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Study on severe accident fuel dispersion behavior in the Advanced Neutron Source reactor at Oak Ridge National Laboratory

Description: Core flow blockage events are a leading contributor to core damage initiation risk in the Advanced Neutron Source (ANS) reactor. During such an accident, insufficient cooling of the fuel could result in core heatup and melting under full coolant flow condition. Coolant inertia forces acting on the melt surface would likely break up the melt into small particles. Under thermal-hydraulic conditions of ANS coolant channel, micro-fine melt particles are expected. Heat transfer between melt particle and coolant, which affects particle breakup, was studied. The study indicates that the thermal effect on melt fragmentation seems to be negligible because the time corresponding to the breakup due to hydrodynamic forces is much shorter than the time for the melt surface to solidify. The study included modeling and analyses to predict transient behavior and transport of debris particles throughout the coolant system. The transient model accounts for the surface forces acting on the particle that results from the pressure variation on the surface, inertia, virtual mass, viscous force due to relative motion of particle in the coolant, gravitation, and resistance due to inhomogenous coolant velocity radially across piping due to possible turbulent coolant motions. Results indicate that debris particles would reside longest in heat exchangers because of lower coolant velocity there. Also core debris tends to move together upon melting and entrainment.
Date: December 31, 1995
Creator: Kim, S.H.; Taleyarkhan, R.P.; Navarro-Valenti, S.; Georgevich, V. & Xiang, J.Y.
Partner: UNT Libraries Government Documents Department

Analysis of LMFBR primary system response to an HCDA using an Eulerian computer code

Description: Applications of an Eulerian code to predict the response of LMFBR containment and primary piping systems to hypothetical core disruptive accidents (HCDA), and to analyze sodium spillage problems, are described. The computer code is an expanded version of the ICECO code. Sample problems are presented for slug impact and sodium spillage, dynamics of the HCDA bubbles, and response of a piping loop. (JWR)
Date: January 1, 1975
Creator: Chang, Y.W.; Wang, C.Y.; Chu, H.Y.; Abdel-Moneim, M.T. & Gvildys, J.
Partner: UNT Libraries Government Documents Department

Recriticality considerations in LMFBR accidents

Description: From confercnce on fast rea ctor safety; Los Angeles, California, USA (2 Apr 1974). Recent studies suggest that fuel which is dispersed upward by a mild prompt critical burst may reenter the core region causing a secondary excuraion. Several modes of fuel reentry recriticality are examined, and phenomena which strongly affect reactivity ramp rate estimates are evaluated. A broad range of ramp rates results from these studies. The results also indicate that a number of effects expected to be present in realistic situations can strongly mitigate the expected reactivity insertion rates. (29 references) (auth)
Date: January 1, 1974
Creator: Boudreau, J.E. & Jackson, J.F.
Partner: UNT Libraries Government Documents Department

Individual plant examination program: Perspectives on reactor safety and plant performance. Part 6, appendices A, B, and C

Description: This report provides perspectives gained by reviewing 75 Individual Plant Examination (IPE) submittals pertaining to 108 nuclear power plant units. IPEs are probabilistic analyses that estimate the core damage frequency (CDF) and containment performance for accidents initiated by internal events (including internal flooding, but excluding internal fire). The U.S. Nuclear Regulatory Commission (NRC), Office of Nuclear Regulatory Research, reviewed the WE submittals with the objective of gaining perspectives in three major areas: (1) improvements made to individual plants as a result of their IPEs and the collective results of the IPE program, (2) plant-specific design and operational features and modeling assumptions that significantly affect the estimates of CDF and containment performance, and (3) strengths and weaknesses of the models and methods used in the IPEs. These perspectives are gained by assessing the core damage and containment performance results, including overall CDF, accident sequences, dominant contributions to component failure and human error, and containment failure modes. In particular, these results are assessed in relation to the design and operational characteristics of the various reactor and containment types, and by comparing the IPEs to probabilistic risk assessment characteristics. Methods, data, boundary conditions, and assumptions used in the IPEs are considered in understanding the differences and similarities observed among the various types of plants.
Date: December 1, 1997
Partner: UNT Libraries Government Documents Department

