SEARCH · Search NASA

Results for “Transient Testing Reactor”

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.

SourceResearch connectionFrom yearThrough year

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

At least 271 records · Page 15

All sources NASA NTRS DOE OSTI

DOE OSTI2017

Room temperature mechanical properties of electron beam welded zircaloy-4 sheet

Room temperature mechanical properties of electron beam welded and plain Zircaloy-4 sheet (1.6mm thick) have been measured and compared. Various welding parameters were utilized to join sheet material. Electron beam welded specimens and as-received sheet specimens show comparable mechanical properties. Zr-4 sheet displays anisotropy; tensile properties measured for transverse display higher elastic modulus, yield strength, reduction of area and slightly lower ductility than for the longitudinal (rolling direction). Higher welding power increases the alloy’s hardness, elastic modulus and yield strength, with a corresponding decrease in tensile strength and ductility. The hardness measured at weld is comparable to the parent metal hardness. Hardness at heat-affected-zone is slightly higher. Electron microscopic examination shows distinct microstructure morphology and grain size at the weld zone, HAZ and parent metal. A correlation between welding parameters, mechanical properties and microstructural features was established for electron beam welded Zircaloy-4 sheet material.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2016

Benchmark Development for TREAT Minimum Critical Mass Core Loading to Support Validation of TREAT Operations

Benchmark experiment data is necessary to validate modeling and simulation activities to support Transient Reactor Experiment Test (TREAT) Facility restart and operations. Key measurements of interest include criticality, control rod worth, excess reactivity, and shutdown margin for varying core loading sizes. Benchmark evaluations are being developed according the guidelines provided in the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook) that can also support advanced modeling and simulation activities, new experiment design for advanced reactors and accident tolerant fuel, and potential low-enriched uranium conversion of the TREAT reactor. This summary discusses development activities supporting a benchmark evaluation of the TREAT minimum critical mass core loading, which will be submitted for peer-review and publication in the IRPhEP Handbook with the identifier TREAT-FUND-RESR-001. Detailed models of TREAT were developed and evaluated to prepare benchmark models of the minimum critical mass configuration. Calculations were performed using MCNP6.1 with ENDF/B-VII.1 nuclear data libraries. Discussion of results.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2021

TREAT Cask Stand Drawings for Unlimited Use

This report contains drawings of the HFEF-15 Cask stand in the TREAT Reactor Building (MFC-720) and the Cask Stand Tower on the south wall of the TREAT High Bay. Public release is sought so these can be sent to DOD-SCO and their contracted companies looking at options for fueling the PELE reactor prior to demonstration in the DOME or elsewhere at INL.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2016

Benchmark Progress towards TREAT M8CAL Core Loading to Support Validation of TREAT Operations

Benchmark experiment data is necessary to validate modeling and simulation activities to support Transient Reactor Experiment Test (TREAT) Facility restart and operations. Key measurements of interest include criticality, control rod worth, excess reactivity, and shutdown margin for varying core loading sizes. Benchmark evaluations are being developed according the guidelines provided in the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook) that can also support advanced modeling and simulation activities, new experiment design for advanced reactors and accident tolerant fuel, and potential low-enriched uranium conversion of the TREAT reactor. This summary discusses development activities supporting a benchmark evaluation of the TREAT M8CAL core loading, which will be submitted for peer-review and publication in the IRPhEP Handbook with the identifier TREAT-FUND-RESR-002. Detailed models of TREAT were developed and evaluated to prepare benchmark models of the minimum critical mass configuration. Calculations were performed using MCNP6.1 with ENDF/B-VII.1 nuclear data libraries. Discussion of results.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2022

DOE-ID-INL-20-007 R1

DOE-ID-INL-20-007 R1 for INL-20-023 R1.

