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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.

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At least 325 records · Page 18
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DOE OSTI2018

Gear Test Assembly - Status Report FY2018

The gear test assembly completed testing in air and distilled water successfully under a range of loading conditions. An external torque sensor assembly was designed and added to confirm the torque readings provided by the Parker control system. Hand calculations were performed to determine the time range that the heaters could be turned on without exceeding the ASME Boiler and Pressure Vessel code Fatigue Stress Limits for 1,000,000 cycles. Transient computational fluid dynamics analyses were performed to determine the heater cycle on/off time. Steady state computational fluid dynamics analyses were performed to determine the final temperature distribution throughout the insulated assembly using a thinner but more effective insulation configuration. The initial heating tests were completed (without insulation installed) to check the heater control system. Testing of the argon pressurized seal chambers were performed and indicated the need for re-machining of several components to eliminate some leaks. Nondestructive examination of the gears continued successfully using eddy current and ultrasonic inspection techniques.

42 ENGINEERING↗
DOE OSTI2019

Annual Status Report (FY 2018): Composite Analysis for Low Level Waste Disposal in the Central Plateau of the Hanford Site

In accordance with U.S. Department of Energy (DOE) requirements in DOE M 435.1-1 Chg 11, and as implemented by DOE/RL-2000-292, the DOE Richland Operations Office has prepared this annual summary of the Hanford Site Composite Analysis for fiscal year (FY) 2018. The Hanford Site Composite Analysis was originally reported in PNNL-118003 and PNNL-11800, Addendum 14 (hereinafter referred to collectively as the Hanford Site Composite Analysis). The Hanford Site Composite Analysis was approved by a 2002 memorandum. As required by DOE/RL-2000-29, an annual evaluation of new information and data developed by a number of onsite programs was completed. The reporting period for this annual evaluation is FY 2018 (October 1, 2017, through September 30, 2018).

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Annual Status Report (FY 2018): Performance Assessment for the Disposal of Low-Level Waste in the 200 West Area Burial Grounds (Rev. 0)

This annual review provides the projected dose estimates of radionuclide inventories disposed in the active 200 West Area Low-Level Waste Burial Grounds (LLBGs) since September 26, 1988. These estimates are calculated using the original dose methodology developed in the performance assessment (PA) analysis. The estimates are compared with performance objectives defined in U.S. Department of Energy (DOE) requirements (DOE 435.1 Chg 1 and its companion documents DOE M 435.1-1 Chg 1 and DOE G 435.1-1). All performance objectives are currently satisfied, and operational waste acceptance criteria and waste acceptance practices continue to be sufficient to maintain compliance with performance objectives. Inventory estimates and associated dose estimates from future waste disposal actions are unchanged from previous years' evaluations that indicate potential impacts well below performance objectives; therefore, future compliance with DOE O 435.1 Chg 1 is expected. Within the active burial grounds, low-level waste and mixed low-level waste currently may be disposed only in two lined trenches in the 218-W-5 Burial Ground (Trenches 31 and 34) until they are either filled or a decision is made to close these trenches. Some mixed low-level waste is also disposed at the Environmental Restoration Disposal Facility in the 200 West Area (which is covered under a separate PA). During this reporting period (fiscal year 2018, from October 1, 2017, through September 30, 2018), waste was disposed to the 200 West Area LLBGs. Results from the multi-year diffusion experiments are summarized for this reporting period to quantify the efficacy of concrete waste forms in retaining key radionuclides (e.g., uranium-238, technetium-99, and iodine-123). The results of sorption experiments initiated in FY 2017 to evaluate sorption coefficients for iodine and technetium are also summarized. The test durations ranged from 1 to 6 months. Continued groundwater monitoring of the 200 West Area LLBGs indicates no groundwater contamination due to LLBG waste. Current assumptions about future land use at the Hanford Site are consistent with PA analysis1 assumptions of a post-closure facility that will not be degraded by human activity. The LLBGs are located in an area identified for waste management and containment of residual contamination. Overall, there are no substantive changes to primary PA assumptions and no changes to the PA analysis conclusion; therefore, compliance with DOE O 435.1 Chg 1 is being maintained.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Annual Status Report (FY 2018): Performance Assessment for the Disposal of Low-Level Waste in the 200 East Area Burial Grounds (Rev. 0)

