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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 271 records · Page 15
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

Power balance analysis at the L-H transition in JET-ILW NBI-heated deuterium plasmas

The understanding of the physics underlying the L-H transition has strong implications for ITER experimental reactor and demonstration power plant (DEMO). In many tokamaks, including JET, it has been observed that, at a particular plasma density, n e,min , the power necessary to access H-mode P L-H is minimum. In the present work, L-H transitions of JET deuterium plasmas heated by neutral beam injection (NBI) are studied for the first time by means of a power balance analysis to characterize the main contributions in the transition, through integrated transport modelling. In the pulses analysed, we do observe a minimum of the L-H power threshold in density, indicating the presence of density branches and of n e,min . Electron and ion heat fluxes at the transition are estimated separately. The electron/ion equipartition power results in favour of the ions, as shown by QuaLiKiz quasilinear gyrokinetic simulations, which predict a larger ion transport that causes T e > T i . The resulting edge ion heat flux also shows a clear change of slope below n e,min , similarly to ASDEX-Upgrade (AUG) NBI pulses (Ryter et al 2014 Nucl. Fusion 54 083003). JET NBI data are compared to radio-frequency heated AUG and Alcator C-mod pulses (Schmidtmayr et al 2018 Nucl. Fusion 58 056003), showing a different trend of the power, coupled to ions at the L-H transition with respect to the linearity observed in the radio-frequency heated plasmas. Finally, the presence of n e,min and the role of the ion heat flux is discussed in the paper, although it seems it is not possible to explain the presence of a P L-H minimum in density by a critical ion heat flux and by the equipartition power for the JET NBI-heated plasmas analysed.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2020

Simulation of the Fast Reactor Fuel Assembly Duct-Bowing Reactivity Effect Using Monte Carlo Neutron Transport and Finite Element Analysis

This paper discusses a new method of simulating the fuel assembly duct-bowing reactivity coefficient for EBR-II run 138B. Quantification of the fuel assembly duct-bowing reactivity effect in liquid metal–cooled fast reactors has been a persistent problem since they were first designed and operated. Simulation of the duct-bowing reactivity effect is difficult because the level of detail required to simulate the effect has exceeded most modeling capabilities. The new method outlined in this paper utilizes the finite element analysis code ANSYS to analyze the thermal and structural components. Here, the displacement of the fuel assembly duct due to thermal expansion and mechanical interaction was calculated by ANSYS using recorded EBR-II run 138B temperature and power boundary value data. The displacement values were incorporated into to a Monte Carlo model of EBR-II run 138B and keff was calculated. Multiple Monte Carlo calculations were performed with duct displacement values corresponding to different reactor temperatures. Using the calculated keff values associated with the different duct displacement results allowed calculation of the duct-bowing reactivity coefficient. The duct-bowing reactivity coefficient was calculated to be –14.5 × 10 –4 $/°C/ ± 4.4%.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Benchmark specifications for EBR-II shutdown heat removal tests

Argonne National Laboratory (ANL) is hosting an IAEA-coordinated research project on benchmark analyses of sodium-cooled fast reactor passive safety tests performed at the Experimental Breeder Reactor-II (EBR-II). The benchmark project involves analysis of a protected and an unprotected loss of flow tests conducted during an extensive testing program within the framework of the U.S. Integral Fast Reactor program to demonstrate the inherently safety features of EBR-II as a pool-type, sodium-cooled fast reactor prototype. The project is intended to improve the participants' design and safety analysis capabilities for sodium-cooled fast reactors through validation and qualification of safety analysis codes and methods. This paper provides a description of the EBR-II tests included in the program, and outlines the benchmark specifications being prepared to support the IAEA-coordinated research project. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Decommissioning of a facility in a historical reactor at SRS: achieving both historical significance preservation and effective in-situ decommissioning - 15275

The Savannah River Site (SRS) is an 802 square-kilometer United States Department of Energy (US DOE) nuclear facility located along the Savannah River near Aiken, South Carolina, where Management and Operations are performed by Savannah River Nuclear Solutions, LLC (SRNS). In 2004, US DOE recognized SRS as a structure within the Cold War Historic District of national, state and local significance because it is composed of the first generation of facilities constructed and operated from 1950 through 1989 to produce plutonium and tritium for our nation's defense. US DOE agreed to manage the SRS 105-C Reactor Facility as a potentially historic property due to its significance in supporting the U.S. Cold War Mission and for potential for future interpretation. The Disassembly Basin (DB) in C Reactor was evaluated for action under CERCLA. Data indicated there were over 830 curies of radioactivity associated with the basin sediments and approximately 9.1 million (M) liters (L) of contaminated water, not including a large quantity of activated reactor equipment, scrap metal, and debris on the basin floor. The release of DB water could potentially migrate to the aquifer and contaminate groundwater. The need for an action was identified to reduce risks to personnel in the facility and to eliminate the possible release of contaminants into the environment. In-situ decommissioning was selected as the remedy. The remedial action proceeded in stages: -Above-grade modifications to the interior of DB building -Relocation of irradiated scrap and reactor components to the deepest segments of the DB - Construction of a metal building to house evaporators - Forced evaporation of the DB's radiologically contaminated water - Grouting with special hybrid grouts - Removal of the evaporators and temporary building. Since the building itself was not demolished and will remain unchanged in appearance with the exception of filling the DB with grout, this ISD removal action/methodology did not adversely impact the historical significance of the 105-C Reactor Building, in accordance with the National Historic Preservation Act, nor did it adversely affect the ability to determine the full extent and details of preservation at some later date. Both special hybrid grout materials met all requirements. Overall average 7-day compressive strength for the dry area mix was 1.54 MPa (223 psi) and for the underwater mix 2.94 MPa (427 psi).Overall 28-day compressive strength for the dry area mix was 2.88 MPa (417 psi) and for the underwater mix 4.99 MPa (724 psi). A limited amount of testing was also performed at 91 days, where the overall compressive strength for the dry area mix was found to be 9.40 MPa (1363 psi) and for the underwater mix 13.98 MPa (2028 psi). All compressive strength values far exceeded the 0.34 MPa (50 psi) minimum requirements. The special, flowable-fill grouts designed and used for P- and R-Reactor ISD and used again for this 105-C ISD removal action resulted in considerable labor, cost, and schedule savings versus conventional materials. As an added bonus/benefit, the CO{sub 2} footprint of the SRS C-Reactor DB ISD was minimized by using a small amount of cement and byproduct material to produce the structural fill materials. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2014

