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Evaluation of the Use of Existing RELAP5-3D Models to Represent the Actinide Burner Test Reactor

Description: The RELAP5-3D code is being considered as a thermal-hydraulic system code to support the development of the sodium-cooled Actinide Burner Test Reactor as part of Global Nuclear Energy Partnership. An evaluation was performed to determine whether the control system could be used to simulate the effects of non-convective mechanisms of heat transport in the fluid that are not currently represented with internal code models, including axial and radial heat conduction in the fluid and subchannel mixing. The evaluation also determined the relative importance of axial and radial heat conduction and fluid mixing on peak cladding temperature for a wide range of steady conditions and during a representative loss-of-flow transient. The evaluation was performed using a RELAP5-3D model of a subassembly in the Experimental Breeder Reactor-II, which was used as a surrogate for the Actinide Burner Test Reactor. An evaluation was also performed to determine if the existing centrifugal pump model could be used to simulate the performance of electromagnetic pumps.
Date: February 1, 2007
Creator: Davis, C. B.
Partner: UNT Libraries Government Documents Department

Applicability of RELAP5-3D for Thermal-Hydraulic Analyses of a Sodium-Cooled Actinide Burner Test Reactor

Description: The Actinide Burner Test Reactor (ABTR) is envisioned as a sodium-cooled, fast reactor that will burn the actinides generated in light water reactors to reduce nuclear waste and ease proliferation concerns. The RELAP5-3D computer code is being considered as the thermal-hydraulic system code to support the development of the ABTR. An evaluation was performed to determine the applicability of RELAP5-3D for the analysis of a sodium-cooled fast reactor. The applicability evaluation consisted of several steps, including identifying the important transients and phenomena expected in the ABTR, identifying the models and correlations that affect the code’s calculation of the important phenomena, and evaluating the applicability of the important models and correlations for calculating the important phenomena expected in the ABTR. The applicability evaluation identified code improvements and additional models needed to simulate the ABTR. The accuracy of the calculated thermodynamic and transport properties for sodium was also evaluated.
Date: July 1, 2006
Creator: Davis, C. B.
Partner: UNT Libraries Government Documents Department

Assessment of RELAP5-3D{copyright} using data from two-dimensional RPI flow tests

Description: The capability of the RELAP5-3D{copyright} computer code to perform multi-dimensional thermal-hydraulic analysis was assessed using data from steady-state flow tests conducted at Rensselaer Polytechnic Institute (RPI). The RPI data were taken in a two-dimensional test section in a low-pressure air/water loop. The test section consisted of a thin vertical channel that simulated a two-dimensional slice through the core of a pressurized water reactor. Single-phase and two-phase flows were supplied to the test section in an asymmetric manner to generate a two-dimensional flow field. A traversing gamma densitometer was used to measure void fraction at many locations in the test section. High speed photographs provided information on the flow patterns and flow regimes. The RPI test section was modeled using the multi-dimensional component in RELAP5-3D Version BF06. Calculations of three RPI experiments were performed. The flow regimes predicted by the base code were in poor agreement with those observed in the tests. The two-phase regions were observed to be in the bubbly and slug flow regimes in the test. However, nearly all of the junctions in the horizontal direction were calculated to be in the stratified flow regime because of the relatively low velocities in that direction. As a result, the void fraction predictions were also in poor agreement with the measured values. Significantly improved results were obtained in sensitivity calculations with a modified version of the code that prevented the horizontal junctions from entering the stratified flow regime. These results indicate that the code`s logic in the determination of flow regimes in a multi-dimensional component must be improved. The results of the sensitivity calculations also indicate that RELAP5-3D will provide a significant multi-dimensional hydraulic analysis capability once the flow regime prediction is improved.
Date: July 1, 1998
Creator: Davis, C.B.
Partner: UNT Libraries Government Documents Department

Overview of the use of ATHENA for thermal-hydraulic analysis of systems with lead-bismuth coolant