IPE results as compared with NUREG-1150

Description: In 1990, the NRC published NUREG-1150 which assessed the risks for five U.S. nuclear power plants. This paper provides a comparison of the results and perspectives obtained from the NUREG-1150 study to those obtained form the Individual Plant Examination (IPE) program. Specifically, results and perspectives on core damage frequency and containment performance are compared.
Date: December 31, 1995
Creator: Pratt, W.T.; Lehner, J.; Camp, A. & Chow, E.
Partner: UNT Libraries Government Documents Department

Improvement and verification of fast reactor safety analysis techniques

Description: An initial analysis of the KIWI-TNT experiment using the VENUS-II disassembly code has been completed. The calculated fission energy release agreed with the experimental value to within about 3 percent. An initial model for analyzing the SNAPTRAN-2 core disassembly experiment was also developed along with an appropriate equation-of-state. The first phase of the VENUS-II/PAD comparison study was completed through the issuing of a preliminary report describing the results. A new technique to calculate a P-V-work curve as a function of the degree of core expansion following a disassembly excursion has been developed. The technique provides results that are consistent with the ANL oxide-fuel equation-of-state in VENUS-II. Evaluation and check-out of this new model are currently in progress. (auth)
Date: January 1, 1975
Creator: Jackson, J.F.
Partner: UNT Libraries Government Documents Department

Accident Sequence Precursor Program Large Early Release Frequency Model Development

Description: The objectives for the ASP large early release frequency (LERF) model development work is to build a Level 2 containment response model that would capture all of the events necessary to define LERF as outlined in Regulatory Guide 1.174, can be directly interfaced with the existing Level 1 models, is technically correct, can be readily modified to incorporate new information or to represent another plant, and can be executed in SAPHIRE. The ASP LERF models being developed will meet these objectives while providing the NRC with the capability to independently assess the risk impact of plant-specific changes proposed by the utilities that change the nuclear power plants' licensing basis. Together with the ASP Level 1 models, the ASP LERF models provide the NRC with the capability of performing equipment and event assessments to determine their impact on a plant's LERF for internal events during power operation. In addition, the ASP LERF models are capable of being updated to reflect changes in information regarding the system operations and phenomenological events, and of being updated to assess the potential for early fatalities for each LERF sequence. As the ASP Level 1 models evolve to include more analysis capabilities, the LERF models will also be refined to reflect the appropriate level of detail needed to demonstrate the new capabilities. An approach was formulated for the development of detailed LERF models using the NUREG-1150 APET models as a guide. The modifications to the SAPHIRE computer code have allowed the development of these detailed models and the ability to analyze these models in a reasonable time. Ten reference LERF plant models, including six PWR models and four BWR models, which cover a wide variety of containment and nuclear steam supply systems designs, will be complete in 1999. These reference models will be used as the ...
Date: January 4, 1999
Creator: Brown, T.D.; Brownson, D.A.; Duran, F.A.; Gregory, J.J. & Rodrick, E.G.
Partner: UNT Libraries Government Documents Department

Analysis of core damage frequency: Nuclear power plant Dukovany, VVER/440 V-213 Unit 1, internal events. Volume 1: Main report

Description: This report presents the final results from the Level 1 probabilistic safety assessment (PSA) for the Dukovany VVER/440 V-213 nuclear power plant, Unit 1. Section 1.1 describes the objectives of this study. Section 1.2 discusses the approach that was used for completing the Dukovany PSA. Section 1.3 summarizes the results of the PSA. Section 1.4 provides a comparison of the results of the Dukovany PSA with the results of other PSAs for different types of reactors worldwide. Section 1.5 summarizes the conclusions of the Dukovany PSA.
Date: December 21, 1994
Creator: Pugila, W. J.
Partner: UNT Libraries Government Documents Department

