Search NASA⌕ Search

SEARCH · Search NASA

Results for “VIPR”

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.

14 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

Validation and Benchmarking of CTF For Single- And Two-Phase Flow Using VIPRE-01

As part of the Consortium for Advanced Simulation of Light Water Reactors (CASL), the subchannel code CTF is being used for single- and two-phase flow analysis under light water reactor operating conditions. Accurate determination of flow distribution, pressure drop, and void content is crucial for prediction of margins to thermal crisis and neutronic feedback, and therefore more efficient plant performance. In preparation for the intended applications, CTF has been validated against experimental facilities comprising the General Electric (GE) 3×3 bundle, the BWR Full-size Fine-mesh Bundle Tests (BFBT), the RISO tube, and the PWR Subchannel and Bundle Tests (PSBT). Meanwhile, the licensed and industrially well-recognized subchannel code VIPRE-01 is used to generate a baseline set of simulations for the targeted tests, and solution parameters are compared to CTF results. The flow split verification problem and single-phase GE 3×3 result in essentially perfect agreement between CTF and VIPRE-01. For two-phase GE 3×3 cases, flow and quality discrepancies arise in the annular-mist flow regime, yet significant improvement is observed in CTF when void drift and two-phase turbulent mixing enhancement are considered. BFBT pressure drop benchmark shows close agreement between predicted and measured results in general, although considerable over-prediction of CTF is observed at relatively high void locations of the facility. This over-estimation tendency is confirmed by RISO cases. While overall statistics are satisfactory, the BFBT bubbly/slug/churn-turbulent flow void contents are markedly over-predicted by CTF, the latter being consistent with PSBT results. Such two-phase closure terms in CTF as turbulent mixing, interfacial and wall friction, and subcooled boiling heat transfer need further improvement using more mechanistic models and high-to-low calibration techniques

Zhao, X↗
DOE OSTI2020

Subchannel codes: CTF and VIPRE-01

In this chapter, two state-of-the-art and well-used subchannel codes are presented: CTF and VIPRE-01. As part of the Consortium for Advanced Simulation of Light Water Reactors, CTF is being used for single- and two-phase flow analysis under light-water reactor normal and anticipated transient operating conditions. Accurate determination of flow distribution, pressure drop, and void content is crucial for predicting margins to thermal crisis and ensuring more efficient nuclear power plant performance. In preparation for the intended applications, CTF is extensively validated against data from experimental facilities. Meanwhile, the licensed commercial code VIPRE-01 is used to generate a baseline set of simulations for the targeted tests and the solution parameters are compared to CTF results. The closure models in CTF that are in the greatest need of improvement are highlighted and discussed.

Zhao, Xingang↗
DOE OSTI2022

Introducing the Bacterial and Viral Bioinformatics Resource Center (BV-BRC): a resource combining PATRIC, IRD and ViPR

The National Institute of Allergy and Infectious Diseases (NIAID) established the Bioinformatics Resource Center (BRC) program to assist researchers with analyzing the growing body of genome sequence and other omics-related data. In this report, we describe the merger of the PAThosystems Resource Integration Center (PATRIC), the Influenza Research Database (IRD) and the Virus Pathogen Database and Analysis Resource (ViPR) BRCs to form the Bacterial and Viral Bioinformatics Resource Center (BV-BRC) https://www.bv-brc.org/. The combined BV-BRC leverages the functionality of the bacterial and viral resources to provide a unified data model, enhanced web-based visualization and analysis tools, bioinformatics services, and a powerful suite of command line tools that benefit the bacterial and viral research communities.

59 BASIC BIOLOGICAL SCIENCES↗
DOE OSTI2011

Investigation of advanced UQ for CRUD prediction with VIPRE.