Description: The INEEL and MIT are investigating the suitability of lead-bismuth cooled fast reactor for producing low-cost electricity as well as for actinide burning. This paper is concerned with the general area of thermal-hydraulics of lead-bismuth cooled reactors. The ATHENA code is being used in the thermal-hydraulic design and analysis of lead-bismuth cooled reactors. The ATHENA code was reviewed to determine its applicability for simulating lead-bismuth cooled reactors. Two modifications were made to the code as a result of this review. Specifically, a correlation to represent heat transfer from rod bundles to a liquid metal and a void correlation based on data taken in a mixture of lead-bismuth and steam were added the code. The paper also summarizes the analytical work that is being performed with the code and plans for future analytical work.
Date: April 2, 2000
Creator: Davis, C. B. & Shieh, A. S.
Partner: UNT Libraries Government Documents Department

Assessment of RELAP5-3D multi-dimensional component model using data from LOFT Test L2-5

Description: The capability of the RELAP5-3D computer code to perform multi-dimensional analysis of a pressurized water reactor (PWR) was assessed using data from the Loss-of-Fluid Test (LOFT) L2-5 experiment. The LOFT facility was a 50 MW PWR that was designed to simulate the response of a commercial PWR during a loss-of-coolant accident (LOCA). Test L2-5 simulated a 200% double-ended cold leg break with an immediate primary coolant pump trip. A three-dimensional model of the LOFT reactor vessel was developed. Calculations of the LOFT L21-5 experiment were performed using the RELAP5-3D computer code. The calculations simulated the blowdown, refill, and reflood portions of the transient. The calculated thermal-hydraulic response of the primary coolant system was generally in reasonable agreement with the test. The calculated results were also generally as good as or better than those obtained previously with RELAP5/MOD3.
Date: July 1, 1998
Creator: Davis, C.B.
Partner: UNT Libraries Government Documents Department

Simulation of three-dimensional hydrodynamic components with a one-dimensional transient analysis code

Description: Significant multidimensional pressure gradients occur in the water plenum region of the reactors at the Savannah River Site (SRS). A multidimensional RELAP5 input model of the L-Reactor was developed and benchmarked against SRS data. Although RELAP5 is a one-dimensional code, its cross-flow junction allows a multidimensional analysis capability. RELAP5 and the model of L-Reactor calculated water plenum pressures that were in good agreement with measured values for tests with symmetric and asymmetric flow patterns within the plenum. The results indicate that a one-dimensional code such as RELAP5, in conjunction with a carefully designed input model, can be used to predict hydraulic response even when large multidimensional effects are present. 6 refs., 6 figs.
Date: January 1, 1990
Creator: Shaw, R.A. & Davis, C.B.
Partner: UNT Libraries Government Documents Department

Simulation of German PKL refill/reflood experiment K9A using RELAP4/MOD7. [PWR]

Description: This paper describes a RELAP4/MOD7 simulation of West Germany's Kraftwerk Union (KWU) Primary Coolant Loop (PKL) refill/reflood experiment K9A. RELAP4/MOD7, a best-estimate computer program for the calculation of thermal and hydraulic phenomena in a nuclear reactor or related system, is the latest version in the RELAP4 code development series. This study was the first major simulation using RELAP4/MOD7 since its release by the Idaho National Engineering Laboratory (INEL). The PKL facility is a reduced scale (1:134) representation of a typical West German four-loop 1300 MW pressurized water reactor (PWR). A prototypical scale of the total volume to power ratio was maintained. The test facility was designed specifically for an experiment simulating the refill/reflood phase of a Loss-of-Coolant Accident (LOCA).
Date: November 1, 1981
Creator: Hsu, M.T.; Davis, C.B. & Behling, S.R.
Partner: UNT Libraries Government Documents Department

Method for quantitative assessment of nuclear safety computer codes. [PWR]

Description: A procedure has been developed for the quantitative assessment of nuclear safety computer codes and tested by comparison of RELAP4/MOD6 predictions with results from two Semiscale tests. This paper describes the developed procedure, the application of the procedure to the Semiscale tests, and the results obtained from the comparison.
Date: January 1, 1979
Creator: Dearien, J.A.; Davis, C.B. & Matthews, L.J.
Partner: UNT Libraries Government Documents Department

The Addition of Noncondensable Gases into RELAP5-3D for Analysis of High Temperature Gas-Cooled Reactors