Effects of a gradually increased load of fish waste silage in co-digestion with cow manure on methane production

Highlights: • New results from continuous anaerobic co-digestion of fish waste silage (FWS) and cow manure (CM). • Co-digestion of FWS and CM has a high biogas potential. • Optimal mixing ratio of FWS/CM is 13–16/87–84 volume%. • High input of FWS leads to accumulation of NH4+ and VFAs and process failure. - Abstract: This study examined the effects of an increased load of nitrogen-rich organic material on anaerobic digestion and methane production. Co-digestion of fish waste silage (FWS) and cow manure (CM) was studied in two parallel laboratory-scale (8 L effective volume) semi-continuous stirred tank reactors (designated R1 and R2). A reactor fed with CM only (R0) was used as control. The reactors were operated in the mesophilic range (37 °C) with a hydraulic retention time of 30 days, and the entire experiment lasted for 450 days. The rate of organic loading was raised by increasing the content of FWS in the feed stock. During the experiment, the amount (volume%) of FWS was increased stepwise in the following order: 3% – 6% – 13% – 16%, and 19%. Measurements of methane production, and analysis of volatile fatty acids, ammonium and pH in the effluents were carried out. The highest methane production from co-digestion of FWS and CM was 0.400 L CH4 gVS{sup −1}, obtained during the period with loading of 16% FWS in R2. Compared to anaerobic digestion of CM only, the methane production was increased by 100% at most, when FWS was added to the feed stock. The biogas processes failed in R1 and R2 during the periods, with loadings of 16% and 19% FWS, respectively. In both reactors, the biogas processes failed due to overloading and accumulation of ammonia and volatile fatty acids.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2012

Godiva IV and Juliet Diagnostics CED-1, Rev. 1 (IER-176)

The Juliet experiment is currently in preliminary design (IER-128). This experiment will utilize a suite of diagnostics to measure the physical state of the device (temperature, surface motion, stress, etc.) and the total and time rate of change of neutron and gamma fluxes. A variety of potential diagnostics has been proposed in this CED-1 report. Based on schedule and funding, a subset of diagnostics will be selected for testing using the Godiva IV pulsed reactor as a source of neutrons and gammas. The diagnostics development and testing will occur over a two year period (FY12-13) culminating in a final set of diagnostics to be fielded for he Juliet experiment currently proposed for execution in FY15.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

National Reactor Innovation Center (NRIC) Infrastructure Integration Plan

The NRIC Infrastructure Integration Plan serves as a roadmap for integrating future test reactors at the Idaho National Laboratory’s (INL’s) desert Site into core infrastructure and services required for reactor operations. This document is not intended to serve as a requirements document, nor does it list all infrastructure requirements for each reactor planned for testing at the Site. It represents an overview of the most likely infrastructure and services that will be required by some or all the test reactors, as well as what infrastructure and services exist at the Site that have the potential to support and enable NRIC test reactors.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

ECAR-7210 Steady-state Thermo-mechanical Analysis of the MARVEL Fuel Bundle

This document reports the steady-state thermal-hydraulic and static mechanical analyses of the MARVEL micro-reactor core. The objective of this work is to investigate whether fuel-to-fuel contact could occur due to the thermal deformation of the fuel cladding during the normal operation and to assess the resulting impact on the peak cladding temperatures (PCT) if contact occurs.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Knowledge Preservation at the Fast Flux Test Facility

The Fast Flux Test Facility (FFTF) is the most recent Liquid Metal Reactor (LMR) to operate in the United States, from 1982 to 1992. The technologies employed in designing and constructing this reactor, along with information obtained from tests conducted during its operation, are currently being secured and archived by the Department of Energy's Office of Nuclear Energy. This report is one in a series documenting the overall project efforts to retrieve and preserve critical information related to advanced reactors

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2011

Knowledge Preservation at the Fast Flux Test Facility

The Fast Flux Test Facility (FFTF) is the most recent Liquid Metal Reactor (LMR) to operate in the United States, from 1982 to 1992. The technologies employed in designing and constructing this reactor, along with information obtained from tests conducted during its operation, are currently being secured and archived by the Department of Energy's Office of Nuclear Energy. This report provides a status update documenting the overall project efforts to retrieve and preserve critical information related to advanced reactors.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

ECAR-6601 MARVEL Project Seismic Accelerations

The MARVEL reactor design requirements are currently listed within TFR-2576. Several MARVEL subsystems are required to be designed to seismic criteria of IBC 2015. This Engineering Analysis (EA) report documents the accelerations that need to be used in an equivalent-static method seismic analysis to satisfy this requirement.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

ECAR-6601 Rev 1 MARVEL Project Seismic Accelerations

The MARVEL reactor design requirements are currently listed within TFR-2576. Several MARVEL subsystems are required to be designed to seismic criteria of IBC 2015. This Engineering Analysis (EA) report documents the accelerations that need to be used in an equivalent-static method seismic analysis to satisfy this requirement.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