99 GENERAL AND MISCELLANEOUS↗

DOE OSTI2022

Alabama A&M University Senior Project

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2023

Alabama A&M University Senior Project Cont'd

This entry is being conducted in order to review documents (drawings and pictures) necessary for senior students of Alabama A&M to conduct their senior project. This project will include helping design an upgraded operating system for the TREAT core clamping units. A previous has been completed for the project per STI # INL/MIS-22-69603 Rev:000, these additional items will further assist the students in there work.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2016

Capabilities development for transient testing of advanced nuclear fuels at TREAT

The TREAT facility is a unique capability at the Idaho National Laboratory currently being prepared for resumption of nuclear transient testing. TREAT is an air-cooled reactor where the driver core is made up of urania dispersed in graphite blocks and encapsulated in zirconium alloy cans. In addition to the reactor itself, the foundation for TREAT's capabilities will also include a suite of irradiation vehicles and supporting infrastructure to provide the desired specimen boundary conditions while supporting a variety of instrumentation needs. Although TREAT historically supported all kinds of nuclear fuels, including LWR's, SFR's, space nuclear propulsion, research reactor plate-type fuels, and even fuels for use in the driver core of other transient test reactors, the majority of TREAT's more-recent experiment programs revolved around SFR's and sodium-environment test capsules and loops. MK-series loop experiment-specific test trains could be outfitted with a variety of instruments to monitor test and specimen conditions while the fast neutron hodoscope collected information about fuel movement before, during, and after transient excursions. The SERTTA (Static Environment Rodlet Transient Test Apparatuses) devices will be general purpose vehicles capable of providing inert gas, water (up to PWR conditions), steam, or perhaps even molten sodium environments within vessel-like capsules. The Multi-SERTTA vehicle design is particularly novel as it makes use of the full active core length in TREAT to test four small rodlets (∼12 cm length), each within its own hermetic boundary, it will be capable of housing pre-irradiated fuel specimens. Additional modules and adaptations of Multi-SERTTA will enable rapid loop pressure blowdown to simulate LOCA conditions. The Super-SERTTA, akin to the Multi-SERTTA will offer more possibilities for simulating transient phenomena.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2016

Westinghouse Accident Tolerant Fuel Program

The Westinghouse accident tolerant fuel (ATF) program is aimed at providing a significant increase in fuel robustness able to increase its tolerance to beyond design basis accidents at an economically attractive price to nuclear power providers. The Westinghouse ATF program, which began in 2004, advanced significantly during the current first phase (1A and 1B) begun in 2012 of this Department of Energy (DOE) program. Phase 1B will culminate in the manufacture of fuel rods that will undergo a 6 year exposure at steady state conditions in the Advanced Test Reactor and the Halden test reactor (Phase 2) and in transient tests in TREAT and Halden. The goal is to load lead test rods in a commercial reactor in 2022. In this paper, the current status of the ATF development and future irradiation plan is discussed in detail. Westinghouse has picked the most viable concepts that will provide maximum accident tolerance as well as economics, and those that will make the most sense for utility adoption. All three options appear to be feasible from a technical and economic point of view. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2026

Development of a BISON validation case for the TRISO transient irradiations in NSRR using effective heat capacity methods

The current tristructural isotropic (TRISO) fuel assessment and validation database in BISON primarily covers steady- state irradiation and high-temperature furnace testing. Transient assessment cases are potentially needed to support U.S. industry efforts in designing and deploying commercial reactors using TRISO fuels. Historical transient tests in- volving TRISO fuels used highly conservative conditions compared to the typical high-temperature gas-cooled reactor accident scenarios. Despite this, modeling historical transient tests is fundamental for evaluating BISON’s predictive capabilities, adapting material properties for high-temperature and high-particle-power regimes, and developing a sys- tematic validation approach for TRISO transient applications. This work developed a 1D model of transient experiments carried out at the Nuclear Safety Research Reactor using BISON. BISON’s predictions of energy deposition, UO 2 melting onset, and molten volume fractions are compared against experimental measurements. Melting was modeled using an effective specific heat capacity model for UO 2 . We found that BISON’s predictions are in reasonable agreement with experimental data for low-energy-deposition cases, and that BISON overpredicts melting at higher energy depositions. We also discuss the potential causes of discrepancies between the simulated and measured results and propose ways to further develop the model. Although these simulations used conservative conditions compared to those expected for actual TRISO-fueled reactors, they extended the range of conditions reflected in the data in the existing BISON database for TRISO fuels.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2019