Description: Oxygen, carbon dioxide, and carbon monoxide have been added to the RELAP5-3D computer code as noncondensable gases to support analysis of high temperature gas-cooled reactors. Models of these gases are required to simulate the effects of air ingress on graphite oxidation following a loss-of-coolant accident. Correlations were developed for specific internal energy, thermal conductivity, and viscosity for each gas at temperatures up to 3000 K. The existing model for internal energy (a quadratic function of temperature) was not sufficiently accurate at these high temperatures and was replaced by a more general, fourth-order polynomial. The maximum deviation between the correlations and the underlying data was 2.2% for the specific internal energy and 7% for the specific heat capacity at constant volume. The maximum deviation in the transport properties was 4% for oxygen and carbon monoxide and 12% for carbon dioxide.
Date: August 1, 2003
Creator: Davis, C. B. & Oh, C. H.
Partner: UNT Libraries Government Documents Department

Reactor Pressure Vessel Temperature Analysis for Prismatic and Pebble-Bed VHTR Designs

Description: Analyses were performed to determine maximum temperatures in the reactor pressure vessel for two potential Very-High Temperature Reactor (VHTR) designs during normal operation and during a depressurized conduction cooldown accident. The purpose of the analyses was to aid in the determination of appropriate reactor vessel materials for the VHTR. The designs evaluated utilized both prismatic and pebble-bed cores that generated 600 MW of thermal power. Calculations were performed for fluid outlet temperatures of 900 and 950 °C, corresponding to the expected range for the VHTR. The analyses were performed using the RELAP5-3D and PEBBED-THERMIX computer codes. Results of the calculations were compared with preliminary temperature limits derived from the ASME pressure vessel code.
Date: April 1, 2006
Creator: Gougar, H. D. & Davis, C. B.
Partner: UNT Libraries Government Documents Department

Phenomena identification and ranking tables for Westinghouse AP600 small break loss-of-coolant accident, main steam line break, and steam generator tube rupture scenarios

Description: This report revision incorporates new experimental evidence regarding AP600 behavior during small break loss-of-coolant accidents. This report documents the results of Phenomena Identification and Ranking Table (PIRT) efforts for the Westinghouse AP600 reactor. The purpose of this PIRT is to identify important phenomena so that they may be addressed in both the experimental programs and the RELAP5/MOD3 systems analysis computer code. In Revision of this report, the responses of AP600 during small break loss-of-coolant accident, main steam line break, and steam generator tube rupture accident scenarios were evaluated by a committee of thermal-hydraulic experts. Committee membership included Idaho National Engineering and Environmental Laboratory staff and recognized thermal-hydraulic experts from outside of the laboratory. Each of the accident scenarios was subdivided into separate, sequential periods or phases. Within each phase, the plant behavior is controlled by, at most, a few thermal-hydraulic processes. The committee identified the phenomena influencing those processes, and ranked & influences as being of high, medium, low, or insignificant importance. The primary product of this effort is a series of tables, one for each phase of each accident scenario, describing the thermal-hydraulic phenomena judged by the committee to be important, and the relative ranking of that importance. The rationales for the phenomena selected and their rankings are provided. This document issue incorporates an update of the small break loss-of-coolant accident portion of the report. This revision is the result of the release of experimental evidence from AP600-related integral test facilities (ROSA/AP600, OSU, and SPES) and thermal-hydraulic expert review. The activities associated with this update were performed during the period from June 1995 through November 1996. 8 refs., 26 figs., 42 tabs.
Date: June 1, 1997
Creator: Wilson, G.E.; Fletcher, C.D. & Davis, C.B.
Partner: UNT Libraries Government Documents Department

Assessment of the RELAP5 multi-dimensional component model using data from LOFT test L2-5

Description: The capability of the RELAP5-3D computer code to perform multi-dimensional analysis of a pressurized water reactor (PWR) was assessed using data from the LOFT L2-5 experiment. The LOFT facility was a 50 MW PWR that was designed to simulate the response of a commercial PWR during a loss-of-coolant accident. Test L2-5 simulated a 200% double-ended cold leg break with an immediate primary coolant pump trip. A three-dimensional model of the LOFT reactor vessel was developed. Calculations of the LOFT L2-5 experiment were performed using the RELAP5-3D Version BF02 computer code. The calculated thermal-hydraulic responses of the LOFT primary and secondary coolant systems were generally in reasonable agreement with the test. The calculated results were also generally as good as or better than those obtained previously with RELAP/MOD3.
Date: January 1, 1998
Creator: Davis, C.B.
Partner: UNT Libraries Government Documents Department

