Search NASA⌕ Search

SEARCH · Search NASA

Results for “JAEA”

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 19 records

DOE and JAEA Field Trial of the Single Chip Shift Register (SCSR)

Los Alamos National Laboratories (LANL) has recently developed a new data acquisition system for multiplicity analysis of neutron detector pulse streams. This new technology, the Single Chip Shift Register (SCSR), places the entire data acquisition system along with the communications hardware onto a single chip. This greatly simplifies the instrument and reduces the size. The SCSR is designed to be mounted into the neutron detector head alongside the instrument amplifiers. The user’s computer connects via USB directly to the neutron detector eliminating the external data acquisition electronics entirely. JAEA, through the INSEP program, asked LANL to demonstrate the functionality of the SCSR in Tokai using the JAEA Epithermal Neutron Multiplicity Counter, ENMC. In late September of 2015 LANL traveled to Tokai to install, demonstrate and uninstall the SCSR in the ENMC. This report documents the results of that field trial.

97 MATHEMATICS AND COMPUTING↗

HTTR-GT/H2 test plant - design study on helium gas turbine for the heat utilization system connected to JAEA's HTTR - 18674

In JAEA (Japan Atomic Energy Agency), the program of verification of combined system with helium gas turbine power generation and hydrogen production using nuclear heat which becomes the first in the world, is proceeding. From 2014 to 2015 Japanese Fiscal Year, the design study of the heat utilization system which is arranged in cascade to the heat supply system to supply the hydrogen production system and helium gas turbine system which are installed in the secondary helium-cooling system of High Temperature Engineering Test Reactor (HTTR) was performed and measure technologies of the system have been verified. MHI participated in the design study work of main equipment of this system. This paper reports the result of the design study of a helium gas turbine among those results. Self-sustaining condition (a power generation output is attained) is required for the gas turbine system in three modes of operation such as a) gas turbine individual operation, b) gas turbine + hydrogen production system combined operation, and c) hydrogen production system load loss and gas turbine independent shift operation]. By adopting a multi-stage centrifugal type impeller as a gas turbine and a compressor as a result of design study, it confirmed that all the mode-of-operation self-sustaining can be attained, and can attain a maximum of approx. 1 MWe power generation output on condition of the turbine inlet temperature of 560 deg. C, the turbine flow of 9.8 kg/sec, the compressor inlet temperature of 30 deg. C, and the compression ratio 1.3. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Uncertainty propagation for the coulometric measurement of the plutonium concentration in CRM126 solution provided by JAEA

This GUM WorkbenchTM propagation of uncertainty is for the coulometric measurement of the plutonium concentration in a Pu standard material (C126) supplied as individual aliquots that were prepared by mass. The C126 solution had been prepared and as aliquoted as standard material. Samples are aliquoted into glass vials and heated to dryness for distribution as dried nitrate. The individual plutonium aliquots were not separated chemically or otherwise purified prior to measurement by coulometry in the F/H Laboratory. Hydrogen peroxide was used for valence adjustment.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Uranium Adsorbent Fibers Prepared by Atom-Transfer Radical Polymerization (ATRP) from Poly(vinyl chloride)- co -chlorinated Poly(vinyl chloride) (PVC- co -CPVC) Fiber

The need to secure future supplies of energy attracts researchers in several countries to a vast resource of nuclear energy fuel: uranium in seawater (estimated at 4.5 billion tons in seawater). In this study, we developed effective adsorbent fibers for the recovery of uranium from seawater via atom-transfer radical polymerization (ATRP) from a poly- (vinyl chloride)-co-chlorinated poly(vinyl chloride) (PVC-co-CPVC) fiber. ATRP was employed in the surface graft polymerization of acrylonitrile (AN) and tert-butyl acrylate (tBA), precursors for uranium-interacting functional groups, from PVC-co-CPVC fiber. The [tBA]/[AN] was systematically varied to identify the optimal ratio between hydrophilic groups (from tBA) and uranyl-binding ligands (from AN). The best performing adsorbent fiber, the one with the optimal [tBA]/[AN] ratio and a high degree of grafting (1390%), demonstrated uranium adsorption capacities that are significantly greater than those of the Japan Atomic Energy Agency (JAEA) reference fiber in natural seawater tests (2.42-3.24 g/kg in 42 days of seawater exposure and 5.22 g/kg in 49 days of seawater exposure, versus 1.66 g/kg in 42 days of seawater exposure and 1.71 g/kg in 49 days of seawater exposure for JAEA). Adsorption of other metal ions from seawater and their corresponding kinetics were also studied. The grafting of alternative monomers for the recovery of uranium from seawater is now under development by this versatile technique of ATRP.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Sodium Fire Collaborative Study Progress CNWG Fiscal Year 2021