TREAT LEU Conversion Analyses Performed 2016-2018

TREAT is an experimental nuclear reactor designed to produce high neutron flux transients for the testing of reactor fuels and other materials. After more than 20 years of non-operational standby, TREAT has been restarted under the U.S. Department of Energy’s Resumption of Transient Testing Program. The conversion of TREAT from its current highly enriched uranium fuel to a low enriched uranium fuel is being pursued in keeping with the mission of the Department of Energy National Nuclear Security Administration’s Material Management and Minimization Reactor Conversion Program. This report documents additional analyses performed over the timeframe of 2016 to early 2018 to evaluate the LEU Conceptual Design for the TREAT reactor, following the initial design studies documented in the 2016 Conceptual Design Report. The analyses documented in this report include neutronic and thermal hydraulic sensitivity studies to evaluate the effect of key design parameters on predicted core behaviors, as well as a simple investigation into the effect of fuel thermal expansion.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2017

Transient Simulation of the Multi-SERTTA Experiment with MAMMOTH

This work details the MAMMOTH reactor physics simulations of the Static Environment Rodlet Transient Test Apparatus (SERTTA) conducted at Idaho National Laboratory in FY-2017. TREAT static-environment experiment vehicles are being developed to enable transient testing of Pressurized Water Reactor (PWR) type fuel specimens, including fuel concepts with enhanced accident tolerance (Accident Tolerant Fuels, ATF). The MAMMOTH simulations include point reactor kinetics as well as spatial dynamics for a temperature-limited transient. The strongly coupled multi-physics solutions of the neutron flux and temperature fields are second order accurate both in the spatial and temporal domains. MAMMOTH produces pellet stack powers that are within 1.5% of the Monte Carlo reference solutions. Some discrepancies between the MCNP model used in the design of the flux collars and the Serpent/MAMMOTH models lead to higher power and energy deposition values in Multi-SERTTA unit 1. The TREAT core results compare well with the safety case computed with point reactor kinetics in RELAP5-3D. The reactor period is 44 msec, which corresponds to a reactivity insertion of 2.685% delta k/k$. The peak core power in the spatial dynamics simulation is 431 MW, which the point kinetics model over-predicts by 12%. The pulse width at half the maximum power is 0.177 sec. Subtle transient effects are apparent at the beginning insertion in the experimental samples due to the control rod removal. Additional difference due to transient effects are observed in the sample powers and enthalpy. The time dependence of the power coupling factor (PCF) is calculated for the various fuel stacks of the Multi-SERTTA vehicle. Sample temperatures in excess of 3100 K, the melting point UO$_2$, are computed with the adiabatic heat transfer model. The planned shaped-transient might introduce additional effects that cannot be predicted with PRK models. Future modeling will be focused on the shaped-transient by improving the control rod models in MAMMOTH and adding the BISON thermo-elastic models and thermal-fluids heat transfer.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2019

Summary Report for Out-of-Pile Transient Testing Database Development

The DOE Advanced Reactor Technology program has supported knowledge preservation efforts to recover and preserve fuel data from the US sodium cooled fast reactor (SFR) program. Efforts have focused on establishing databases of experimental data generated during the Integral Fast Reactor program (at EBR-II, FFTF, and TREAT) as well as during related out-of-pile examinations of the fuel irradiated at EBR-II. The information in these databases is essential to support further development and licensing activities for advanced fast reactor designs. OPTD is a database of out-of-pile furnace transient tests on metallic fuel. It includes records of 62 tests on irradiated fuel pins conducted in furnace apparatuses constructed in the Alpha-Gamma Hot Cell Facility at ANL. This report describes efforts during FY19 to improve visibility and access, expand the content of, and improve the user experience with the FIPD and OPTD databases.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2023

Sizing and Selection of Pressure Relief Valves for High-Pressure Thermal–Hydraulic Systems