This document summarizes the results from a level 3 milestone study within the CASL VUQ effort. It demonstrates the application of 'advanced UQ,' in particular dimension-adaptive p-refinement for polynomial chaos and stochastic collocation. The study calculates statistics for several quantities of interest that are indicators for the formation of CRUD (Chalk River unidentified deposit), which can lead to CIPS (CRUD induced power shift). Stochastic expansion methods are attractive methods for uncertainty quantification due to their fast convergence properties. For smooth functions (i.e., analytic, infinitely-differentiable) in L{sup 2} (i.e., possessing finite variance), exponential convergence rates can be obtained under order refinement for integrated statistical quantities of interest such as mean, variance, and probability. Two stochastic expansion methods are of interest: nonintrusive polynomial chaos expansion (PCE), which computes coefficients for a known basis of multivariate orthogonal polynomials, and stochastic collocation (SC), which forms multivariate interpolation polynomials for known coefficients. Within the DAKOTA project, recent research in stochastic expansion methods has focused on automated polynomial order refinement ('p-refinement') of expansions to support scalability to higher dimensional random input spaces [4, 3]. By preferentially refining only in the most important dimensions of the input space, the applicability of these methods can be extended from O(10{sup 0})-O(10{sup 1}) random variables to O(10{sup 2}) and beyond, depending on the degree of anisotropy (i.e., the extent to which randominput variables have differing degrees of influence on the statistical quantities of interest (QOIs)). Thus, the purpose of this study is to investigate the application of these adaptive stochastic expansion methods to the analysis of CRUD using the VIPRE simulation tools for two different plant models of differing random dimension, anisotropy, and smoothness.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2011

CASL L1 Milestone Report: CASL.P4.01, Sensitivity and Uncertainty Analysis for CIPS with VIPRE-W and BOA

The CASL Level 1 Milestone CASL.P4.01, successfully completed in December 2011, aimed to 'conduct, using methodologies integrated into VERA, a detailed sensitivity analysis and uncertainty quantification of a crud-relevant problem with baseline VERA capabilities (ANC/VIPRE-W/BOA).' The VUQ focus area led this effort, in partnership with AMA, and with support from VRI. DAKOTA was coupled to existing VIPRE-W thermal-hydraulics and BOA crud/boron deposit simulations representing a pressurized water reactor (PWR) that previously experienced crud-induced power shift (CIPS). This work supports understanding of CIPS by exploring the sensitivity and uncertainty in BOA outputs with respect to uncertain operating and model parameters. This report summarizes work coupling the software tools, characterizing uncertainties, and analyzing the results of iterative sensitivity and uncertainty studies. These studies focused on sensitivity and uncertainty of CIPS indicators calculated by the current version of the BOA code used in the industry. Challenges with this kind of analysis are identified to inform follow-on research goals and VERA development targeting crud-related challenge problems.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

In Search of a Cost-Effective Approach for Estimating Dam Passage Survival

Abstract Passage of downstream-migrating fishes through hydropower dams poses a management issue in rivers throughout the world. As such, regulations have been enacted at many locales to understand and limit fish losses by requiring dam passage survival estimation. However, limiting estimates of survival to the dam and immediate tailrace presents a challenge to researchers. In this study, two alternative release–recapture methods were used to estimate dam passage survival of juvenile salmonids through Lower Granite Dam on the Snake River, Washington, in 2018. One approach, the virtual/paired-release (ViPRe) model, was extensively used in the Columbia–Snake River basin at federally operated hydroelectric dams during 2010–2014. This existing approach uses three releases of tagged smolts to isolate dam passage survival, defined as survival from the upstream dam face to the tailrace mixing zone 1–2 km downstream of the dam. An alternative approach, the virtual release/dead-fish correction (ViRDCt) model, uses one live-release group paired with a release of dead tagged fish at the dam to estimate the same survival parameter. The alternative estimation approaches were tested on two spring out-migrating stocks, yearling Chinook Salmon Oncorhynchus tshawytscha and steelhead Oncorhynchus mykiss, and on a summer out-migrating stock of subyearling Chinook Salmon. The alternative estimates for these stocks were all within 1 SE. However, the ViRDCt model produced survival estimates with SEs that were 59% smaller than those of the existing ViPRe model and did so using 42% fewer tagged fish and one less downstream acoustic detection array. Because of the reduced sample size, nearly US$540,000 less in tag costs and 4,000 fewer labor-hours were required to implement the ViRDCt model compared to the ViPRe model. As such, the ViRDCt model represents a cost-effective and precise approach to characterizing dam passage survival.

