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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 1,027 records · Page 57
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

Numerical Simulation of Twin Jets using LES with the Nek5000 Code

The twin jet activity, a joint effort between Texas A and M University, University of Tennessee - Knoxville, and Argonne National Laboratory, is being used to analyze the performance of various computational fluid dynamics (CFD) code packages. The twin jet water facility (TJWF) is based off a semi-classical problem of combining planar turbulent jets in a large free-shear environment or enclosure. The problem is similar to physics seen in advanced reactor designs such as in the large upper plena of a sodium fast reactor. Utilizing the experimental data sets created using laser Doppler velocimetry (LDV) and particle image velocimetry (PIV), the Nek5000 code, among others, is qualitatively compared to the experimental work. This effort is a starting point for determining the appropriate simulation techniques needed for resolving the twin jet physics.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

Development of the multigroup cross section library for the CASL neutronics simulator MPACT: Method and procedure

The MPACT neutronics module of the Consortium for Advanced Simulation of Light Water Reactors (CASL) core simulator is a 3D whole core transport code being developed for the CASL toolset known as the Virtual Environment for Reactor Analysis (VERA). Key characteristics of the MPACT code include a subgroup method for resonance self-shielding and a whole-core transport solver with a 2D/1D synthesis method. The AMPX and SCALE code packages have been significantly improved to include intermediate resonance parameters, as well as new modules to enhance the accuracy of resonance data and scattering matrices for the Bondarenko approach and transport cross sections of hydrogen. In addition, CASL XSTools have been developed to generate the MPACT multigroup cross section library. A simple super-homogenization method for 238U has been developed to resolve reaction rate discrepancy issues caused by angle-dependent total cross sections, poor scattering matrices, and a poor resonance interference model at coarse group structure. Here, due to the limitation of computing capacity that can occur even in high performance computing, a new 51-group structure has been developed for efficient simulation that is applicable for both pressurized water reactor (PWR) and boiling water reactor (BWR) simulations. A new ENDF/B-VII.1 MPACT 51-group library has been developed using the AMPX/SCALE and XSTools code packages, and it has been verified and validated by performing various benchmark calculations.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Measurement of prompt neutron generation time at the VIR-2M pulsed nuclear reactor

The prompt neutron generation time is measured in the core of the VIR-2M research nuclear reactor. The measurements are performed using the Babala method while the reactor is in the subcritical state. The VIR-2M reactor and the relevant experimental equipment are briefly described, and the experimental procedure and data processing technique are presented. It is shown that the prompt neutron generation time with empty experimental channels is 35 ± 1 μs.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

Material Interactions in Severe Accidents – Benchmarking the MELCOR V2.2 Eutectics Model for a BWR-3 MARK-I Station Blackout: Part I – Single Case Analysis

Here in this analysis, the two material interaction models available in the MELCOR code are benchmarked for a severe accident at a BWR under representative Fukushima Daiichi boundary conditions. This part of the benchmark investigates the impact of each material interaction model on accident progression through a detailed single case analysis. It is found that the eutectics model simulation exhibits more rapid accident progression for the duration of the accident. The slower accident progression exhibited by the interactive materials model simulation, however, allows for a greater degree of core material oxidation and hydrogen generation to occur, as well as elevated core temperatures during the ex-vessel accident phase. The eutectics model simulation exhibits more significant degradation of core components during the late in-vessel accident phase – more debris forms and relocates to the lower plenum before lower head failure. The larger debris bed observed in the eutectics model simulation also reaches higher temperatures, presenting a more significant thermal challenge to the lower head until its failure. At the end of the simulated accident scenario, however, core damage is comparable between both simulations due to significant core degradation that occurs during the ex-vessel phase in the interactive materials model simulation. A key difference between the two models’ performance is the maximum temperatures that can be reached in the core and therefore the maximum ΔT between any two components. When implementing the interactive materials model, users have the option to modify the liquefaction temperature of the ZrO 2 -interactive and UO 2 -interactive materials as a way to mimic early fuel rod failure due to material interactions. Through modification of the liquefaction of high melting point materials with significant mass, users may inadvertently limit maximum core temperatures for fuel, cladding, and debris components.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2013