Direct containment heating models in the CONTAIN code

Description: The potential exists in a nuclear reactor core melt severe accident for molten core debris to be dispersed under high pressure into the containment building. If this occurs, the set of phenomena that result in the transfer of energy to the containment atmosphere and its surroundings is referred to as direct containment heating (DCH). Because of the potential for DCH to lead to early containment failure, the U.S. Nuclear Regulatory Commission (USNRC) has sponsored an extensive research program consisting of experimental, analytical, and risk integration components. An important element of the analytical research has been the development and assessment of direct containment heating models in the CONTAIN code. This report documents the DCH models in the CONTAIN code. DCH models in CONTAIN for representing debris transport, trapping, chemical reactions, and heat transfer from debris to the containment atmosphere and surroundings are described. The descriptions include the governing equations and input instructions in CONTAIN unique to performing DCH calculations. Modifications made to the combustion models in CONTAIN for representing the combustion of DCH-produced and pre-existing hydrogen under DCH conditions are also described. Input table options for representing the discharge of debris from the RPV and the entrainment phase of the DCH process are also described. A sample calculation is presented to demonstrate the functionality of the models. The results show that reasonable behavior is obtained when the models are used to predict the sixth Zion geometry integral effects test at 1/10th scale.
Date: August 1, 1995
Creator: Washington, K.E. & Williams, D.C.
Partner: UNT Libraries Government Documents Department

Evaluation of potential severe accidents during low power and shutdown operations at Grand Gulf, Unit 1: Summary of results. Volume 1

Description: During 1989 the Nuclear Regulatory Commission (NRC) initiated an extensive program to examine the potential risks during low power and shutdown operations. Two plants, Surry and Grand Gulf, were selected as the plants to be studied by Brookhaven National Laboratory (Surry) and Sandia National Laboratories (Grand Gulf). This report documents the work performed during the analysis of the Grand Gulf plant. A phased approach was used for the overall study. In Phase 1, the objectives were to identify potential vulnerable plant configurations, to characterize (on a high, medium, or low basis) the potential core damage accident scenario frequencies and risks, and to provide a foundation for a detailed Phase 2 analysis. It was in Phase 1 that the concept of plant operational states (POSs) was developed to allow the analysts to better represent the plant as it transitions from power operation to nonpower operation than was possible with the traditional technical specification divisions of modes of operation. This phase consisted of a coarse screening analysis performed for all POSs, including seismic and internal fire and flood for some POSs. In Phase 2, POS 5 (approximately cold shutdown as defined by Grand Gulf Technical Specifications) during a refueling outage was selected as the plant configuration to be analyzed based on the results of the Phase 1 study. The scope of the Level 1 study includes plant damage state analysis and uncertainty analysis and is documented in a multi-volume NUREG/CR report (i.e., NUREG/CR-6143). The internal events analysis is documented in Volume 2. Internal fire and internal flood analyses are documented in Volumes 3 and 4, respectively. A separate study on seismic analysis, documented in Volume 5, was performed for the NRC by Future Resources Associates, Inc. The Level 2/3 study of the traditional internal events is documented in Volume 6, and a ...
Date: July 1, 1995
Creator: Whitehead, D. W.; Staple, B. D. & Daniel, S. L.
Partner: UNT Libraries Government Documents Department

Preliminary topical report on comparison reactor disassembly calculations

Description: Preliminary results of comparison disassembly calculations for a representative LMFBR model (2100-l voided core) and arbitrary accident conditions are described. The analytical methods employed were the computer programs: FX2- POOL, PAD, and VENUS-II. The calculated fission energy depositions are in good agreement, as are measures of the destructive potential of the excursions, kinetic energy, and work. However, in some cases the resulting fuel temperatures are substantially divergent. Differences in the fission energy deposition appear to be attributable to residual inconsistencies in specifying the comparison cases. In contrast, temperature discrepancies probably stem from basic differences in the energy partition models inherent in the codes. Although explanations of the discrepancies are being pursued, the preliminary results indicate that all three computational methods provide a consistent, global characterization of the contrived disassembly accident. (auth)
Date: November 1, 1975
Creator: McLaughlin, T.P.
Partner: UNT Libraries Government Documents Department

Status of accident analysis for fast breeder reactors

Description: There are still considerable deficiencies in computational tools available even for following accidents to initial disassembly. Present indications are that such a disassembly will be mild, without much sensitivity of this result to modeling assumptions, but the LOF-driven TOP requires better analysis. It is not possible at present to follow the course of an accident mechanistically beyond a first disassembly. Plugging and recriticality are possibilities, but so far do not appear likely to lead to a substantial increase in accident severity. It seems unlikely that the destructive work in an HCDA will be large enough to cause difficulties in containment. However, complete melting of the core material is a desirable assumption for PAHR analysis. It is not possible at present to predict what the relative upward and downward ejection of core material will be. (auth)
Date: November 1, 1975
Creator: Hummel, H.H.
Partner: UNT Libraries Government Documents Department

