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HTGR safety philosophy

Description: The accident at the Three Mile Island has focused public attention on reactor safety. Many public figures advocate a safer method of generating nuclear electricity for the second nuclear era in the US. The paper discusses the safety philosophy of a concept deemed suitable for this second nuclear era. The HTGR, in the course of its evolution, included safety as a significant determinant in design philosophy. This is particularly evident in the design features which provide inherent safety. Inher… more
Date: August 1, 1980
Creator: Joskimovic, V. & Fisher, C.R.
Partner: UNT Libraries Government Documents Department
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Nuclear alkali metal Rankine power systems for space applications

Description: Nucler power systems utilizing alkali metal Rankine power conversion cycles offer the potential for high efficiency, lightweight space power plants. Conceptual design studies are being carried out for both direct and indirect cycle systems for steady state space power applications. A computational model has been developed for calculating the performance, size, and weight of these systems over a wide range of design parameters. The model is described briefly and results from parametric design st… more
Date: August 1, 1986
Creator: Moyers, J. C. & Holcomb, R. S.
Partner: UNT Libraries Government Documents Department
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Criticality experiments with mixed oxide fuel pin arrays in plutonium-uranium nitrate solution

Description: A series of critical experiments was completed with mixed plutonium-uranium solutions having a Pu/(Pu + U) ratio of approximately 0.22 in a boiler tube-type lattice assembly. These experiments were conducted as part of the Criticality Data Development Program between the United States Department of Energy (USDOE) and the Power Reactor and Nuclear Fuel Development Corporation (PNC) of Japan. A complete description of the experiments and data are included in this report. The experiments were perf… more
Date: August 1, 1988
Creator: Lloyd, R.C. (Pacific Northwest Lab., Richland, WA (United States)) & Smolen, G.R. (Oak Ridge National Lab., TN (United States))
Partner: UNT Libraries Government Documents Department
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Probabilistic risk assessment of HTGRs

Description: Probabilistic Risk Assessment methods have been applied to gas-cooled reactors for more than a decade and to HTGRs for more than six years in the programs sponsored by the US Department of Energy. Significant advancements to the development of PRA methodology in these programs are summarized as are the specific applications of the methods to HTGRs. Emphasis here is on PRA as a tool for evaluating HTGR design options. Current work and future directions are also discussed.
Date: August 1980
Creator: Fleming, K. N.; Houghton, W. J.; Hannaman, G. W. & Joksimovic, V.
Partner: UNT Libraries Government Documents Department
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Physics of reactor safety. Volume II. Quarterly report, April-June 1980

Description: The work in the Applied Physics Division includes reports on reactor safety modeling and assessment by members of the Reactor Safety Appraisals Section. Work on reactor core thermal-hydraulics is performed in ANL's Components Technology Division, emphasizing 3-dimensional code development for LMFBR accidents under natural convection conditions.
Date: August 1, 1980
Partner: UNT Libraries Government Documents Department
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High Flux Isotope Reactor. Quarterly report, October, November, and December 1979

Description: Routine reactor operation with four end-of-cycle shutdowns and one scheduled midcycle shutdown resulted in an on-stream time of 93.6% for the quarter. This gave the HFIR an on-stream time for the year of 91.3%. The outer control plates were replaced, and the annual core components inspection was made.
Date: August 1, 1980
Creator: Corbett, B.L. & Poteet, K.H.
Partner: UNT Libraries Government Documents Department
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Medium-size high-temperature gas-cooled reactor

Description: This report summarizes high-temperature gas-cooled reactor (HTGR) experience for the 40-MW(e) Peach Bottom Nuclear Generating Station of Philadelphia Electric Company and the 330-MW(e) Fort St. Vrain Nuclear Generating Station of the Public Service Company of Colorado. Both reactors are graphite moderated and helium cooled, operating at approx. 760/sup 0/C (1400/sup 0/F) and using the uranium/thorium fuel cycle. The plants have demonstrated the inherent safety characteristics, the low activatio… more
Date: August 1, 1980
Creator: Peinado, C.O. & Koutz, S.L.
Partner: UNT Libraries Government Documents Department
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User's guide for revised SPEC-4 neutron spectrum unfolding code

Description: The SPEC-4 computer code was developed in the United Kingdom to solve the spectrum unfolding problem for spherical gas-filled proton-recoil neutron spectrometers. This report describes the ORNL version of SPEC-4 which has been applied to the analysis of data from the Tower Shielding Facility. Recent modifications are described which largely pertain to the graphical output routines. In addition, the input requirements are presented in considerable detail including suggestions and recommendations… more
Date: August 1, 1980
Creator: Johnson, J.O. & Ingersoll, D.T.
Partner: UNT Libraries Government Documents Department
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Constitutive relations in TRAC-P1A