This annual review provides the projected dose estimates of radionuclide inventories disposed in the active 200 East Area Low-Level Waste Burial Grounds (LLBGs) since September 26, 1988. These estimates are calculated using the original dose methodology developed in the performance assessment (PA) analysis. The estimates are compared with performance objectives defined in U.S. Department of Energy (DOE) requirements (DOE O 435.1 Chg 1 and companion documents DOE M 435.1-1 Chg 1 and DOE G 435.1-1). All performance objectives are currently satisfied, and operational waste acceptance criteria and waste acceptance practices continue to be sufficient to maintain compliance with performance objectives. Inventory estimates and associated dose estimates from future waste disposal actions are unchanged from previous years' evaluations that indicate potential impacts well below performance objectives; therefore, future compliance with DOE O 435.1 Chg 1 is expected. Within the active burial grounds, low-level waste and mixed low-level waste currently may be disposed only in the dedicated U.S. Navy reactor compartment trench in the 218-E-12B Burial Ground (Trench 94). Naval reactor compartment disposal at Trench 94 will continue. During this reporting period (fiscal year 2018, from October 1, 2017, through September 30, 2018), two reactor compartments were disposed in Trench 94. Results from multi-year diffusion experiments are summarized for this reporting period to quantify the efficacy of concrete waste forms in retaining key radionuclides (e.g., uranium-238, technetium-99, and iodine-129). The results of sorption experiments initiated in FY 2017 to evaluate sorption coefficients for iodine and technetium are also summarized. The test durations ranged from 1 to 6 months. Continued groundwater monitoring of the 200 East Area LLBGs indicates no groundwater contamination due to LLBG waste. Current assumptions about future land use at the Hanford Site are consistent with PA analysis assumptions of a post-closure facility that will not be degraded by human activity. The LLBGs are located in an area identified for waste management and containment of residual contamination. This area will remain after final environmental remediation and the proposed shrinkage of Hanford Site boundaries to small sections within the 200 East Area and 200 West Area in the Central Plateau. Overall, there are no substantive changes to primary PA assumptions and no changes to the PA analysis conclusion; therefore, compliance with DOE O 435.1 Chg 1 is being maintained.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Annual Status Report (FY 2018): Performance Assessment for the Environmental Restoration Disposal Facility

DOE O 435.1 and DOE M 435.1-1 require that a determination of continued adequacy of the performance assessment (PA), composite analysis, and disposal authorization statement be made on an annual basis, and it must consider the results of data collection and analysis from research, field studies, and monitoring. Beginning in 1996, the Environmental Restoration Disposal Facility (ERDF) started accepting low-level radioactive, hazardous, and mixed wastes that were generated during cleanup activities at the Hanford Site. ERDF is composed of a series of cells or disposal areas and can accommodate future design expansions as needed. Currently, there are 10 cells. During this reporting period (fiscal year 2018, which extends from October 1, 2017, through September 30, 2018), approximately 193,358 U.S. tons of waste was disposed at ERDF. From ERDF inception through September 30, 2018, approximately 18.3 million U.S. tons of waste had been disposed at ERDF, which equates to consumption of approximately 87% of the disposal volume. As a condition of the Disoposal Authorization Statement, disposal operations within ERDF must be in accordance with the waste acceptance criteria (WAC) that provides specific radionuclide disposal limits, waste form restrictions, and descriptions of acceptable waste packages in compliance with the requirements of DOE 435-1.1. The ERDF WAC stipulates that waste destined for disposal at ERDF be controlled based on source, physical form, and contaminant concentration and activity levels. Concentration limits included in the ERDF WAC were extracted from the ERDF PA. A sum of fraction analysis of of the ERDF inventory against the threshold concentration limits from inception through fiscal year 2018 produces a result <1. This result provides confirmation of compliance with PA imposed radionuclide limits.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Modeling K-DEMO Plasma Power Exhaust with UEDGE (Status Report through Oct. 2018)