This report discusses the progress on the collaboration between Sandia National Laboratories (Sandia) and Japan Atomic Energy Agency (JAEA) on the sodium fire research in fiscal year (FY) 2021 and is a continuation of the FY2020 progress report. In this report, we only report the changes made to the current sodium pool fire model in MELCOR. We modified and corrected many control functions to enhance the current model from the FY2020 report. This year’s enhancements relate to better agreement of the suspended aerosol measurement from JAEA’s F7 series tests. Staff from Sandia and JAEA conducted the validation studies of the sodium pool fire model in MELCOR. To validate this pool fire model with the latest enhancement, JAEA sodium pool fire experiments (F7-1 and F7-2) were used. The results of the calculation, including the code-to-code comparisons are discussed as well as suggestions for further model improvement. Finally, recommendations are made for new MELCOR simulations for FY2022.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Sodium Fire Collaborative Study Progress (CNWG Fiscal Year 2022)

This report discusses the progress on the collaboration between Sandia National Laboratories (Sandia) and Japan Atomic Energy Agency (JAEA) on the sodium fire research in fiscal year (FY) 2022 and is a continuation of the FY 2021 progress report. We only report the changes made to the current sodium pool fire model in MELCOR. We modified and corrected many control functions to enhance the fraction of oxygen consumed that reacts to form monoxide (FO2) parameter in the current model from the FY2021 report. This year's enhancements relate to better agreement of the suspended aerosol measurement from JAEA's F7 series tests. Staff from Sandia and JAEA conducted the validation studies of the sodium pool fire model in MELCOR. To validate this pool fire model with the latest enhancement, JAEA sodium pool fire experiments (F7-1 and F7-2) were used. The results of the calculation, including the code-to-code comparisons are discussed as well as suggestions for further model improvement. Finally, recommendations are made for new MELCOR simulations for FY 2023.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DECOVALEX-2023: Task D Final Report

Task D of DECOVALEX-2023 is focused on the simulation of the coupled thermal hydraulic-mechanical (THM) behaviour in the full-scale engineered barrier system (EBS). The Horonobe EBS experiment is the demonstration of the full-scale EBS in the underground research laboratory (URL) (performed by JAEA in the Horonobe URL in Japan). Task D consisted of the three steps, a preliminary step (Step 0), simulation of the laboratory tests (Step 1) and simulation of the in-situ full-scale EBS experiment (Step 2). Since the Horonobe EBS experiment demonstrates the vertical emplacement option of the EBS, the experiment gallery is also backfilled with the backfill material. Therefore, interaction between the EBS and the backfill material can also be demonstrated, such as deformation (change of density) of the buffer material. The underground water in the Horonobe URL is saline. This fact adds chemical processes to THM behaviour. For example, mechanical properties (such as swelling pressure of the buffer material and backfill material) and hydraulic properties (such as permeability of the buffer material and backfill material) change depending on the water chemistry. Task D was therefore a challenging Task focused on not only the relatively simple THM behaviour but also complex THM behaviour including chemical processes. Six research teams (BGR, CAS, JAEA, KAERI, SNL and Taipower) participated the Task D. BGR, CAS, JAEA, KAERI and Taipower research teams selected a THM approach, while the SNL research team selected a TH approach. Step 1 involved the simulation of laboratory test results and was important to check the numerical codes developed by the research teams. Step 1 was divided into four sub steps. The simulation results through the Step 1 identified the parameters for simulation of the Step 2. Basic parameters of the materials (buffer material, backfill material, rock mass, concrete, sand) were provided by JAEA. Special parameters which research team needed were identified by back analysis of Step 1. Most notably the mechanical behaviour of swelling and displacement depended on the applied model (elastic model or elastoplastic model). Parameters such as Young’s modulus were found to need smaller values than characterised in the fundamental laboratory test results (Step 1-1, 1-2) for the elastic model. Although laboratory experiments are usually simple, test results contained some error. For example, if the saturation level is 100 % or higher, it should be considered an error. This situation was presented in the Step 1-3. A possible reason is that the buffer material is a mixture of bentonite and silica sand. When a specimen is cut to measure volume or weight, sand grains will affect the measurement data. In Step 2, boundary conditions such as temperature on the surface of the simulated overpack, heater power of the electrical heaters installed in the simulated overpack, injection pressure and inflow rate of the test water, were applied. The outer boundary conditions can be selected using measured data (injection pressure and inflow rate of the test water that is controlled by the injection systems installed in the sand layer around the buffer material and in the boundary between backfill material and concrete support). Since such measured data has some noise, research teams developed their own simplified boundary conditions. Inner boundary conditions can be selected using measured data as heater power and temperature on the surface of the simulated overpack. These data also contain some noise, so research teams developed their own simplified developed boundary conditions. Task D validated various approaches thorough the simulation of the in-situ full scale EBS system including backfill of the gallery: variations in the coupling processes (THM or THC), analysis codes, and boundary conditions. Temperature distribution in the buffer material was simulated well by all research teams. This means thermal behaviour is not sensitive to the simulation approaches. Although the water content distribution on the outside of the buffer material was well simulated by all research teams, the simulation results differ from the measured values inside the buffer material (at the centre and inside, near the simulated overpack). The buffer material is made from tap water, but in the in-situ experiment, saline groundwater infiltrates the buffer material. Therefore, the selection of the hydraulic parameters of the buffer material greatly affects the simulation results of the re saturation behaviour of the buffer material. In the Horonobe EBS experiment, measured values suitable for validating the simulation results were not obtained near the simulated overpack. When simulating the pressure and deformation of the buffer material, the measurement data is easily affected by the installation conditions of the measurement sensors, so verifying the measurement data itself remains an issue. Mechanical simulation results differ depending on whether they are considered as elastic or elastoplastic phenomena. The accuracy of measured in-situ data can be assessed by detailed analysis comparing sampling specimen analysis and measured data. The Horonobe EBS experiment is scheduled to be dismantled in the future (FY2026 and 2027). This detailed dismantling investigation will finally confirm the measured data.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Comparative study on neutronics characteristics of a 1500 MWe metal fuel sodium-cooled fast reactor