Implicit Eulerian method for analyzing transient phenomena in fast reactors

Description: An Eulerian compressible hydrodynamic method is presented for analyzing transient phenomena in nuclear reactors following hypothetical excursions. The method uses the implicit integration scheme to solve nonlinear equations of fluid dynamics in conjunction with a thin-shell analysis to calculate the response of the wall boundary. Detailed formulations are given. Results are presented for two example problems and compared with available experimental data. 6 references. (auth)
Date: October 1, 1975
Creator: Wang, C.Y.
Partner: UNT Libraries Government Documents Department

Development of thin shell equations for reactor subassembly dynamics

Description: By means of Hamilton's principle the elastic equations of motion for thin cylindrical shells are developed. Explicitly included in these equations are the possibilities of loading due to interior pressure pulses and to thermal stresses. A test problem is considered for the case of a uniform pressure pulse traveling along the axis; an exact analytic solution is compared with the results found by a finite-difference method. The two solutions agree within 1 percent; which agreement is satisfactory for present purposes.
Date: January 1, 1976
Creator: DeVault, G.P. & Blewett, P.J.
Partner: UNT Libraries Government Documents Department

Effect of cross-sectional buckling on the behavior of ACS support columns

Description: These analyses of the performance of the support columns for the above-core structures (ACS) have two principal aims: (1) to predict the forces exerted by the column in a hypothetical core-disruptive accident (HCDA) so that the motion of the ACS can be predicted in a coupled fluid-structure analysis, (2) to provide the strains and deformations of the columns so that situations which lead to complete failure of the support system can be identified. In previous studies, the columns were represented by beam elements so changes in the cross section could not be treated. While the columns in many designs are relatively thick-walled, scale-model tests performed at SRI indicate significant changes in the cross section. Therefore, models have been developed in which the portions of the column which undergo significant changes in cross section are modeled by plate elements. For the purpose of comparing the plate-beam models in the context of the loads expected in an HCDA, its predictions were compared to experimental results obtained in the SRI scale model tests. The solutions were obtained by the code SAFE/RAS; a new plate element was added to that program to perform these studies.
Date: January 1, 1983
Creator: Kennedy, J.M. & Belytschko, T.B.
Partner: UNT Libraries Government Documents Department

Study on severe accident fuel dispersion behavior in the Advanced Neutron Source reactor at Oak Ridge National Laboratory

Description: Core flow blockage events have been identified as a leading contributor to core damage initiation risk in the Advanced Neutron Source (ANS) reactor. During such an accident, insufficient cooling of the fuel in a few adjacent blocked coolant channels out of several hundred channels, could also result in core heatup and melting under full coolant flow condition in other coolant channels. Coolant inertia forces acting on the melt surface would likely break up the melt into small particles. Under thermal-hydraulic conditions of ANS coolant channel, micro-fine melt particles are expected. Heat transfer between melt particle and coolant, which affects the particle breakup characteristics, was studied. The study indicates that the thermal effect on melt fragmentation seems to be negligible because the time corresponding to the breakup due to hydrodynamic forces is much shorter than the time for the melt surface to solidify. The study included modeling and analyses to predict transient behavior and transport of debris particles throughout the coolant system. The transient model accounts for the surface forces acting on the particle that result from the pressure variation on the surface, inertia, virtual mass, viscous force due to the relative motion of the particle in the coolant, gravitation, and resistance due to inhomogeneous coolant velocity radially across piping due to expected turbulent coolant motions. The results indicate that debris particles would reside longest in the heat exchangers because of lower coolant velocity there. Also they are entrained and move together in a cloud.
Date: September 1, 1995
Creator: Kim, S.H.; Taleyarkhan, R.P.; Navarro-Valenti, S. & Georgevich, V.
Partner: UNT Libraries Government Documents Department