59 BASIC BIOLOGICAL SCIENCES↗
DOE OSTI2016

CASL COBRA-TF improvements for PWR DNB analysis

COBRA-TF (CTF) is a thermal hydraulic (T/H) subchannel code using either three-dimensional (3D) Cartesian or subchannel coordinate formulations for two-phase fluid flow and heat transfer solutions. CTF has been improved under the Consortium for Advanced Simulation of Light Water Reactors (CASL) program for Pressurized Water Reactor (PWR) applications, including software optimization, new closure models, pre- and post-processing and parallelization for modeling full reactor core T/H responses under normal operating and accident conditions. As a result of collaboration among CASL partners including the Westinghouse Electric Company, the Oak Ridge National Laboratory (ORNL), and the Sandia National Laboratories, additional modeling improvements were made to CTF specifically for PWR Departure from Nucleate Boiling (DNB) analysis, including a code option to evaluate fuel thermal margin in terms of DNB Ratio (DNBR) and an axial shape factor to account for effect of non-uniform axial power distribution on DNB. Multiple DNB correlations are now linked with CTF for different applications, including the Westinghouse proprietary WRB-1 correlation for fuel designs containing mixing vane grid spacers. The improved CTF code with the WRB-1 correlation (CTF/WRB-1) was validated using the DNB data from the PWR Subchannel Bundle Tests (PSBT). In addition to the comparison with the test data, the CTF/WRB-1 DNBR results and the associated local fluid conditions were compared to the results of the Westinghouse T/H design code, VIPRE-W, which is an enhanced version of the VIPRE-01 code originally developed by the Electric Power Research Institute (EPRI). The comparisons showed that CTF/WRB-1 DNBR predictions are in good agreement with the VIPRE-W results within the applicable range of the DNB correlation. A model sensitivity study was performed to confirm that the CTF void drift model had an insignificant effect on DNBR under the steam line break (SLB) low pressure condition. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Demonstration of a New Water Vapor Radar (Field Campaign Report)

The purpose of the project was to demonstrate and validate observations from a new differential absorption radar, which uses an active sounding approach to remotely measure water vapor within clouds and precipitation. The instrument, the vapor in-cloud profiling radar (VIPR), is the first radar of its kind using a differential absorption technique to profile water vapor. The demonstration and validation of these data is essential to proving the utility of these observations in combination with the standard remote-sensing approaches employed at SGP. This initial effort is an essential element to facility the incorporation of the VIPR radar or its derivatives in future focused efforts to understand the role of water vapor in shaping the convective environment.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2016

DNB evaluation of PWR steam-line break event using Westinghouse and CASL coupled code systems

ANCKVIPRE and the VERA-CS code systems have been successfully applied to predicting PWR core response with respect to departure from nucleate boiling (DNB) at the limiting time step of a postulated hot zero power (HZP) main steam-line break (MSLB) event with and without offsite power available. ANCKVIPRE links the ANC code with the VIPRE-W code. ANC is Westinghouse's multi-dimensional nodal diffusion code for all nuclear core design calculations. It predicts core reactivity, assembly power, rod power, detector thimble flux, and other relevant core characteristics. VIPRE-W is the Westinghouse version of a thermalhydraulic subchannel code developed for light water reactor core design applications. The reactor core simulation was based on input of the core boundary conditions from RETRAN system transient and STAR-CCM+ CFD core inlet distribution calculations. To account for the asymmetric power distribution in the reactor core due to the broken steam pipe in one loop and the stuck RCCA, both ANCKVIPRE and the VERA-CS models were set up for the whole reactor core. The advanced modeling and simulations confirmed that the HZP MSLB high-flow case with offsite power available is more limiting with respect to the DNB acceptance criterion, consistent with the limiting case analyzed in the 4-loop plant Safety Analysis Report (SAR). The results also demonstrated that the currently NRC-approved Westinghouse coupled code system such as ANCKVIPRE remains conservative.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Assessment of the Subchannel Code CTF for Single- and Two-Phase Flows

