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Power excursion analysis for high burnup cores

Description: A study was undertaken of power excursions in high burnup cores. There were three objectives in this study. One was to identify boiling water reactor (BWR) and pressurized water reactor (PWR) transients in which there is significant energy deposition in the fuel. Another was to analyze the response of BWRs to the rod drop accident (RDA) and other transients in which there is a power excursion. The last objective was to investigate the sources of uncertainty in the RDA analysis. In a boiling wat… more
Date: February 1996
Creator: Diamond, D. J.; Neymotin, L. & Kohut, P.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Basis for requirements for Advanced Neutron Source (ANS) Control Element Test Facility

Description: The process used to determine the requirements for the Advanced Neutron Source (ANS) Control Element Test Facility (CETF) is explained. The requirements for the CETF are tabulated.
Date: August 1, 1995
Creator: Hendrich, W. R.; Yahr, G. T.; Anderson, J. L.; Battle, R. E.; Litherland, P. S.; Oakes, L. S. et al.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Analysis of flow reversal test

Description: A series of tests has been conducted to measure the dryout power associated with a flow transient whereby the coolant in a heated channel undergoes a change in flow direction. An analysis of the test was made with the aid of a system code, RELAP5. A dryout criterion was developed in terms of a time-averaged void fraction calculated by RELAP5 for the heated channel. The dryout criterion was also compared with several CHF correlations developed for the channel geometry.
Date: March 1, 1996
Creator: Cheng, L. Y. & Tichler, P. R.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Experimental study of static flow instability in subcooled flow boiling in parallel channels

Description: Experimental data for static flow instability or flow excursion (FE) at conditions applicable to the Advanced Neutron Source Reactor are very limited. A series of FE tests with light water flowing vertically upward was completed covering a local exit heat flux range of 0.7--18 MW/m{sup 2}, exit velocity range of 2.8--28.4 m/s, exit pressure range of 0.117--1.7 MPa, and inlet temperature range of 40-- 50{degrees}C. Most of the tests were performed in a ``stiff`` (constant flow) system where the … more
Date: December 31, 1995
Creator: Siman-Tov, M.; Felde, D. K.; McDuffee, J. L. & Yoder, G. L.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Summary of results from the IPIRG-2 round-robin analyses

Description: This report presents a summary of the results from three one-day international round-robin workshops which were organized by Battelle in conjunction with the Second International Piping Integrity Research Group (IPIRG-2) Program. The objective of these workshops was to develop a consensus in handling difficult analytical problems in leak-before-break and pipe flaw evaluations. The workshops, which were held August 5, 1993, March 4, 1994, and October 21, 1994 at Columbus, Ohio, involved various … more
Date: February 1, 1996
Creator: Rahman, S.; Olson, R.; Rosenfield, A. & Wilkowski, G.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Melcor Benchmarking Against Integral Severe Fuel Damage Tests

Description: MELCOR is a fully integrated computer code that models all phases of the progression of severe accidents in light water reactor nuclear power plants, and is being developed for the US Nuclear Regulatory Commission (NRC) by Sandia National Laboratories (SNL). Brookhaven National Laboratory (BNL) has a program with the NRC to provide independent assessment of MELCOR, and a very important part of this program is to benchmark MELCOR against experimental data from integral severe fuel damage tests a… more
Date: December 31, 1994
Creator: Madni, I.K.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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RELAP5 model for advanced neutron source reactor thermal-hydraulic transients, three-element-core design

Description: In order to utilize reduced enrichment fuel, the three-element-core design has been proposed. The proposed core configuration consists of inner, middle, and outer elements, with the middle element offset axially beneath the inner and outer elements, which are axially aligned. The three-element-core RELAP5 model assumes that the reactor hardware is changed only within the core region, so that the loop piping, heat exchangers, and pumps remain as assumed for the two-element-core configuration. Ho… more
Date: February 1, 1996
Creator: Chen, N. C. J.; Wendel, M. W. & Yoder, G. L.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Thermal-hydraulic instabilities in pressure tube graphite-moderated boiling water reactors

Description: Thermally induced two-phase instabilities in non-uniformly heated boiling charmers in RBMK-1000 reactor have been analyzed using RELAP5/MOD3 code. The RELAP5 model of a RBMK-1000 reactor was developed to investigate low flow in a distribution group header (DGH) supplying 44 fuel pressure tubes. The model was evaluated against experimental data. The results of the calculations indicate that the period of oscillation for the high power tube varied from 3.1s to 2.6s, over the power range of 2.0 MW… more
Date: September 1, 1995
Creator: Tsiklauri, G. & Schmitt, B.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Implementation of the high-order schemes QUICK and LECUSSO in the COMMIX-1C Program

