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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 1,603 records · Page 89
All sourcesNASA NTRSDOE OSTI
DOE OSTI2014

Benchmark Specifications and Results for ßeff in a HEU Metal System Using ORSphere

In the early 1970s Dr. John T. Mihalczo (team leader), J.J. Lynn, and J.R. Taylor performed experiments at the Oak Ridge Critical Experiments Facility (ORCEF) with highly enriched uranium (HEU) metal (called Oak Ridge Alloy or ORALLOY) in an attempt to recreate GODIVA I results with greater accuracy than those performed at Los Alamos National Laboratory in the 1950s (HEU-MET-FAST-001). The purpose of the Oak Ridge ORALLOY Sphere (ORSphere) experiments was to estimate the unreflected and unmoderated critical mass of an idealized sphere of uranium metal corrected to a density, purity, and enrichment such that it could be compared with the GODIVA I experiments. Part of the experimental series was the measurement of the delayed neutron fraction, ßeff, using time correlation measurements and using the central void reactivity measurement. The time correlations measurements were rejected by the experimenter. The measurements using the central void reactivity measurement yielded a ßeff value of 0.00657, which agrees well with the value measured with GODIVA I (0.0066). This measurement is evaluated, found to be acceptable, and discussed in extensive detail in “ORSphere: Physics Measurements for Bare, HEU(93.2) Metal Sphere". In order to determine the delayed neutron fraction using the central void reactivity delayed neutron parameters must be used. The experimenter utilized the delayed neutron parameters set forth by Keepin, Wimment, and Zeigler. If the derivation of the ßeff is repeated with different delayed neutron parameters from various modern nuclear data sets the resulting values vary greatly from the expected results.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Benchmarking and Disclosing Resilience Performance

This fact sheet provides an overview of trends and barriers related to disclosure of climate and resilience performance. It reviews major reporting frameworks for resilience, compares the frameworks, and provides next steps for building owners.

Resilience, Resiliency, Benchmarking, Sustainabili↗
DOE OSTI2012

Benchmark of Atucha-2 PHWR RELAP5-3D control rod model by Monte Carlo MCNP5 core calculation

Atucha-2 is a Siemens-designed PHWR reactor under construction in the Republic of Argentina. Its geometrical complexity and peculiarities require the adoption of advanced Monte Carlo codes for performing realistic neutronic simulations. Therefore core models of Atucha-2 PHWR were developed using MCNP5. In this work a methodology was set up to collect the flux in the hexagonal mesh by which the Atucha-2 core is represented. The scope of this activity is to evaluate the effect of obliquely inserted control rod on neutron flux in order to validate the RELAP5-3D{sup C}/NESTLE three dimensional neutron kinetic coupled thermal-hydraulic model, applied by GRNSPG/UNIPI for performing selected transients of Chapter 15 FSAR of Atucha-2. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

TRACE and TRAC-BF1 benchmark against Leibstadt plant data during the event inadvertent opening of relief valves

In framework of introducing TRACE code to transient analyses system codes for Leibstadt Power Plant (KKL), a conversion process of existing TRAC-BF1 model to TRACE has been started within KKL. In the first step, TRACE thermal-hydraulic model for KKL has been developed based on existing TRAC-BF1 model. In order to assess the code models a simulation of plant transient event is required. In this matter simulations of inadvertent opening of 8 relief valves event have been performed. The event occurs at KKL during normal operation, and it started when 8 relief valves open resulting in depressurization of the Reactor Pressure Vessel (RPV). The reactor was shutdown safely by SCRAM at low level. The high pressure core spray (HPCS) and the reactor core isolation cooling (RCIC) have been started manually in order to compensate the level drop. The remaining water in the feedwater (FW) lines flashes due to saturation conditions originated from RPV depressurization and refills the reactor downcomer. The plant boundary conditions have been used in the simulations and the FW flow rate has been adjusted for better prediction. The simulations reproduce the plant data with good agreement. It can be concluded that the TRAC-BF1 existing model has been used successfully to develop the TRACE model and the results of the calculations have shown good agreement with plant recorded data. Beside the modeling assessment, the TRACE and TRAC-BF1 capabilities to reproduce plant physical behavior during the transient have shown satisfactory results. The first step of developing KKL model for TRACE has been successfully achieved and this model is further developed in order to simulate more complex plant behavior such as Turbine Trip. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Benchmark studies of thermal jet mixing in SFRs using a two-jet model