As part of the Consortium for Advanced Simulation of Light Water Reactors, the subchannel code CTF is being used for single-phase and two-phase flow analysis under light water reactor operating conditions. Accurate determination of flow distribution, pressure drop, and void content is crucial for predicting margins to thermal crisis and ensuring more efficient plant performance. In preparation for the intended applications, CTF has been validated against data from experimental facilities comprising the General Electric (GE) 3 × 3 bundle, the boiling water reactor full-size fine-mesh bundle tests (BFBTs), the Risø tube, and the pressurized water reactor subchannel and bundle tests (PSBTs). Meanwhile, the licensed, well-recognized subchannel code VIPRE-01 was used to generate a baseline set of simulations for the targeted tests and solution parameters were compared to the CTF results.The flow split verification problem and single-phase GE 3 × 3 results are essentially in perfect agreement between the two codes. For the two-phase GE 3 × 3 cases, flow and quality discrepancies arise in the annular-mist flow regime, yet significant improvement is observed in CTF when void drift and two-phase turbulent mixing enhancement are considered. The BFBT pressure drop benchmark shows close agreement between predicted and measured results in general, although considerable overprediction by CTF is observed at relatively high void locations of the facility. This overestimation tendency is confirmed by the Risø cases. While overall statistics are satisfactory, both BFBT and PSBT bubbly-to-churn flow void contents are markedly overpredicted by CTF.The issues with two-phase closures such as turbulent mixing, interfacial and wall friction, and subcooled boiling heat transfer need to be addressed. As a result, preliminary sensitivity studies are presented herein, but more advanced models and code stability analysis require further investigation.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Review of CTF’s Fuel Rod Modeling Using FRAPCON-4.0’s Centerline Temperature Predictions

Coolant Boiling in Rod Arrays–Two Fluid (COBRA-TF), or CTF1 [1], is a nuclear thermal hydraulic subchannel code used throughout academia and industry. CTF’s fuel rod modeling is originally developed for VIPRE code [2]. Its methodology is based on GAPCON [3] and FRAP [4] fuel performance codes, and material properties are included from MATPRO handbook [5]. This work focuses on review of CTF’s fuel rod modeling to address shortcomings in CTF’s temperaturepredictions. CTF is compared to FRAPCON which is U.S. NRC’s steady-state fuel performance code for light-water reactor fuel rods. FRAPCON calculates the changes in fuel rod variables and temperatures including the eects of cladding hoop strain, cladding oxidation, hydriding, fuel irradiation swelling, densification, fission gas release and rod internal gas pressure. It uses fuel, clad and gap material properties from MATPRO. Additionally, it has its own models for fission gas release, cladding corrosion and cladding hydrogen pickup. It allows finite dierence or finite element approaches for its mechanical model.In this study, FRAPCON-4.0 [6] is used as a reference fuel performance code. In comparison, Halden Reactor Data for IFA432 Rod 1 and Rod 3. CTF simulations are performed in two ways; informing CTF with gap conductance value from FRAPCON, and using CTF’s dynamic gap conductance model.First case is chosen to show temperature is predicted correctly with CTF’s models for thermal and cladding conductivities once gap conductance is provided. Latter is to review CTF’s dynamic gap conductance model. These Halden test cases are selected to be representative of cases with and without any physical contact between fuel-pellet and clad while reviewing functionality of CTF’s dynamic gap conductance model. Improving the CTF’s dynamic gap conductance model will allowprediction of fuel and cladding thermo-mechanical behavior under irradiation, and better temperature feedbacks from CTF in transient calculations.

Toptan, Aysenur↗
DOE OSTI2012

3D analysis of the reactivity insertion accident in VVER-1000

Fuel parameters such as peak enthalpy and temperature during rod ejection accident are calculated. The calculations are performed by 3D neutron kinetics code NESTLE and 3D thermal-hydraulic code VIPRE-W. Both hot zero power and hot full power cases were studied for an equilibrium cycle with Westinghouse hex fuel in VVER-1000. It is shown that the use of 3D methodology can significantly increase safety margins for current criteria and met future criteria. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Nitride Fuel Potential for Power Uprates

