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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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All sourcesNASA NTRSDOE OSTI
DOE OSTI2019

Salt Heater Test (FY19)

This report discusses the fiscal year 2019 (FY19) design, implementation, and preliminary data interpretation plan for a set of borehole heater tests call the brine availability tests in salt (BATS), which is funded by the DOE Office of Nuclear Energy (DOE-NE) at the Waste Isolation Pilot Plant (WIPP). The organization of BATS is outlined in Project Plan: Salt In-Situ Heater Test. An early design of the field test is laid out in Kuhlman et al., including extensive references to previous field tests, which illustrates aspects of the present test. The previous test plan by Stauffer et al., places BATS in the context of a multi-year testing strategy, which involves tests of multiple scales and processes, possibly culminating in a drift-scale disposal demonstration.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Free-field Ground Motion Induced by Underground Explosions at Aqueduct Mesa with Predictions for Physical Experiment One (PE1)

We fit standard power-law models for ground acceleration and particle velocity using scaled ground motion observations from underground nuclear explosions at Aqueduct Mesa on the Nevada National Security Site. The models are then validated using small-scale high-explosive (HE) test data from nearby Rainier Mesa. The comparison between model predictions and observations is good and can be made more favorable if the assumed yield for the HE tests are doubled. Further validation is made with a numerical experiment that additionally shows a transition in ground motion attenuation to elastic propagation (1 = r). An extended model is provided to incorporate the long-range transition in attenuation, which results in increased ground motions at farther range. We then make predictions for the field experiment, Physical Experiment One (PE1), to take place in P-tunnel on Aqueduct Mesa, and suggest that the models could be used in planning various operations for PE1.

45 MILITARY TECHNOLOGY, WEAPONRY, AND NATIONAL DEF↗
DOE OSTI2021

Update on the Simulation of Commercial Drying of Spent Nuclear Fuel

The purpose of this report is to document improvements in the simulation of commercial vacuum drying procedures at the Nuclear Energy Work Complex at Sandia National Laboratories. Validation of the extent of water removal in a dry spent nuclear fuel storage system based on drying procedures used at nuclear power plants is needed to close existing technical gaps. Operational conditions leading to incomplete drying may have potential impacts on the fuel, cladding, and other components in the system. A general lack of data suitable for model validation of commercial nuclear canister drying processes necessitates additional, well-designed investigations of drying process efficacy and water retention. Scaled tests that incorporate relevant physics and well-controlled boundary conditions are essential to provide insight and guidance to the simulation of prototypic systems undergoing drying processes.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Response of a Pressurized Water Reactor Dashpot Region to Commercial Drying Cycles

The purpose of this report is to document updates to the simulation of commercial vacuum drying procedures at the Nuclear Energy Work Complex at Sandia National Laboratories. Validation of the extent of water removal in a dry spent nuclear fuel storage system based on drying procedures used at nuclear power plants is needed to close existing technical gaps. Operational conditions leading to incomplete drying may have potential impacts on the fuel, cladding, and other components in the system. A general lack of data suitable for model validation of commercial nuclear canister drying processes necessitates additional, well-designed investigations of drying process efficacy and water retention. Scaled tests that incorporate relevant physics and well-controlled boundary conditions are essential to provide insight and guidance to the simulation of prototypic systems undergoing drying processes. This report documents testing updates for the Dashpot Drying Apparatus (DDA), an apparatus constructed at a reduced scale with multiple Pressurized Water Reactor (PWR) fuel rod surrogates and a single guide tube dashpot. This apparatus is fashioned from a truncated 5×5 section of a prototypic 17×17 PWR fuel skeleton and includes the lowest segment of a single guide tube, often referred to as the dashpot region. The guide tube in this assembly is open and allows for insertion of a poison rod (neutron absorber) surrogate.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

DANCE and NEUANCE measurements of $^{233}U(n, \gamma$) (Q3 Report to NCSP on FY22)

The experiment was performed at the end of the CY20 runcycle and CY21. A total of 20 days were measured using the 20 mg sample, and the 10 mg sample was placed in the beam for a total of 2 days. The rest of the beam time was used to measure radioactive γ sources for energy calibration, background measurements, and tests to define the 233 U windows required during the data taking.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2024

Once-Through Steam Generator Model Analysis Using Python and Advanced Optimization Tools (Summer Internship Report)