Confinement improvement by fluctuating input power

An interesting effect of the fluctuating input power on plasma confinement is reported. Specifically, the addition of the sinusoidal perturbation, Asin(wt), to constant power q{sub 0} is shown to promote the confinement, leading to the L-H transition at a lower value of q{sub 0}, as compared to the case of constant q{sub 0} without the sinusoidal perturbation. In general, higher amplitude (A) and lower frequency (w) are found to be more favorable for the L-H transition while an interesting linear relation between A and w leading to the L-H transition is established for different values of q{sub 0}.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2014

Progress and present status of ITER cryoline system

The cryoline system at ITER forms a very complex network localized inside the Tokamak building, on a dedicated plant bridge and in cryoplant areas. The cooling power produced in the cryoplant is distributed via these lines with a total length of about 3.7 km and interconnecting all the cold boxes of the cryogenic system as well as the cold boxes of various clients (magnets, cryopumps and thermal shield). Distinct layouts and polygonal geometry, nuclear safety and confinement requirements, difficult installation and in-service inspection/repair demand very high reliability and availability for the cryolines. The finalization of the building-embedded plates for supporting the lines, before the detailed design, has made this project technologically more challenging. The conceptual design phase has been completed and procurement arrangements have been signed with India, responsible for providing the system of cryolines and warm lines to ITER, as in kind contribution. The prototype test for the design and performance validation has been planned on a representative cryoline section. After describing the basic features and general layout of the ITER cryolines, the paper presents key design requirements, conceptual design approach, progress and status of the cryolines project as well as challenges to build such a complex cryoline system.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

Simulation of site and topographic effects on ground motion in Los Alamos, NM mesas

Mesas, which consist of an elevated area of land with a flat top and steep cliffs at the sides, are one of the common geological formations present in the Los Alamos region. Previous research has shown that geological formations such as sedimentary canyons can amplify soil response during earthquakes. There have also been parametric studies to understand the response of an idealized and isolated mountain/canyon under inclined plane waves. In this study, a 2-D linear anelastic soil domain, with topography modelled after the Los Alamos region with many mesas and canyons, is considered to understand site-specific topographic effects in the presence of non-isolated topographical features. Various earthquake sources configurations ranging from point sources to finite faults with varying rupture length, dip angles and spatial slip distributions are considered. To isolate the effect of topography, three other soil domains—a homogeneous half-space, homogeneous half-space with mesas and canyons, and a computational domain with just the soil properties from the Los Alamos region on a flat soil domain is also considered. 2-D site-response analyses of these soil domains under earthquake excitation show that the free-field response of the soil can be amplified or de-amplified depending on the topography of the region and the location of the station relative to the fault. These studies also show that even relatively small mesas with height less than 100 m can significantly amplify the response (by a factor of 2 or more), which differ from the much smaller amplification factors (≤1.2) specified by standard building codes such as Eurocode-8 (2000) for topographical features with a similar slope. De-amplifications are also fairly common, especially close to the canyons. The results from this study agree qualitatively with those from the ambient vibration study conducted by Stolte et al. on a mesa from the Los Alamos region. Furthermore, such site-specific studies provide important insights into the variability of the topographic amplification factors within a region of interest. Such knowledge is important in the design of safety-related critical infrastructure located within that region.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2015