The goal of this project is to develop simulation capability for the edge plasma region of K-DEMO in order to assess the heat and particle loads on plasma facing components. The roles of Lawrence Livermore National Laboratory (LLNL) are to provide the UEDGE simulation code for this purpose, to obtain initial edge-plasma solutions for KDEMO, and to train K-DEMO staff members in both the use of the 2D UEDGE and the code structure such that future modification can be made. A schematic of K-DEMO highlighting the power exhaust is shown in Fig. 1. The major radius of the device is 6.8 m, and the total power to be exhausted is 280 MW in the edge plasma region. Evaluating impact of different divertor geometries and the use of seeded impurity radiation to manage the peak heat flux to surfaces are key aspects being addressed.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2019

Complete Status Report Documenting the Processing, Microstructure and High-Temperature Mechanical Properties of the Nanostructured Ferritic Alloy OFRAC

This report provides a summary of the new nanostructured ferritic alloy, OFRAC (Oak Ridge Fast Reactor Advanced Fuel Cladding), that was developed for the advanced sodium fast reactor fuel cladding and core internals. The composition of OFRAC, which is Fe-12Cr-1Mo-03.Ti-0.3Nb + 0.3Y2O3 (wt. %), was chosen to achieve high temperature strength as well as high temperature creep performance and swelling resistance requirements that past structural materials such as austenitic stainless steels and ferritic/martensitic steels, i.e. HT-9, are not suitable as SFR cladding materials. The mechanical alloying procedures were described for producing OFRAC and the detailed characterization studies revealed uniformly distributed, high concentration of Y-Ti-O enriched oxide particles, or nanoclusters. The Nb addition that was intended to sequester interstitial C and N that are commonly introduced as contamination during ball milling of powder was found to be very effective in forming coarse size Nb carbonitrides along with excess Ti that removes these interstitial elements from the bcc-Fe matrix of OFRAC. The tensile tests on the OFRAC in the extruded condition showed excellent balance in strength and ductility from room temperature to 800°C. Both the microstructure and tensile properties were not significantly degraded after annealing the extruded OFRAC samples for 8 h at 1050°C and 1150°C. The latter temperature is ~76% of the melting point of OFRAC. These results demonstrated the stability of the microstructure and tensile properties to extremely high temperatures. The thermal creep resistance at elevated temperatures and stresses was investigated using strain rate jump testing. The results showed that OFRAC possessed high stress exponents at high stresses and temperatures of 550°C, 600°C, 700ºC and 800°C. This thermal dependence of minimum creep rate indicates that OFRAC will show quite low creep rates at higher temperatures and stresses compared to HT-9 and stainless steels. These significant improvements indicate that the NFA OFRAC alloy is a compelling and viable choice for sodium fast reactor fuel cladding.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Thermal Hydraulic Experimental Test Article: Status Report for FY2019

The Thermal Hydraulic Experimental Test Article (THETA) is a facility that will be used to develop sodium components and instrumentation as well as acquire experimental data for validation of reactor thermal hydraulic and safety analysis codes. The facility will simulate nominal conditions as well as protected/unprotected loss of flow accidents in an SFR. High fidelity distributed temperature profiles of the developed flow field will be acquired with Rayleigh backscatter based optical fiber temperature sensors. The facility is being designed in partnership with systems code experts to tailor the experiment to ensure the most relevant and highest quality data for code validation. Figure 1 illustrates the main components of the primary system. THETA will be installed in the Mechanisms Engineering Test Loop (METL) with the primary system in the 28 inch Test Vessel #4, the secondary system in the 18-inch Test Vessel #6, and its sodium-to-air heat exchanger will occupy the location for 18-inch Test Vessel #5 (Test Vessel #5 will not be installed), Figure 2. The design work for the facility has been completed and is currently in the procurement, manufacturing and initial testing stage. The THETA primary system, including the redan, inlet and outlet plenum, 38 kW electric core simulant heater, submerged flowmeter, pump, and all instrumentation has either been procured or is currently being tested at Argonne. The secondary sodium system for cooling the tube-side sodium of the intermediate heat exchanger has all been analyzed according to ASME code and is in the final development stages before procurement.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