Posttest analysis of international standard problem 10 using RELAP4/MOD7. [PWR]

Description: RELAP4/MOD7, a best estimate computer code for the calculation of thermal and hydraulic phenomena in a nuclear reactor or related system, is the latest version in the RELAP4 code development series. This paper evaluates the capability of RELAP4/MOD7 to calculate refill/reflood phenomena. This evaluation uses the data of International Standard Problem 10, which is based on West Germany's KWU PKL refill/reflood experiment K9A. The PKL test facility represents a typical West German four-loop, 1300 MW pressurized water reactor (PWR) in reduced scale while maintaining prototypical volume-to-power ratio. The PKL facility was designed to specifically simulate the refill/reflood phase of a hypothetical loss-of-coolant accident (LOCA).
Date: January 1, 1981
Creator: Hsu, M.; Davis, C.B.; Peterson, A.C. Jr. & Behling, S.R.
Partner: UNT Libraries Government Documents Department

Thermal Hydraulic Analyses for Coupling High Temperature Gas-Cooled Reactor to Hydrogen Plant

Description: The US Department of Energy is investigating the use of high-temperature nuclear reactors to produce hydrogen using either thermochemical cycles or high-temperature electrolysis. Although the hydrogen production processes are in an early stage of development, coupling either of these processes to the high-temperature reactor requires both efficient heat transfer and adequate separation of the facilities to assure that off-normal events in the production facility do not impact the nuclear power plant. An intermediate heat transport loop will be required to separate the operations and safety functions of the nuclear and hydrogen plants. A next generation high-temperature reactor could be envisioned as a single-purpose facility that produces hydrogen or a dual-purpose facility that produces hydrogen and electricity. Early plants, such as the proposed Next Generation Nuclear Plant (NGNP), may be dual-purpose facilities that demonstrate both hydrogen and efficient electrical generation. Later plants could be single-purpose facilities. At this stage of development, both single- and dual-purpose facilities need to be understood. A number of possible configurations for a system that transfers heat between the nuclear reactor and the hydrogen and/or electrical generation plants were identified. These configurations included both direct and indirect cycles for the production of electricity. Both helium and liquid salts were considered as the working fluid in the intermediate heat transport loop. Methods were developed to perform thermal-hydraulic and cycle-efficiency evaluations of the different configurations and coolants. The thermal-hydraulic evaluations estimated the sizes of various components in the intermediate heat transport loop for the different configurations. The relative sizes of components provide a relative indication of the capital cost associated with the various configurations. Estimates of the overall cycle efficiency of the various configurations were also determined. The evaluations determined which configurations and coolants are the most promising from thermalhydraulic and efficiency points of view.
Date: August 1, 2006
Creator: Oh, C.H.; Barner, R.; Davis, C. B.; Sherman, S. & Pickard, P.
Partner: UNT Libraries Government Documents Department

Thermal-Hydraulic Analyses of Heat Transfer Fluid Requirements and Characteristics for Coupling A Hydrogen Production Plant to a High-Temperature Nuclear Reactor

Description: The Department of Energy is investigating the use of high-temperature nuclear reactors to produce hydrogen using either thermochemical cycles or high-temperature electrolysis. Although the hydrogen production processes are in an early stage of development, coupling either of these processes to the hightemperature reactor requires both efficient heat transfer and adequate separation of the facilities to assure that off-normal events in the production facility do not impact the nuclear power plant. An intermediate heat transport loop will be required to separate the operations and safety functions of the nuclear and hydrogen plants. A next generation high-temperature reactor could be envisioned as a single-purpose facility that produces hydrogen or a dual-purpose facility that produces hydrogen and electricity. Early plants, such as the proposed Next Generation Nuclear Plant, may be dual-purpose facilities that demonstrate both hydrogen and efficient electrical generation. Later plants could be single-purpose facilities. At this stage of development, both single- and dual-purpose facilities need to be understood. Seven possible configurations for a system that transfers heat between the nuclear reactor and the hydrogen and/or electrical generation plants were identified. These configurations included both direct and indirect cycles for the production of electricity. Both helium and liquid salts were considered as the working fluid in the intermediate heat transport loop. Methods were developed to perform thermalhydraulic and cycle-efficiency evaluations of the different configurations and coolants. The thermalhydraulic evaluations estimated the sizes of various components in the intermediate heat transport loop for the different configurations. The relative sizes of components provide a relative indication of the capital cost associated with the various configurations. Estimates of the overall cycle efficiency of the various configurations were also determined. The evaluations determined which configurations and coolants are the most promising from thermal-hydraulic and efficiency points of view. These evaluations also determined which configurations and options ...
Date: June 1, 2005
Creator: Davis, C. B.; Oh, C. H.; Barner, R. B. & Wilson, D. F.
Partner: UNT Libraries Government Documents Department