On-line measurement of gaseous iodine species during a PWR severe accident

A long-range remote sensing of severe accidents in nuclear power plants can be obtained by monitoring the online emission of volatile fission products such as xenon, krypton, caesium and iodine. The nuclear accident in Fukushima was ranked at level 7 of the International Nuclear Event Scale by the NISA (Nuclear and Industrial Safety Agency) according to the importance of the radionuclide release and the off-site impact. Among volatile fission products, iodine species are of high concern, since they can be released under aerosols as well as gaseous forms. Four years after the Fukushima accident, the aerosol/gaseous partition is still not clear. Since the iodine gaseous forms are less efficiently trapped by the Filtered Containment Venting Systems than aerosol forms, it is of crucial importance to monitor them on-line during a nuclear accident, in order to improve the source term assessment in such a situation. Therefore, we propose to detect and quantify these iodine gaseous forms by the use of highly sensitive optical methods. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Constellation's best estimate alternate source term methodology overview

Safety analyses for a nuclear power plants need to consider postulated accidents that results in at risk of accidental release of radiation. The regulations require plant specific safety analysis reports to include an evaluation of the requirements of 10CFR50.67. Such a safety analysis report mandates limits such that calculated radiological consequences relative to certain dose locations do not exceed a total effective dose equivalent (TEDE) limits following a postulated release of radioactivity. Calculations are performed to estimate the radiological consequences, in terms of dose, to people and equipment to ensure the estimated doses are within the prescribed limits. Analysis should demonstrate, with reasonable assurance, that these prescribed limits are complied with. Conventional methodologies utilize conservative approaches to address lack of uncertainty quantification in the utilized approaches, methods, and/or inputs. These built-in excess conservatisms often result in compounding effects and hence overly conservative results in the estimated radiological consequences, which leads to inaccurate margin evaluation for operation and accident mitigation. Therefore, evaluating accurate dose consequences is needed for both operational and safety reasons. The research documented in this paper provides an overview of Constellation's Best Estimate Alternate Source Term (BEAST) Methodology. BEAST methodology relies upon the use of realistic yet bounding input distributions for key analysis parameters, which replaces use of conservative deterministic singular inputs that bound overall analysis domain. This approach enables evaluating a more accurate accident analysis response, while enabling a bounding licensing basis envelope via more accurate quantification of uncertainty in the application. Therefore, built-in margin for a given scenario is more accurately evaluated.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

Embedded Sensors and Controls to Improve Component Performance and Reliability Conceptual Design Report

The objective of this project is to demonstrate improved reliability and increased performance made possible by deeply embedding instrumentation and controls (I&C) in nuclear power plant (NPP) components and systems. The project is employing a highly instrumented canned rotor, magnetic bearing, fluoride salt pump as its I&C technology demonstration platform. I&C is intimately part of the basic millisecond-by-millisecond functioning of the system; treating I&C as an integral part of the system design is innovative and will allow significant improvement in capabilities and performance. As systems become more complex and greater performance is required, traditional I&C design techniques become inadequate and more advanced I&C needs to be applied. New I&C techniques enable optimal and reliable performance and tolerance of noise and uncertainties in the system rather than merely monitoring quasistable performance. Traditionally, I&C has been incorporated in NPP components after the design is nearly complete; adequate performance was obtained through over-design. By incorporating I&C at the beginning of the design phase, the control system can provide superior performance and reliability and enable designs that are otherwise impossible. This report describes the progress and status of the project and provides a conceptual design overview for the platform to demonstrate the performance and reliability improvements enabled by advanced embedded I&C.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

Flowsheet Evaluation for the Dissolving and Neutralization of Sodium Reactor Experiment Used Nuclear Fuel

This report includes the literature review, hydrogen off-gas calculations, and hydrogen generation tests to determine that H-Canyon can safely dissolve the Sodium Reactor Experiment (SRE; thorium fuel), Ford Nuclear Reactor (FNR; aluminum alloy fuel), and Denmark Reactor (DR-3; silicide fuel, aluminum alloy fuel, and aluminum oxide fuel) assemblies in the L-Bundles with respect to the hydrogen levels in the projected peak off-gas rates. This is provided that the number of L-Bundles charged to the dissolver is controlled. Examination of SRE dissolution for potential issues has aided in predicting the optimal batching scenario. The calculations detailed in this report demonstrate that the FNR, SRE, and DR-3 used nuclear fuel (UNF) are bounded by MURR UNF and may be charged using the controls outlined for MURR dissolution in a prior report.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Flowsheet Evaluation for the Dissolving and Neutralization of Sodium Reactor Experiment Used Nuclear Fuel