Gear Test Assembly - Status Report For FY2019

The final stages of pre-sodium commissioning were completed for the Gear Test Assembly (GTA) in late 2018 and all the components were moved to building 308 for final preparations for insertion into the Mechanisms Engineering Test Loop (METL). The GTA was fully assembled in the experimental test assembly workstation and all supporting electrical and hardware systems were installed on the METL Mezzanine. The GTA was installed in Test Vessel 1 after performing a final systems check, and the vessel and article were raised to operating temperature and held for two weeks to bake-out any moisture present in the components. Test Vessel 1 was filled with sodium and a purification process was completed using the cold trap system. When the sodium was sufficiently clean, Test Vessel 1 was isolated from the rest of METL, and experimental operation of the GTA commenced. The experimental operations consisted of rotating two shafts using Parker servomotors to apply a prototypic torque of a fuel handling mechanism removing and inserting a sodium fast reactor fuel assembly. This removing and inserting motion was repeated until the system seized, indicating component degradation of either the gears or the bearings in the test gearbox. A total of 19,600 cycles were completed, corresponding to the removal and insertion of 9,800 fuel assemblies. The system was shut down and then prepared for removal. The Flexicask System was successfully used to remove the GTA from Test Vessel 1 while keeping both vessel and article inert during the process. The GTA was installed in the Carbonation System, where the process of slowly reacting the sodium away using moist CO 2 has been started. Once the GTA is fully cleaned, it will be disassembled and the gears will undergo the same NDE process performed before testing to quantify the wear that occurred. A report on this, along with the data analysis from the experimental operations will be submitted in late 2019.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

SNL Research into Stress Corrosion Cracking of SNF Dry Storage Canisters (FY19 Status Report)

This progress report describes work done in FY19 at Sandia National Laboratories (SNL) to assess the localized corrosion performance of container/cask materials used in the interim storage of spent nuclear fuel (SNF). Of particular concern is stress corrosion cracking (SCC), by which a through-wall crack could potentially form in a canister outer wall over time intervals that are shorter than possible dry storage times. Work in FY19 refined our understanding of the chemical and physical environment on canister surfaces and evaluated the relationship between chemical and physical environment and the form and extent of corrosion that occurs.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Thermal Hydraulic Experimental Test Article - Status Report for FY2019 (Rev. 1)

The Thermal Hydraulic Experimental Test Article (THETA) is a facility that will be used to develop sodium components and instrumentation as well as acquire experimental data for validation of reactor thermal hydraulic and safety analysis codes. The facility will simulate nominal conditions as well as protected/unprotected loss of flow accidents in an SFR. High fidelity distributed temperature profiles of the developed flow field will be acquired with Rayleigh backscatter based optical fiber temperature sensors. The facility is being designed in partnership with systems code experts to tailor the experiment to ensure the most relevant and highest quality data for code validation. The design work for the facility has been completed and is currently in the procurement, manufacturing and initial testing stage. The THETA primary system, including the redan, inlet and outlet plenum, 38 kW electric core simulant heater, submerged flowmeter, pump, and all instrumentation has either been procured or is currently being tested at Argonne. The secondary sodium system for cooling the tube-side sodium of the intermediate heat exchanger has all been analyzed according to ASME code and is in the final development stages before procurement.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

Discharge Permit-1845 Quarterly Status Report October 2019

Trichloroethene (TCE) and nitrate have been identified as constituents of concern in groundwater at the Sandia National Laboratories, New Mexico (SNL/NM) Technical Area (TA)-V Groundwater (TAVG) Area of Concern (AOC) based on detections above the U.S. Environmental Protection Agency (EPA) maximum contaminant levels in samples collected from monitoring wells. The maximum contaminant levels and the State of New Mexico drinking water standards, as specified in 20.6.2.3103 New Mexico Administrative Code (NMAC) for TCE and nitrate (as nitrogen) are 5 micrograms per liter (p,g/L) and 10 milligrams per liter (mg/L), respectively.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2019

Status Report on Uncertainty Quantification and Sensitivity Analysis Tools in the Geologic Disposal Safety Assessment (GDSA) Framework