This study covers the critical concerns related to the sizing, selection, installation, maintenance, and testing of pressure safety valves (PSVs). The aim is to ensure the safety of pressurized systems, hydrostatic transmission systems, and hydraulic plants, including process plants, thermal power plants, and nuclear reactor systems. PSVs are devices that ensure the safety and reliability of pressurized vessels, lines, and systems during overpressure events. The task of selecting which PSV features are of greatest value for a specific purpose is complex—especially in the design of a high-pressure experimental thermal–hydraulic facility for hydrostatic and transient testing of the reactor system—when the systems are in the design and development phases and require qualification and demonstration to prove that they have reached a given level of technological readiness. The present study highlights the required steps for users to follow the associated rules, guidelines, and recommendations. As a part of this research, case studies are presented to help readers better understand the applicable strategy and standards. A discussion and a review of PSV performance degradation and failure are summarized to provide a better understanding of varied process applications and conditions, including fluid flow dynamics, boundary-layer formation and pressure drops, gas bubble formation and collapse, geometric configurations, inlet/outlet piping, abrupt pressure fluctuations, and acoustic resonance. Moreover, this study discusses the servicing and testing of PSVs in a multiphase pressurized system. Overall, it provides a basic overview of how PSVs ensure the safety of pressurized systems, supported by case studies and industrial practices.

02 PETROLEUM↗

DOE OSTI2020

Foreword Special issue on the Kilopower Project, Kilowatt Reactor Using Stirling TechnologY (KRUSTY) Test

This special issue of Nuclear Technology contains full-length, peer-reviewed papers describing the design work leading up to and the results of the Kilowatt Reactor Using Stirling TechnologY (KRUSTY) test. KRUSTY was the centerpiece of the National Aeronautics and Space Administration (NASA) Kilopower Project to design, build, and test a space nuclear reactor. This test was the first such test since the end of the Space Nuclear Auxiliary Power (SNAP) project at the end of the 1960s. In this issue, the introduction paper presents the goals of the Kilopower Project and the potential missions this reactor concept could serve in NASA. Kilopower was intended to serve both human exploration needs on planetary surfaces as well as science needs for deep-space exploration. The design work for the experiment by Poston and the power conversion development by Gibson present the pre-work required to perform the eventual KRUSTY test. A paper on regulatory analysis follows, to show the path used to gain approval of the proposed experiment. Then, the early zero-power critical experiments are presented by Sanchez and Grove. These experiments were essential data used to enhance model predictions prior to the high-temperature test. Next, Poston presents the three experiments (warm criticals) that increase the temperature in an incremental fashion prior to the final experiment. These experiments were used to achieve final regulatory approval of the final high-temperature experiment. The last paper by Poston presents the results of the steady-state and transient testing of the reactor at full power and at the design temperature. These results show that the reactor design and as-built experiment met all of the requirements that NASA had developed for the system. Any experiment of this magnitude was accomplished only because of the hard work and dedication of a large number of people at multiple institutions, including the NASA Glenn Research Center, NASA Marshall Space Flight Center, Y-12 National Security Site, Los Alamos National Laboratory, and the Nevada National Security Site prime contractor (formerly National Security Technologies and now the Mission Support and Test Services). The project was jointly funded by the Space Technology Mission Directorate at NASA and the Criticality Safety Program at the National Nuclear Security Administration (NNSA). A special thanks to Jerry McKamy and Angela Chambers at the NNSA and Lee Mason and Don Palac at NASA for their support and guidance. For the authors and multitude of staff that performed this work, the passion for space nuclear reactors was the key to making this experiment a reality.