This work presents the methodology and current results for design analysis of grid supported pressurized water reactor cores using uranium nitride (UN) fuel in square arrays. Uranium nitride has a higher theoretical density and higher thermal conductivity than uranium oxide, allowing for lower fuel temperatures, longer cycle length and higher power densities with similar enrichment. In order to avoid the UN and high temperature water reaction, silicon carbide cladding has been proposed in place of Zircaloy cladding. The analysis aims at finding the maximum thermal power over a range of bundle geometries (rod diameter and pitch). The same methodology is applied to the design of UO{sub 2} oxide fueled cores to provide a fair comparison of the achievable power rating between the two fuel types. Steady-state limits are considered, including fuel bundle pressure drop, minimum departure from nucleate boiling ratio (MDNBR), fuel temperature, fuel rod vibrations and pin fretting and sliding wear. Transients such as loss of flow and overpower accident will be considered. A steady state analysis was undertaken to determine the maximum achievable power that can be sustained by square array fuel assemblies loaded with nitride and oxide fuels with same or better safety performances as the reference design. Steady-state design limits were applied including MDNBR, fuel bundle pressure drop, fuel temperature, cumulative sliding and fretting wear. The major conclusions of this study are as follows: 1) It is possible to achieve significant power increase for UN and UO{sub 2} compared to the reference core power output of 3400 MWth with smaller but more numerous rods and higher pitch over diameter ratios. A power uprate of 25% can be reached according to the VIPRE-MATLAB interface for full core analysis with the same number of spacer grids, and 32% according to the new MATLAB script for single hot channel analysis, with five more spacer grids. For this latter core, a higher enrichment would be required. 2) The nitride fuel temperature constraint from dissociation doesn't influence steady state performances as other design constraints such as MDNBR and pressure loss are more restrictive. The oxide fuel temperature constraint from fission gas release is reached in part of the study domain, but not in the region that allows for possible uprates. 3) As a result, the thermal hydraulic performance of UN and UO{sub 2} are similar. Differences in possible power uprates will arise from fretting and sliding wear. Wear depends of the time spent by the fuel in the core, which depends on burnup and heavy metal inventory. UN being denser, it can allow for power uprates or increased cycle length. There is a trade off between power uprates and cycle length with regards to fuel pin wear. Reducing fuel pin wear by reducing cycle length can allow for higher thermal power. This affects the plants' economics as well. This work will be completed by a transient analysis using CASMO4E, SIMULATE3 and SIMULATE3k to model the neutronics of a nitride fueled core. The transient performances will be used to determine the maximum power uprates while performing as well or better during transients as the reference design. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Review of CTF's Fuel Rod Modeling Using FRAPCON-4.0's Centerline Temperature Predictions

Coolant Boiling in Rod Arrays-Two Fluid (COBRA-TF), or CTF1, is a nuclear thermal hydraulic subchannel code used throughout academia and industry. CTF's fuel rod modeling is originally developed for VIPRE code. Its methodology is based on GAPCON and FRAP fuel performance codes, and material properties are included from MATPRO handbook. This work focuses on review of CTF's fuel rod modeling to address shortcomings in CTF's temperature predictions. CTF is compared to FRAPCON which is U.S. NRC's steady-state fuel performance code for light-water reactor fuel rods. FRAPCON calculates the changes in fuel rod variables and temperatures including the effects of cladding hoop strain, cladding oxidation, hydriding, fuel irradiation swelling, densification, fission gas release and rod internal gas pressure. It uses fuel, clad and gap material properties from MATPRO. Additionally, it has its own models for fission gas release, cladding corrosion and cladding hydrogen pickup. It allows finite difference or finite element approaches for its mechanical model. In this study, FRAPCON-4.0 is used as a reference fuel performance code. In comparison, Halden Reactor Data for IFA432 Rod 1 and Rod 3. CTF simulations are performed in two ways; informing CTF with gap conductance value from FRAPCON, and using CTF's dynamic gap conductance model. First case is chosen to show temperature is predicted correctly with CTF's models for thermal and cladding conductivities once gap conductance is provided. Latter is to review CTF's dynamic gap conductance model. These Halden test cases are selected to be representative of cases with and without any physical contact between fuel-pellet and clad while reviewing functionality of CTF's dynamic gap conductance model. Improving the CTF's dynamic gap conductance model will allow prediction of fuel and cladding thermo-mechanical behavior under irradiation, and better temperature feedbacks from CTF in transient calculations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