This report includes the literature review, hydrogen off-gas calculations, and hydrogen generation tests to determine that H-Canyon can safely dissolve the Sodium Reactor Experiment (SRE; thorium fuel), Ford Nuclear Reactor (FNR; aluminum alloy fuel), and Denmark Reactor (DR-3; silicide fuel, aluminum alloy fuel, and aluminum oxide fuel) assemblies in the L-Bundles with respect to the hydrogen levels in the projected peak off-gas rates. This is provided that the number of L-Bundles charged to the dissolver is controlled. Examination of SRE dissolution for potential issues has aided in predicting the optimal batching scenario.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Tritium Management Loop Design Status

This report summarizes physical, chemical, and engineering analyses that have been performed to support development of a test loop to study tritium migration in FLiBe (2LiF-BeF 2 ) salts. The loop will operate under turbulent flow, and a schematic of the apparatus has been used to develop a model in Mathcad to suggest flow parameters that should be targeted in loop operation. The introduction of tritium into the loop has been discussed, as well as various means to capture or divert the tritium from egress through a test assembly. Permeation was calculated, starting with the development of a Modelica model for a transport through a nickel window into a vacuum, followed by modification of the model for a FLiBe system with an argon sweep gas on the downstream side of the permeation interface. Results suggest that tritium removal with a simple tubular permeation device will occur readily. Although this system is idealized, it suggests that rapid measurement capability in the loop may be necessary to study and understand tritium removal from the system.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

Progress Report on Performance of A709 and G91 Steels in Sodium

Specimens in six different processing and heat treatment conditions of A709 H58776 were exposed to sodium at 550, 600, and 650°C, respectively for various exposure times. G91 base metal was tested in sodium at 550, 600, and 650°C, and G91 weldment at 550 and 600°C, respectively. Parallel thermal aging experiments were conducted on G91 and A709 to obtain thermal aging data for comparison with sodium exposure results to separate the thermal and sodium effects. The corrosion data obtained on A709 steel continue to show low corrosion rates, which indicate good compatibility of A709 with sodium when oxygen content is controlled. G91 has also shown acceptable corrosion rates over the temperature range of 550-650°C in sodium. Thermal aging or sodium exposures reduced the tensile strength, uniform elongation, and total elongation for A709 ESR, AOD specimens. These specimens consistently showed higher yield stress and tensile strength and lower uniform and total elongations after sodium exposure than after thermal aging under comparable testing conditions. It implies that there was an additional effect of sodium exposure that gave rise to an increase in tensile strength of A709. Dynamic strain aginginduced flow serrations were completely removed by either thermal aging or sodium exposure at 650°C, which imply that carbon or nitrogen in solution was removed from the solution and formed precipitates during aging or sodium exposures. The tensile data ruled out the possibility of decarburization under the sodium exposure conditions. The HOMO specimens behaved somewhat differently from the ESR and AOD specimens in sodium. While thermal aging and sodium exposures at 550 and 600°C decreased the yield stress and the ultimate tensile strength of G91, there was virtually no additional effect resulting from sodium exposures at these two temperatures. In contrast, sodium exposures at 650°C had a drastic effect on the yield stress and the ultimate tensile strength of G91. It is suggested that G91 experienced decarburization in the 650°C sodium environments. Carbon concentrations in sodium in the SMT-1 and SMT-2 loops were determined by a foil equilibration method. The estimated carbon concentration was in the range of 0.8-1.2 ppm in the SMT-1 loop and 0.3-0.7 ppm in the SMT-2 loop. The carbon activity in sodium in the SMT-2 loop was estimated to be 0.03-0.08 at 600°C, and 0.08-0.2 at 550°C. The carbon activity in sodium at 650°C in the SMT-1 loop was 0.04-0.07. Equilibrium simulation of the carburization – decarburization processes was conducted for A709 and G91 steels exposed in sodium environments at temperatures of 550-700°C. The carbon activity-concentration relationship for G91 was re-evaluated by considering four phases in G91, i.e. bcc ferrite, M 23 C 6 , NbC and VC carbides. It was found that the carburization-decarburization process in G91 steel was dictated by M 23 C 6 carbides at high carbon activities, while NbC carbides dominated the process at low carbon activities. Formation of NbC increases the decarburization resistance of G91 steel. It remains to be understood whether the decarburization resistance provided by MC carbides in G91 can be maintained during long-term operations of SFRs. The carbon concentration-activity relationship for A709 was calculated based on the equilibrium of fcc-austenite and M 23 C 6 carbide phase. The calculations showed that A709 decarburizes at 650°C in the SMT-1 loop environments, which is different from the experimental findings. The carbon concentration-activity relationship in A709 was further evaluated by considering four phases, including fcc-austenite, M 23 C 6 , NbC, and TiC carbide phases. TiC is the most stable carbides among the three carbide phases. The carburization-decarburization process in A709 is dictated by M 23 C 6 carbides at high carbon activities and by TiC at low carbon activities. Formation of TiC increases the decarburization resistance of A709 steel. Because of the complexity of the precipitation process in A709 during sodium exposures, detailed characterization of precipitates in sodium-exposed specimens and improved thermodynamic models are needed to understand the carburization-decarburization behavior of A709. Kinetic analysis of carbon transfer will be investigated for G91 and A709 in future work.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Effects of Sodium Exposure on the Tensile Properties of Grade 91 steel