Description: Multidimensional analysis computer programs based on the finite volume method, such as COMMIX-1C, have been commonly used to simulate thermal-hydraulic phenomena in engineering systems such as nuclear reactors. In COMMIX-1C, the first-order schemes with respect to both space and time are used. In many situations such as flow recirculations and stratifications with steep gradient of velocity and temperature fields, however, high-order difference schemes are necessary for an accurate prediction o… more
Date: August 1, 1995
Creator: Sakai, K.; Sun, J.G. & Sha, W.T.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Investigation of a Steam Generator Tube Rupture Sequence Using VICTORIA

Description: VICTORIA-92 is a mechanistic computer code for analyzing fission product behavior within the reactor coolant system (RCS) during a severe reactor accident. It provides detailed predictions of the release of radionuclides and nonradioactive materials from the core and transport of these materials within the RCS. The modeling accounts for the chemical and aerosol processes that affect radionuclide behavior. Coupling of detailed chemistry and aerosol packages is a unique feature of VICTORIA; it al… more
Date: December 31, 1995
Creator: Bixler, N. E.; Erickson, C. M. & Schaperow, J. H.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Core damage frequency prespectives for BWR 3/4 and Westinghouse 4-loop plants based on IPE results

Description: This paper discusses the core damage frequency (CDF) insights gained by analyzing the results of the Individual Plant Examinations (IPES) for two groups of plants: boiling water reactor (BWR) 3/4 plants with Reactor Core Isolation Cooling systems, and Westinghouse 4-loop plants. Wide variability was observed for the plant CDFs and for the CDFs of the contributing accident classes. On average, transients-with loss of injection, station blackout sequences, and transients with loss of decay heat r… more
Date: December 31, 1995
Creator: Dingman, S.; Camp, S.; LaChance, J. & Mary Drouin
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Current and prospective safety issues at the HFBR

Description: The Brookhaven High Flux Beam Reactor (HFBR) was designed primarily to produce external neutron beams for experimental research. It is cooled, moderated and reflected by heavy water and uses MTR-ETR type fuel elements containing enriched uranium. The reactor power when operation began in 19965 was 40 MW, was raised to 60 MW in 1982 after a number of plant modifications, and operated at that level until 1989. Since that time safety questions have been raised which resulted in extended shutdowns … more
Date: March 1, 1996
Creator: Tichler, P.R.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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A direct comparison of MELCOR 1.8.3 and MAAP4 results for several PWR & BWR accident sequences

Description: This paper presents a comparison of calculations of severe accident progression for several postulated accident sequences for representative Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR) nuclear power plants performed with the MELCOR 1.8.3 and the MAAP4 computer codes. The PWR system examined in this study is a 1100 MWe system similar in design to a Westinghouse 3-loop plant with a large dry containment; the BWR is a 1100 MWe system similar in design to General Electric BWR/… more
Date: August 1, 1996
Creator: Leonard, M.T.; Ashbaugh, S.G.; Cole, R.K.; Bergeron, K.D. & Nagashima, K.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Dense inclined flows: Theory and experiments. Final report

Description: Rapid, gravity-driven flows of granular materials down inclines pose a challenge to our understanding. Even in situations in which the flow is steady and two-dimensional, the details of how momentum ad energy are balanced within the flow and at the bottom boundary are not well understood. Thus we have undertaken a research program integrating theory, computer simulation, and experiment that focuses on such flows. the effort involves the development of theory informed by the results of simultane… more
Date: December 1, 1995
Creator: Jenkins, J.T. & Louge, M.Y.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Hydraulic properties of Hanford Waste Vitrification Project 39-4 frit slurries

Description: Pacific Northwest Laboratory (PNL) conducted this study for Fluor Daniel Inc. and Westinghouse Hanford Company. The purpose of the study was to assess the effect of solids loading on the hydraulic properties of frit slurries. The effect of solids loading on the hydraulic properties of the fret slurries was evaluated by testing various concentrations of frit slurries in various sized schedule 40 stainless steel piping. The pressure drop in straight and 90-degree long radius elbow sections was me… more
Date: March 1, 1996
Creator: Abrigo, G. P.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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PRODIAG: Combined expert system/neural network for process fault diagnosis. Volume 2, Code manual