This study focuses on the parametric analysis of design parameters for a once-through steam generator (OTSG) model, using python and advanced optimization tools to facilitate applications such as the flowing autoclave steam generator (FASG) test cases. Building on previous research involving another OTSG with a different design, this project aims to enhance our understanding of how steam generators (SGs) behave and how their outputs are influenced by changes in design. The reason for this design change is to allow for more precise modeling and optimization of SG performance, to provide a comparative analysis between the two designs, and to set up the model for integration with the FASG test case. The OTSG python-model is a mathematical representation (including fluid flow and heat transfer equations/models/correlations) of a steam-generating unit in a pressurized water reactor-type small modular reactor system. Design studies involve changing the model’s input design parameters to observe the resulting effects on the output of the system. By using advanced optimization tools, such as the Risk Analysis Virtual Environment (RAVEN) developed at Idaho National Laboratory, detailed design parametric studies and model optimization were performed. Six input parameters—pressure, temperature, and mass flow rate for the inlet of the primary-side (hot fluid) and secondary-side (cold fluid), respectively, of the SG—were randomly perturbed via RAVEN’s Monte Carlo Sampler module, using uniform distributions (i.e., ±1%, ±5% and ±10% relative changes) for 600 samples. The analysis provides valuable insights into SG optimization and can be used for sensor placement optimization to effectively monitor and obtain experimental data in other tests.

20 FOSSIL-FUELED POWER PLANTS↗
DOE OSTI2020

Hadron-argon cross-section measurements in ProtoDUNE-SP

The single-phase liquid argon ProtoDUNE detector (ProtoDUNE-SP) is located at CERN’s neutrino platform facility and serves as a prototype to validate the technology for the 10-kton fiducial mass liquid argon detectors for the DUNE experiment. ProtoDUNE-SP was exposed to a variety of test-beam particles (electrons, muons, pions, kaons, and protons) towards the end of 2018 collecting data in a broad range of momenta, from 0.3 - 7 GeV/c. The results will improve the understanding of final-state interactions in neutrino-argon cross-sections. The progress of the analyses towards the cross-section measurements on argon from the test-beam data collected at protoDUNE-SP will be presented.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
NASA NTRS1972

NASA-Lewis experiences with multigroup cross sections and shielding calculations

The nuclear reactor shield analysis procedures employed at NASA-Lewis are described. Emphasis is placed on the generation, use, and testing of multigroup cross section data. Although coupled neutron and gamma ray cross section sets are useful in two dimensional Sn transport calculations, much insight has been gained from examination of uncoupled calculations. These have led to experimental and analytic studies of areas deemed to be of first order importance to reactor shield calculations. A discussion is given of problems encountered in using multigroup cross sections in the resolved resonance energy range. The addition to ENDF files of calculated and/or measured neutron-energy-dependent capture gamma ray spectra for shielding calculations is questioned for the resonance region. Anomalies inherent in two dimensional Sn transport calculations which may overwhelm any cross section discrepancies are illustrated.

Lahti, G. P.↗
NASA NTRS1990

Current and future uses of accelerators in particle astrophysics

Beams of artificially accelerated heavy ions, protons, antiprotons, electrons, and positrons currently available at (and planned for) numerous facilities around the world are a valuable resource to the Cosmic Ray community. Such beams have been used to test detector concepts, calibrate balloon-borne and space flight experiments and to measure fundamental nuclear physics parameter necessary for the interpretation of Cosmic Ray data. As new experiments are flown the quality and extent of Cosmic Ray measurements will continue to improve. It will be necessary to increase activity at ground based accelerators in order to test/calibrate these new instruments and to maintain (or possibly improve) the ability to interpret these data. In this area, the newly formed Transport Collaboration, supported by NASA, will be providing new nuclear interaction cross section measurements for beams with Z less than or = 58 and supporting new instrument calibrations at the Lawrence Berkeley Laboratory Bevalac accelerator.

Guzik, T. G.↗
DOE OSTI2016

TREAT Multi-SERTTA Neutronics Design and Analysis Support

Transient Reactor Experiment Test (TREAT) experiment vehicle design is necessary in preparation for TREAT facility restart and the resumption of transient testing to support current Accident Tolerant Fuel (ATF) characterization needs and future fuels testing requirements. Currently the most mature vehicle design is the Multi-SERTTA (Static Environments Rodlet Transient Test Apparatuses), which can accommodate up to four concurrent rodlet-sized specimens under separate environmental conditions. Development of this robust test vehicle design requires neutronics analyses to support optimization of the Power Coupling Factor (PCF) to efficiently maximize energy generation in the fuel rodlets and to perform essential safety analyses. Current Multi-SERTTA calculations support final design of this vehicle, the design process for other TREAT vehicles, and the development of analyses practices for upcoming transient testing experiments. Models of the preliminary design Multi-SERTTA vehicle with rodlets were prepared and neutronics calculations were performed using MCNP6.1 with ENDF/B-VII.1 nuclear data libraries. Discussion of results.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Strength Oxidation of NBG-17 Graphite

This poster is to show the strength after oxidation of NBG-17 graphite at various mass loss percentages. This data is compared to previous data that had no standards of testing to establish an ASTM standard for future application.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