The Spent Fuel and Waste Science and Technology (SFWST) Campaign of the U.S. Department of Energy (DOE) Office of Nuclear Energy (NE), Office of Fuel Cycle Technology (FCT) is conducting research and development (R&D) on geologic disposal of spent nuclear fuel (SNF) and high-level nuclear waste (HLW). Two high priorities for SFWST disposal R&D are design concept development and disposal system modeling. These priorities are directly addressed in the SFWST Geologic Disposal Safety Assessment (GDSA) control account, which is charged with developing a geologic repository system modeling and analysis capability, and the associated software, GDSA Framework, for evaluating disposal system performance for nuclear waste in geologic media. GDSA Framework is supported by SFWST Campaign and its predecessor the Used Fuel Disposition (UFD) campaign.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

MSRE TRANSFORM Model for Thermal-Hydraulic Benchmarking (Status Report)

Over the last several years, there has been renewed interest in the development of molten salt reactor concepts for the next generation of nuclear power plants. The modeling tools used to design these reactors must be validated in order to accurately predict the thermal hydraulic behavior expected during operation. The majority of the practical experience with molten salt reactors comes from the Molten Salt Reactor Experiment (MSRE) that operated at Oak Ridge National Laboratory (ORNL) from 1964 until 1969. The work discussed in this paper uses the design information and experimental data from that period to create and validate a detailed model of the MSRE. Once completed, this model’s performance will be compared to other system analysis codes for thermal hydraulic benchmarking, and their ability to predict the thermal hydraulic performance of molten salt reactor systems will also be compared.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Performance Assessment for the Environmental Restoration Disposal Facility (FY 2019 Annual Status Report)

DOE O 435.1 and DOE M 435.1-1 require that a determination of continued adequacy of the performance assessment (PA), composite analysis, and disposal authorization statement (DAS) be annually, and it must consider the results of data collection and analysis from research, field studies, and monitoring as well as update any Radioactive Waste Management Basis documents. Beginning in 1996, the Environmental Restoration Disposal Facility (ERDF) has accepted low-level radioactive, hazardous, and mixed wastes that were generated during cleanup activities at the Hanford Site. ERDF is composed of a series of cells (currently 10) or disposal areas and can accommodate future design expansions as needed. During this reporting period (fiscal year 2019, October 1, 2018, to September 30, 2019). approximately 139,721 U.S. tons of waste was disposed at ERDF. From inception through September 30, 2019, ERDF has received approximately 18.5 million U.S. tons of waste, which equates to consumption of approximately 88% of the disposal volume. As a condition of the DAS, disposal operations within ERDF must be in accordance with the waste acceptance criteria (WAC) that provide specific radionuclide disposal limits, waste from restrictions, and descriptions of acceptable waste packages in compliance with DOE M 435-1.1 requirements. The ERDF WAC stipulates that waste destined for disposal at ERDF be controlled based on source, physical form, and containment concentration and activity levels

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Composite Analysis for Low Level Waste Disposal in the Central Plateau of the Hanford Site (FY2019 Annual Status Report)

In accordance with DOE M 435.1-1 requirements and as implemented by DOE/RL-2000-29, the U.S. Department of Energy (DOE), Richland Operations Office has prepared this annual summary for fiscal year (FY) 2019. Originally reported in PNNL-11800 and PNNL-11800, Addendum 1, the Hanford Site Composite Analysis was approved through issuance of a 2002 memorandum. As required by DOE/RL-2000-29, an annual evaluation of new information and data developed by a number of onsite programs was completed. The reporting period for this annual evaluation is FY 2019 (October 1, 2018, through September 30, 2019). The information provided in this evaluation includes activities performed in FY 2019 that are considered pertinent to the Hanford Site Composite Analysis - information that could change source terms considered, and monitoring, research, and development results.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Status Report on the MSRE Transform Model for Thermal-Hydraulic Benchmarking

Over the last several years, there has been renewed interest in the development of molten salt reactor concepts for the next generation of nuclear power plants. The modeling tools used to design these reactors must be validated in order to accurately predict the thermal hydraulic behavior expected during operation. The majority of the practical experience with molten salt reactors comes from the Molten Salt Reactor Experiment (MSRE) that operated at Oak Ridge National Laboratory (ORNL) from 1964 until 1969. The work discussed in this paper uses the design information and experimental data from that period to create and validate a detailed model of the MSRE. Once completed, this model’s performance will be compared to other system analysis codes for thermal hydraulic benchmarking, and their ability to predict the thermal hydraulic performance of molten salt reactor systems will also be compared.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