Description: We recommend the reader first review Volume 1 of this document, Code Theory, before reading Volume 2. In this volume we make extensive use of terms and concepts described and defined in Volume 1 which are not redefined here to the same extent. To try to reduce the amount of redundant information, we have restricted this volume to the presentation of the expert system code and refer back to the theory described in Volume 1 when necessary. Verification and validation of the results are presented … more
Date: September 1, 1995
Creator: Reifman, J. & Wei, T.Y.C.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Bayesian risk-based acceptance criteria

Description: Mechanistic (or deterministic) analysis is traditionally performed in the process of designing a new nuclear reactor or reactor core and also as part of the safety analysis of existing reactors or reload cores. Mechanistic accident analysis is characterized by the specification of an initial operating condition, an initiating event, and subsequent system faults. These subsequent faults are often chosen, through such mechanisms as the worst single failure criterion, so as to maximize the consequ… more
Date: April 1, 1996
Creator: Martz, H. F.; Abramson, L. R. & Johnson, J. W.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Improved timestep-size diagnostic edits for TRAC-P

Description: Improvements have been made to the timestep-size selection logic diagnostic edits of the Transient Reactor Analysis Code (TRAC), specifically to the TRAC-P version. These include both a precise account of the reason for the selection for individual timesteps and thermal-hydraulic information on mesh cells that control the timestep size. The new edits can be specified by user input as a range of timestep numbers, problem time, or both. A description of the current timestep controls in effect in … more
Date: April 1, 1996
Creator: Giguere, P.T.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Vibration and stability of a group of tubes in crossflow

Description: This paper presents an unsteady flow theory for flow-induced vibration and instability of tube arrays in crossflow. It includes measurements of motion-dependent fluid forces, mathematical model, and experiments on nonlinear response of tube arrays. The unsteady flow theory can be used to provide answers to complex vibration problems in steam generators.
Date: December 31, 1995
Creator: Chen, S.S. & Cai, Y.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Flow blockage analysis for the advanced neutron source reactor

Description: The Advanced Neutron Source (ANS) reactor was designed to provide a research tool with capabilities beyond those of any existing reactors. One portion of its state-of-the-art design required high-speed fluid flow through narrow channels between the fuel plates in the core. Experience with previous reactors has shown that fuel plate damage can occur when debris becomes lodged at the entrance to these channels. Such debris disrupts the fluid flow to the plate surfaces and can prevent adequate coo… more
Date: January 1, 1996
Creator: Stovall, T. K.; Crabtree, J. A.; Felde, D. K. & Park, J. E.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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TMI-2 analysis using SCDAP/RELAP5/MOD3.1

Description: SCDAP/RELAP5/MOD3.1, an integrated thermal hydraulic analysis code developed primarily to simulate severe accidents in nuclear power plants, was used to predict the progression of core damage during the TMI-2 accident. The version of the code used for the TMI-2 analysis described in this paper includes models to predict core heatup, core geometry changes, and the relocation of molten core debris to the lower plenum of the reactor vessel. This paper describes the TMI-2 input model, initial condi… more
Date: November 1, 1994
Creator: Hohorst, J. K.; Polkinghorne, S. T.; Siefken, L. J.; Allison, C. M. & Dobbe, C. A.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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The Advanced Neutron Source liquid deuterium cold source

Description: The Advanced Neutron Source will employ two cold sources to moderate neutrons to low energy (<10 meV). The cold neutrons produced are then passed through beam guides to various experiment stations. Each cold source moderator is a sphere of 410-mm internal diameter. The moderator material is liquid deuterium flowing at a rate of 1 kg/s and maintained at subcooled temperatures at all points of the circuit, to prevent boiling. Nuclear beat deposited within the liquid deuterium and its containment … more
Date: August 1, 1995
Creator: Lucas, A.T.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Thermal-hydraulic studies of the Advanced Neutron Source cold source

Description: The Advanced Neutron Source (ANS), in its conceptual design phase at Oak Ridge National Laboratory, was to be a user-oriented neutron research facility producing the most intense steady-state flux of thermal and cold neutrons in the world. Among its many scientific applications, the production of cold neutrons was a significant research mission for the ANS. The cold neutrons come from two independent cold sources positioned near the reactor core. Contained by an aluminum alloy vessel, each cold… more
Date: August 1, 1995
Creator: Williams, P.T. & Lucas, A.T.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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