The MELCOR Analysis of Chinshan Nuclear Power Plant Spent Fuel Pool for Fukushima-like Accident

The safety analysis of the nuclear power plant (NPP) is very important work in the NPP safety. Especially after the Fukushima NPP accident occurred, the importance of NPP safety analysis has been raised and there is more concern for the safety of NPPs in Taiwan. Because the earthquake and tsunami occurred, the cooling system of spent fuel pool failed and the safety issue of spent fuel pool generated in Japan's Fukushima NPP. Chinshan NPP is the first NPP in Taiwan which is BWR/4 plant and OLTP (Original Licensed Thermal Power) for each unit is 1775 MWt. Chinshan NPP finished SPU (stretch power uprate) and the operating power is 103.66% of the OLTP, which is 1840 MWt now. After Fukushima NPP accident occurred, in order to concern the safety of Chinshan NPP spent fuel pool, we performed the safety analysis of spent fuel pool by using MELCOR/SNAP which also assumed the cooling system of spent fuel pool failed. The geometry of Chinshan NPP spent fuel pool is 12.17 m x 7.87 m x 11.61 m and the initial condition is 60 deg. (water temperature) / 1.013 x 10{sup 5} Pa. And, the total power of the fuels is roughly 8.9 MWt initially. MELCOR is developed by Sandia National Laboratories. MELCOR is a fully integrated, engineering-level computer code that models the progression of severe accidents in light water reactor NPPs, including containment response and leakage of activity to the environment. A broad spectrum of severe accident phenomena in both BWRs and PWRs is treated in MELCOR in a unified framework. These include thermalhydraulic response in the reactor coolant system, reactor cavity, containment, and confinement buildings; core heatup, degradation, and relocation; core-concrete attack; hydrogen production, transport, and combustion; fission product release and transport behavior. Current uses of MELCOR include estimation of severe accident source terms and their sensitivities and uncertainties in a variety of applications. According to Dr. Carbajo's paper, MELCOR was used to perform the study of spent fuel pool of Fukushima Daiichi Unit 4. Niina E. Koenoenen used MELCOR to establish the spent fuel pool model of a Nordic BWR. The loss-of-pool-cooling accidents was simulated and analyzed by this model. The above studies indicate that MELCOR is capable of handling the simulation of the spent fuel pool. An increasing number of researchers are using MELCOR code to analyze test facilities and nuclear power plants. Jun Wang et al. used MELCOR to perform core degradation simulation and MELCOR results were compared with CORA experimental data. Through this work, they reviewed the performance of MELCOR COR package in detail. Tae Woon Kim et al. developed MELCOR model of APR1400 (an advanced pressurized water reactor). They used this model to perform the LBLOCA transient simulation and sensitivity study. Longze Li et al. established the MELCOR model of CPR1000 (a Chinese pressurized reactor 1000-MW power plant). The station blackout (SBO) with failure of the steam generator (SG) safety relief valve (SRV) transient was simulated and analyzed by this model. SNAP is a graphic user interface program that processes the inputs and outputs of MELCOR. In addition, there is an animation function in SNAP which can present the animation of analysis results. There are two main steps in this research. The first step was the establishment of Chinshan NPP spent fuel pool MELCOR/SNAP model. And the transient analysis of Chinshan NPP spent fuel pool MELCOR/SNAP model under the cooling system failure condition was performed. The next step was the comparison among the MELCOR, TRACE and CFD data. TRACE and CFD data were from INER reports. In addition, the animation model of Chinshan NPP spent fuel pool was presented by using the animation function of SNAP with MELCOR/SNAP analysis results. This study has developed MELCOR/SNAP model of Chinshan NPP spent fuel pool successfully. By using the above model, the safety analysis of the spent fuel pool was performed under the cooling system of spent fuel pool failed condition. The analysis results of MELCOR, TRACE and CFD were similar in this case. It indicated that there was a respectable accuracy in MELCOR/SNAP model. The analysis results depicted that the uncovered of the fuels occurred at 2.6 day and the metal-water reaction of fuels occurred roughly at 3.7 day after the cooling system failed. The above results indicated that the failure of cladding occurred after 3.7 day. This study's results can help to evaluate the safety issue of Chinshan NPP spent fuel pool. In addition, the cooling system and some mitigation measures of spent fuel pool is not simulated in the above model. However, we will do it in the future. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Korean Localization Program for Performance Verification Test of APR1400's RCP