To guide the modeling, simulations and design of Sodium Fast Reactors (SFRs), we explore and compare the predictive capabilities of two numerical solvers COMSOL and OpenFOAM in the thermal jet mixing of two buoyant jets typical of the outlet flow from a SFR tube bundle. This process will help optimize on-going experimental efforts at obtaining high resolution data for V and V of CFD codes as anticipated in next generation nuclear systems. Using the k-{epsilon} turbulence models of both codes as reference, their ability to simulate the turbulence behavior in similar environments was first validated for single jet experimental data reported in literature. This study investigates the thermal mixing of two parallel jets having a temperature difference (hot-to-cold) {Delta}T{sub hc}= 5 deg. C, 10 deg. C and velocity ratios U{sub c}/U{sub h} = 0.5, 1. Results of the computed turbulent quantities due to convective mixing and the variations in flow field along the axial position are presented. In addition, this study also evaluates the effect of spacing ratio between jets in predicting the flow field and jet behavior in near and far fields. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2013

Benchmarking the New RESRAD-OFFSITE Source Term Model with DUST-MS and GoldSim - 13377

RESRAD-OFFSITE is a computer code developed by Argonne National Laboratory under the sponsorship of U.S. Department of Energy (DOE) and U.S. Nuclear Regulatory Commission (NRC). It is designed on the basis of RESRAD (onsite) code, a computer code designated by DOE and NRC for evaluating soil-contaminated sites for compliance with human health protection requirements pertaining to license termination or environmental remediation. RESRAD-OFFSITE has enhanced capabilities of modeling radionuclide transport to offsite locations and calculating potential radiation exposure to offsite receptors. Recently, a new source term model was incorporated into RESRAD-OFFSITE to enhance its capability further. This new source term model allows simulation of radionuclide releases from different waste forms, in addition to the soil sources originally considered in RESRAD (onsite) and RESRAD-OFFSITE codes. With this new source term model, a variety of applications can be achieved by using RESRAD-OFFSITE, including but not limited to, assessing the performance of radioactive waste disposal facilities. This paper presents the comparison of radionuclide release rates calculated by the new source term model of RESRAD-OFFSITE versus those calculated by DUST-MS and GoldSim, respectively. The focus of comparison is on the release rates of radionuclides from the bottom of the contaminated zone that was assumed to contain radioactive source materials buried in soil. The transport of released contaminants outside of the primary contaminated zone is beyond the scope of this paper. Overall, the agreement between the RESRAD-OFFSITE results and the DUST-MS and GoldSim results is fairly good, with all three codes predicting identical or similar radionuclide release profiles over time. Numerical dispersion in the DUST-MS and GoldSim results was identified as potentially contributing to the disagreement in the release rates. In general, greater discrepancy in the release rates was found for short-lived, fast-moving radionuclides than for long-lived, slow-moving radionuclides. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Benchmarking of activation rate in special core in LR-0 reactor

The experiments focused on the determination of activation reaction rates of selected IRDFF materials activated in a well-defined special core assembled in the LR-0 reactor are presented in the paper. The used core is designed and verified for IRDFF library testing by reactivity characterization, fission rates distribution and also neutron spectra measurement at the center of the core. {sup 23}Na(n,2n), {sup 75}As(n,2n), {sup 89}Y(n,2n) and {sup 90}Zr(n,2n) reaction cross sections were measured in this special core. The presented neutron cross sections were derived from experimentally determined reaction rates of irradiated samples of NaF, As{sub 2}O{sub 3}, Y{sub 2}O{sub 3}, and ZrO{sub 2}. These reaction rates are derived from gamma activities of the activation products of studied activation reactions. Net Peak Areas (NPA) of photons accompanying decay were measured using a semiconductor High Purity Germanium (HPGe) detector. Despite the fact that mentioned reactions have a threshold above 10 MeV, where the neutron spectrum in the special LR-0 core is nearly undistinguishable from {sup 235}U prompt fission neutron spectrum, the measured integral cross sections averaged in LR-0 spectrum were corrected for spectral shift. The obtained values of {sup 235}U spectral average cross section are: (3.84 ± 0.19) μb for {sup 23}Na(n,2n); (0.320 ± 0.015) mb for {sup 75}As(n,2n); (0.171 ± 0.006) mb for {sup 89}Y(n,2n); (0.107 ± 0.005) mb for {sup 90}Zr(n,2n). (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Benchmark for Free-Gas Elastic Scattering with the Heavy-Gas Model and Piecewise-Constant Cross Sections

Accurate Monte Carlo calculations involving thermal neutrons require that codes implement the thermal scattering physics correctly. The heavy-gas model, which is valid for large atomic mass, with constant cross sections in energy has an analytic solution for the thermal neutron spectrum that can be readily used to verify the implementation of the free-gas thermal scattering routines. Previous work has shown that MCNP6.1 correctly computes the free-gas thermal flux with constant cross sections. This summary extends the verification to piecewise-constant cross sections in energy that can better approximate physically realistic cross sections. The obtained solutions for the piecewise-constant cases agree with the respective thermal spectra obtained by MCNP6.1, further verifying that it is accurately simulating the free-gas model. (authors)

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