Two heats of G91 steel have been investigated in liquid sodium at 550- 650°C to understand their corrosion behaviour, microstructural evolution, and tensile properties. Sodium exposure experiments of subsized, sheet-type tensile specimens of G91 were performed in Argonne’s forced convection sodium materials testing loops, SMT-1 and SMT-2. Maximum exposure times of ~49,000 h at ~38,000 h, and ~20,000 h have been achieved at 550, 600, and 650°C, respectively. Thermal aging experiments of G91 steel were also conducted in parallel, and the results were compared with those of the sodium-exposed specimens to isolate the effect of thermal aging from the effect of sodium exposure. It was found that the yield stress and ultimate tensile strength of G91 were reduced by more than 50% after sodium exposures at 650°C due to decarburization. The reduction in tensile strength was accompanied by the dissolution of M 23 C 6 carbides and excessive grain growth. The effect of sodium at 550 and 600°C was insignificant for exposure times up to ~49,000 h and ~38,000 h, respectively. The effect of sodium on the microstructural stability and tensile properties of G91 is largely dependent on the carbon mass transfer in the alloy-sodium system.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Merger of the Dissolved Gas Method into RELAP5-3D

The Dissolved Gas Model (DGM) was developed for Bettis Atomic Power Laboratory in their own special version of RELAP5-3D known as NUPAC by a subcontractor to. Incorporation of this important code capability in RELAP5-3D was necessary for collaboration between INL and Bettis on future code development efforts. The changes necessary to implement DGM are fundamental, changing the very solver technology at the core of RELAP5-3D, and pervasive, requiring the addition of new subprograms and data, alteration of existing data structures, and modification of more than one hundred subprograms. Moreover, NUPAC and RELAP5-3D had evolved separately and in different directions for a decade and were significantly dissimilar, so the effort was both complex and time-consuming. To ensure its correctness, a large test set comprised thousands of input cases had been constructed and had to run correctly to prove DGM was implemented correctly. A novel method for transporting and adjusting the updates for DGM into RELAP5-3D was undertaken. Tens of thousands of lines of code were either changed or added resulting in a 100% successful implementation of DGM in RELAP5-3D.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