering scientific data concerning ingredients we, as humans, consider essential: water, energy and necessary chemical elements. The JIM0 is an ambitious mission which will implore propulsion from an ION thruster powered by a nuclear fission reactor. Glenn Research Center is responsible for the development of the dynamic power conversion, power management and distribution, heat rejection and ION thrusters. The first test phase for the JIM0 program concerns the High Power AC Power Management and Distribution (PMAD) Test Bed. The goal of this testing is to support electrical performance verification of the power systems. The test bed will incorporate a 2kW Brayton Rotating Unit (BRU) to simulate the nuclear reactor as well as two ION thrusters. The first module of the PMAD Test Bed to be designed is the Main Power Distribution Unit (MPDU) which relays the power input to the various propulsion systems and scientific instruments. The MPDU involves circuitry design as well as mechanical design to determine the placement of the components. The MPDU consists of fourteen relays of four different variations used to convert the input power into the appropriate power output. The three phase system uses 400 Vo1ts(sub L-L) rms at 1000 Hertz. The power is relayed through the circuit and distributed to the scientific instruments, the ION thrusters and other controlled systems. The mechanical design requires the components to be positioned for easy electrical wiring as well as allowing adequate room for the main buss bars, individual circuit boards connected to each component and power supplies. To accomplish creating a suitable design, AutoCAD was used as a drafting tool. By showing a visual layout of the components, it is easy to see where there is extra room or where the components may interfere with one another. By working with the electrical engineer who is designing the circuit, the specific design requirements for the MPDU were determined and used as guidelines. Space is limited due to the size of the mounting plate therefore each component must be strategically placed. Since the MPDU is being designed to fit into a simulated model of the spacecraft systems on the JIMO, components must be positioned where they are easily accessible to be wired to the other onboard systems. Mechanical and electrical requirements provided equally important limits which are combined to produce the best possible design of the MPDU.

Papa, Melissa R.↗
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↗