Evaluation of operational safety at Babcock and Wilcox Plants: Volume 2, Thermal-hydraulic results

Description: The Nuclear Regulatory Commission has initiated a research program to develop a methodology to assess the operational performance of Babcock and Wilcox plants and to apply this methodology on a trial basis. The methodology developed for analyzing Babcock and Wilcox plants integrated methods used in both thermal-hydraulics and human factors and compared results with information used in the assessment of risk. The integrated methodology involved an evaluation of a selected plant for each pressurized water reactor vendor during a limited number of transients. A plant was selected to represent each vendor, and three transients were identified for analysis. The plants were Oconee Unit 1 for Babcock and Wilcox, H.B. Robinson Unit 2 for Westinghouse, and Calvert Cliffs Unit 1 for Combustion Engineering. The three transients were a complete loss of all feedwater, a small-break loss-of-coolant accident, and a steam-generator overfill with auxiliary feedwater. Included in the integrated methodology was an assessment of the thermal-hydraulic behavior, including event timing, of the plants during the three transients. Thermal-hydraulic results are presented in this volume (Volume 2) of the report. 26 refs., 30 figs., 7 tabs.
Date: November 1, 1987
Creator: Wheatley, P.D.; Davis, C.B.; Callow, R.A.; Fletcher, C.D.; Dobbe, C.A. & Beelman, R.J.
Partner: UNT Libraries Government Documents Department

Engineering Analysis of Intermediate Loop and Process Heat Exchanger Requirements to Include Configuration Analysis and Materials Needs

Description: The need to locate advanced hydrogen production facilities a finite distance away from a nuclear power source necessitates the need for an intermediate heat transport loop (IHTL). This IHTL must not only efficiently transport energy over distances up to 500 meters but must also be capable of operating at high temperatures (>850oC) for many years. High temperature, long term operation raises concerns of material strength, creep resistance and general material stability (corrosion resistance). IHTL design is currently in the initial stages. Many questions remain to be answered before intelligent design can begin. The report begins to look at some of the issues surrounding the main components of an IHTL. Specifically, a stress analysis of a compact heat exchanger design under expected operating conditions is reported. Also the results of a thermal analysis performed on two ITHL pipe configurations for different heat transport fluids are presented. The configurations consist of separate hot supply and cold return legs as well as annular design in which the hot fluid is carried in an inner pipe and the cold return fluids travels in the opposite direction in the annular space around the hot pipe. The effects of insulation configurations on pipe configuration performance are also reported. Finally, a simple analysis of two different process heat exchanger designs, one a tube in shell type and the other a compact or microchannel reactor are evaluated in light of catalyst requirements. Important insights into the critical areas of research and development are gained from these analyses, guiding the direction of future areas of research.
Date: September 1, 2005
Creator: Lillo, T.M.; Williamson, R.L.; Reed, T.R.; Davis, C.B. & Ginosar, D.M.
Partner: UNT Libraries Government Documents Department