99 GENERAL AND MISCELLANEOUS↗

DOE OSTI2020

Modular high temperature gas reactor core modeling with RELAP5-3D/PHISICS – Optimization schemes for load following

The objective of this study is to develop a 3D/NK model to study the load following operation of a Modular High-Temperature Gas-Cooled Reactor (MHTGR). The selected design is the 350 MW prismatic, graphite moderated, helium cooled thermal reactor based on the MHTGR-350 transient benchmark led by the High-Temperature Gas-Cooled Reactor (HTGR) Methods Core Simulation Group at Idaho National Laboratory (INL) in the framework of the Organization for Economic Cooperation and Development Nuclear Energy Agency (OECD/NEA). This kind of reactor usually reacts very slowly to the perturbation of the core parameters due to the large amount of graphite in the core. This kind of behavior promotes the reactor stability but on the other hand limits the reactor load-following operability. Recently advances in gas reactor designs have made load-following in gas reactors a feasible and achievable goal. Modern nuclear reactors (such as the MHTGR-350) are designed to withstand the structural stresses associated with load-following. Operating the reactor in load-following mode will inevitably reduce the load factor. Although a higher load factor means more revenue and the best usage of the fuel, following the variable energy demand will increase the maximum achievable fraction of generated nuclear power, being no more limited to the base load power generation. In general, if low impact on material aging and the safe operability are demonstrated the plant economics will be minimally affected. In order to ensure that the system can be safely operated in a load-following mode an extensively study has been carried out. Some tests with linear coolant mass flow rate change demonstrated that the reactor behavior is suitable for operation in load following mode. This assumption has been confirmed by a 4 days load following transient test in which the reactor supplied the requested power with a negligible error.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2017

HELIOS vs. DRAGON – Code-to-Code Comparison for High Temperature Gas Reactors

Full core neutronics modeling for depletion and transient analysis in commercial and test reactors requires a lattice physics code to obtain material homogenized and energy group condensed cross- sections at an assembly level. These cross-sections are then used in a full-core model for depletion and transient analysis. HELIOS and DRAGON are two popular lattice physics codes that are capable of modeling the HTGR, which is not straight forward due to the large neutron mean free path in graphite, the presence of burnable poisons, proximity to reflectors, and the double heterogeneous nature of prismatic TRISO fuel. To understand on how HELIOS and DRAGON perform, a simple hexagonal fuel pin cell based on the MHTGR design was analyzed. In total, there are three benchmark cases, with each case increasing in complexity. These cases were selected to help distinguish and identify limitations and benefits of each reactor physics code.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2023

Code Benchmark of the HTTF Pressurized Conduction Cooldown Test Using SAM

The High Temperature Test Facility (HTTF) at Oregon State University is an integral system test facility to simulate postulated reactor transients of prismatic high-temperature gas-cooled reactors(HTGRs). A series of test campaigns was launched, providing abundant test data that could be used to benchmark reactor system analysis codes like the System Analysis Module (SAM). In this study, a SAM model of the facility is developed based on the two-dimensional (2D) ring model approach. All components including the ceramic matrix, graphite heaters, helium coolant channels, core barrel, upcomer, pressure vessel, and reactor cavity cooling system are modeled as concentric cylindrical rings. The model is used to simulate one of the benchmark problems-Pressurized Conduction Cooldown (PCC)-within the scope of the Organisation for Economic Co-operation and Development Nuclear Energy Agency International HTTF Benchmark. The simulations consist of two parts. In the first part, operating and boundary conditions as well as thermophysical properties of materials are specified for the benchmark problem. In this work, results from the first part will be used in code-to-code comparison. In the second part, the SAM model is used to simulate Test PG-27, which is the first PCC test carried out in the HTTF, with only two of the ten heater banks activated. The results in the second part are used for code-to-data comparison. Because the helium coolant flow rate is not measured in this facility, it is estimated using the input power and inlet/outlet coolant temperatures. Additionally, radial heat flow in the ceramic blocks is complicated by hundreds of cylindrical coolant channels and heater rods embedded in them. As such, it is necessary to deduce an effective thermal conductivity for the ceramic to analyze the core thermal behavior. SAM predictions of the helium coolant and ceramic temperatures are compared with test data measured in three equivalent sectors. Overall, the SAM results agree reasonably well with test data within the variation of data among the three sectors, which demonstrates SAM's capability in capturing transient effects in HTGR using the simplified 2D ring model.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