Under the cooperative effort of the Civil Nuclear Energy R&D Working Group within the framework of the U.S.-Japan bilateral, Argonne National Laboratory (ANL) and Japan Atomic Energy Agency (JAEA) have been performing benchmark study using Japan Sodium-cooled Fast Reactor (JSFR) design with metal fuel. In this benchmark study, core characteristic parameters at the beginning of cycle were evaluated by the best estimate deterministic and stochastic methodologies of ANL and JAEA. The results obtained by both institutions show a good agreement with less than 200 pcm of discrepancy on the neutron multiplication factor, and less than 3% of discrepancy on the sodium void reactivity, Doppler reactivity, and control rod worth. The results by the stochastic and deterministic approaches were compared in each party to investigate impacts of the deterministic approximation and to understand potential variations in the results due to different calculation methodologies employed. From the detailed analysis of methodologies, it was found that the good agreement in multiplication factor from the deterministic calculations comes from the cancellation of the differences on the methodology (0.4%) and nuclear data (0.6%). The different treatment in reflector cross section generation was estimated as the major cause of the discrepancy between the multiplication factors by the JAEA and ANL deterministic methodologies. Impacts of the nuclear data libraries were also investigated using a sensitivity analysis methodology. Furthermore, the differences on the inelastic scattering cross sections of U-238, ν values and fission cross sections of Pu-239 and µ-average of Na-23 are the major contributors to the difference on the multiplication factors.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Sodium Fire Collaborative Study Progress (CNWG FY 2019)

This report discusses the progress on the collaboration between Sandia National Laboratories (Sandia) and Japan Atomic Energy Agency (JAEA) on the sodium fire research in fiscal year 2019. First, the current sodium pool fire model in MELCOR, which is adapted from CONTAIN-LMR code, is discussed. The associated sodium fire input requirements are also presented. A proposed model improvement developed at Sandia is discussed. Finally, the validation study of the sodium pool fire model in MELCOR carried out by a JAEA's staff is described. To validate this model, a JAEA sodium pool fire experiment (F7-1 test) is used. A preliminary calculation is performed using a modified MELCOR model from a previous experiment simulation. The results of the calculation are discussed as well as suggestions for improvement. Finally, recommendations are made for new MELCOR simulations for next fiscal year, 2020.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

Sodium Fire Collaborative Study Progress -- CNWG Fiscal Year 2020

This report discusses the progress on the collaboration between Sandia National Laboratories (Sandia) and Japan Atomic Energy Agency (JAEA) on the sodium fire research in fiscal year 2020. First, the current sodium pool fire model in MELCOR, which is adapted from CONTAIN-LMR code, is discussed. The associated sodium fire input requirements are also presented. These input requirements are flexible enough to permit further model development via control functions to enhance the current model without modifying the source code. The theoretical pool fire model improvement developed at Sandia is discussed. A control function model has been developed from this improvement. Then, the validation study of the sodium pool fire model in MELCOR carried out by both Sandia and JAEA’s staff is described. To validate this pool fire model with the enhancement, a JAEA sodium pool fire experiment (F7-1 test) is used. The results of the calculation are discussed as well as suggestions for further model improvement. Finally, recommendations are made for new MELCOR simulations for next fiscal year, 2021.

36 MATERIALS SCIENCE↗