Development of HyPEP, A Hydrogen Production Plant Efficiency Calculation Program

Description: The Department of Energy envisions the next generation very high temperature gas-cooled reactor (VHTR) as a single-purpose or dual-purpose facility that produces hydrogen and electricity. The Ministry of Science and Technology (MOST) of the Republic of Korea also selected VHTR for the Nuclear Hydrogen Development and Demonstration (NHDD) Project. The report will address the evaluation of hydrogen and electricity production cycle efficiencies for such systems as the VHTR and NHDD, and the optimization of system configurations. Optimization of such complex systems as VHTR and NHDD will require a large number of calculations involving a large number of operating parameter variations and many different system configurations. The research will produce (a) the HyPEP which is specifically designed to be an easy-to-use and fast running tool for the hydrogen and electricity production evaluation with flexible system layout, (b) thermal hydraulic calculations using reference design, (c) verification and validation of numerical tools used in this study, (d) transient analyses during start-up operation and off-normal operation. This project will also produce preliminary cost estimates of the major components.
Date: March 1, 2006
Creator: Oh, C. H.; Davis, C. B.; Sherman, S. R.; Vilim, S.; Lee, Y. J. & Lee, W. J.
Partner: UNT Libraries Government Documents Department

Simulation of three-dimensional hydrodynamic components with a one-dimensional transient analysis code

Description: The RELAP5 series of transient analysis codes was developed to provide the United States Nuclear Regulatory Commission with a fast-running and user convenient reactor analysis tool. Although it was developed primarily for best-estimate transient simulation of pressurized water reactors, it has been used to simulate a wide spectrum of hydraulic and thermal transients in both nuclear and non-nuclear systems involving steam-water-noncondensible fluid mixtures. In recent years it has also been applied to thermal-hydraulic analyses of various US Department of Energy production reactors. RELAP5 is a one-dimensional code, meaning that the basic field equations are solved only in the axial direction of a component. Thus, for example, only axial flow is calculated in a reactor vessel; radial and azimuthal flows are not considered. This has been a minor limitation of the code because most hydraulic situations in reactor systems can be modeled adequately with a one-dimensional code. In those situations where three-dimensional flows were anticipated, the TRAC-PF1 code has generally been used. (TRAC-PF1 has the capability to model three-dimensional components; however, that option is normally only used in the reactor vessel model.) This paper describes the RELAP5 hexagonal model of the SRS L-Reactor as well as comparisons of benchmark calculations with SRS data. Emphasis is placed on the multidimensional phenomena. 6 refs., 6 figs.
Date: January 1, 1990
Creator: Shaw, R.A. (Los Alamos National Lab., NM (USA)) & Davis, C.B. (EG and G Idaho, Inc., Idaho Falls, ID (USA))
Partner: UNT Libraries Government Documents Department

Benchmarking the RELAP5/MOD2. 5 r-. Theta. model of an SRS (Savannah River Site) reactor to the 1989 L-Reactor tests

Description: Benchmarking calculations utilizing RELAP5/MOD2.5 with a detailed multi-dimensional r-{theta} model of the SRS L-Reactor will be presented. This benchmarking effort has provided much insight into the two-component two-phase behavior of the reactor under isothermal conditions with large quantities of air ingested from the moderator tank to the external loops. Initial benchmarking results have illuminated several model weaknesses which will be discussed in conjunction with proposed modeling changes. The benchmarking work is being performed to provide a fully qualified RELAP5 model for use in computing the system response to a double ended large break LOCA. 5 refs., 14 figs.
Date: January 1, 1990
Creator: Bollinger, J.S. (Westinghouse Savannah River Co., Aiken, SC (USA)) & Davis, C.B. (EG and G Idaho, Inc., Idaho Falls, ID (USA))
Partner: UNT Libraries Government Documents Department

Developmental assessment of the multidimensional component in RELAP5 for Savannah River Site thermal hydraulic analysis

Description: This report documents ten developmental assessment problems which were used to test the multidimensional component in RELAP5/MOD2.5, Version 3w. The problems chosen were a rigid body rotation problem, a pure radial symmetric flow problem, an r-[theta] symmetric flow problem, a fall problem, a rest problem, a basic one-dimensional flow test problem, a gravity wave problem, a tank draining problem, a flow through the center problem, and coverage analysis using PIXIE. The multidimensional code calculations are compared to analytical solutions and one-dimensional code calculations. The discussion section of each problem contains information relative to the code's ability to simulate these problems.
Date: July 1, 1992
Creator: Hanson, R.G.; Johnson, E.C. (eds.); Carlson, K.E.; Chou, C.Y.; Davis, C.B.; Martin, R.P. et al.
Partner: UNT Libraries Government Documents Department