The Korean government together with the Korean nuclear society completed the development of technologies for the design, manufacturing, and performance verification test of the APR1400's Reactor Coolant Pump (RCP) as one of the national programs for the localization of deficient nuclear technology. The RCP localization program was driven by KAERI, which was responsible for the development of the performance verification and evaluation technology, and by DOOSAN heavy industrial company, which was in charge of the design and manufacturing of the RCP. The test facility for the localized APR1400's RCP was constructed at the KAERI site at the end of 2012. The thermal hydraulic and electric capability of the RCP test facility (RCPTF) covers up to 18.5 MPa, 343 deg. C, 11.7 m{sup 3}/s, and 14.0 MW for the design pressure, temperature, flow rate, and maximum electric power, respectively. In 2013, a commissioning test was performed to verify its designed capability, followed by several modifications in the RCPTF including signal processing and control logic to enhance the verification and evaluation capability of the RCP performance. After finishing the commissioning and modification of the RCPTF, a type test for the new-type of RCP was performed successfully. In 2014 and 2015, eight production tests for the localized RCPs, which will be installed in a commercial nuclear power plant in Korea, were successfully completed. As a final stage of the localization program of the RCP, KAERI will perform a seal LOCA test to verify the leakage characteristics and produce experimental data on the localized seal assembly for a safety analysis during a station black out (SBO) condition. In this summary, several technical issues developed through a series of research processes for a performance verification test of the localized RCP are described.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Using Information Automation and Human Technology Integration to Implement Integrated Operations for Nuclear

The purpose of the research effort described in this report is to develop and demonstrate an approach to the design and implementation of advanced, automated systems intended to increase operational efficiencies at nuclear power plants. In particular, we describe methods for considering human-technology integration (HTI) issues throughout the various phases of system design, test, and implementation and how these considerations help promote effective design. This research project will develop planning tools and comprehensive guidance on how HTI principles and methods, in combination with information automation technologies, can enable effective data integration and coordination for full nuclear plant modernization. Specifically, this research project will develop an approach to automate the mapping of data from plant systems and processes to application needs, thereby significantly reducing the amount of human workload currently required for the execution of these tasks. In addition to developing an automated solution as a replacement for these tasks, this research will also analyze digitalization’s effectiveness in reducing operational costs of compliance related activities. Compliance activities are estimated to account for as much as 50% of operations and maintenance (i.e., non-fuel and non-capital) costs of plant operation.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2018

FY2018 ILAW Glass SPFT Testing for Disposal at IDF: Glasses ORLEC46 and ORLEC52 (Final Report, Rev. 0)

About 50 million gallons of high-level mixed waste is currently stored in underground tanks at The United States Department of Energy’s (DOE’s) Hanford site in the State of Washington. The Hanford Tank Waste Treatment and Immobilization Plant (WTP) will provide DOE’s Office of River Protection (ORP) with a means of treating this waste by vitrification for subsequent disposal. The tank waste will be separated into low- and high-activity waste fractions, which will then be vitrified respectively into Immobilized Low Activity Waste (ILAW) and Immobilized High Level Waste (IHLW) products. The ILAW product will be disposed in an engineered facility – the Integrated Disposal Facility (IDF) – on the Hanford site, while the IHLW product is designed for deep geological disposal in a national facility for high level nuclear waste. The ILAW and IHLW products must meet a variety of requirements with respect to protection of the environment before they can be accepted for disposal. The objective of the work described in this report is to perform testing, data collection, and analyses for the ILAW glass product for subsequent use in the performance assessment (PA) of the IDF to assess potential environmental risks associated with long-term storage.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Searching for Flavor Dependence in Nuclear Quark Behavior

The observed correlation between the EMC effect and the contribution of short-range correlations (SRCs) in nuclei suggests that the modification of the quark distributions of bound protons and neutrons might occur within SRCs. This raises the possibility that the EMC effect may have an isospin dependence arising from the np dominance of SRCs. We discuss previous attempts to test this possibility and perform a new analysis of existing data. We find no experimental support for the observation of an isospin dependence of the EMC effect.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2016

Collision of Physics and Software in the Monte Carlo Application Toolkit (MCATK)

The topic is presented in a series of slides organized as follows: MCATK overview, development strategy, available algorithms, problem modeling (sources, geometry, data, tallies), parallelism, miscellaneous tools/features, example MCATK application, recent areas of research, and summary and future work. MCATK is a C++ component-based Monte Carlo neutron-gamma transport software library with continuous energy neutron and photon transport. Designed to build specialized applications and to provide new functionality in existing general-purpose Monte Carlo codes like MCNP, it reads ACE formatted nuclear data generated by NJOY. The motivation behind MCATK was to reduce costs. MCATK physics involves continuous energy neutron & gamma transport with multi-temperature treatment, static eigenvalue (k eff and α) algorithms, time-dependent algorithm, and fission chain algorithms. MCATK geometry includes mesh geometries and solid body geometries. MCATK provides verified, unit-test Monte Carlo components, flexibility in Monte Carlo application development, and numerous tools such as geometry and cross section plotters.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