Description: The purpose of this document is to describe the basic thermal-hydraulic models and correlations that are in the TRAC-P1A code, as released in March 1979. It is divided into two parts, A and B. Part A describes the models in the three-dimensional vessel module of TRAC, whereas Part B focuses on the loop components that are treated by one-dimensional formulations. The report follows the format of the questions prepared by the Analysis Development Branch of USNRC and the questionnaire has been att… more
Date: August 1, 1980
Creator: Rohatgi, U.S. & Saha, P.
Partner: UNT Libraries Government Documents Department
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Sandia Sodium Purification Loop (SNAPL) description and operations manual

Description: Sandia's Sodium Purification Loop was constructed to purify sodium for fast reactor safety experiments. An oxide impurity of less than 10 parts per million is required by these in-pile experiments. Commercial, reactor grade sodium is purchased in 180 kg drums. The sodium is melted and transferred into the unit. The unit is of a loop design and purification is accomplished by ''cold trapping.'' Sodium purified in this loop has been chemically analysed at one part per million oxygen by weight. 5 … more
Date: August 1, 1985
Creator: Acton, R. U.; Weatherbee, R. L.; Smith, L. A.; Mastin, F. L. & Nowotny, K. E.
Partner: UNT Libraries Government Documents Department
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Radioactive materials released from nuclear power plants. Annual report, 1983. Volume 4

Description: Releases of radioactive materials in airborne and liquid effluents from commercial light water reactors during 1983 have been compiled and reported. Data on solid waste shipments as well as selected operating information have been included. This report supplements earlier annual reports issued by the former Atomic Energy Commission and the Nuclear Regulatory Commission. The 1983 release data are summarized in tabular form. Data covering specific radionuclides are summarized.
Date: August 1, 1986
Creator: Tichler, J. & Norden, K.
Partner: UNT Libraries Government Documents Department
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Light-Water-Reactor Safety Research Program: Quarterly Progress Report, January-March 1980

Description: This progress report summarizes the Argonne National Laboratory work performed during January, February, and March 1980 on water-reactor-safety problems. The research and development area covered is Transient Fuel Response and Fission-Product Release.
Date: August 1, 1980
Creator: Massey, Walter E. & Kyger, Jack A.
Partner: UNT Libraries Government Documents Department
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PWR loss-of-coolant accident analysis capability of the WRAP-EM system

Description: The modular computational system known as the Water Reactor Analysis Package (WRAP) has been extended to provide the computational tools required to perform a complete analysis of loss-of-coolant accidents (LOCAs) in pressurized water reactors (PWR). The new system is known as the WRAP-EM (Evaluation Model) system and will be used by NRC to interpret and evaluate reactor vendor EM methods and computed results. The system for PWR-EM analysis is comprised of several computer codes which have been… more
Date: August 1, 1980
Creator: Gregory, M. V. & Beranek, F.
Partner: UNT Libraries Government Documents Department
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National Low-Temperature Neutron-Irradiation Facility

Description: The Materials Sciences Division of the United States Department of Energy will establish a National Low Temperature Neutron Irradiation Facility (NLTNIF) which will utilize the Bulk Shielding Reactor (BSR) located at Oak Ridge National Laboratory. The facility will provide high radiation intensities and special environmental and testing conditions for qualified experiments at no cost to users. This report describes the planned experimental capabilities of the new facility.
Date: August 1, 1983
Creator: Coltman, R.R. Jr.; Klabunde, C.E. & Young, F.W. Jr.
Partner: UNT Libraries Government Documents Department
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Laboratory manual for salt-mixing test in 37- and 217-pin bundles. [LMFBR]

Description: This laboratory manual deals with the procedure employed during salt tracer experiments used in evaluating the hydraulic characteristics of a rod bundle. A description of the standard equipment used is given together with the details of manufacture of probes used for detecting the salt concentration. Details of the bundle construction have been excluded as they are availble in the reference cited. An attempt has been made to point out potential trouble areas and procedures.
Date: August 1, 1980
Creator: Chan, Y.N. & Todreas, N.E.
Partner: UNT Libraries Government Documents Department
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Determination of coefficients for the universal laws of friction and heat transfer for CFTL calculations

Description: The friction factor and Stanton number for flow past a roughened surface are determined by the parameters A and R(h/sup +/) of the universal law of friction and A/sub H/ and G(h/sup +/) of the universal law of heat transfer. The methods to be used for determination of these parameters for the particular roughness to be used in the Core Flow Test Loop (CFTL) are presented. Examples are given concerning the application of these methods to both transitional and fully rough flow using experimental … more
Date: August 1, 1980
Creator: Hodge, S.A.
Partner: UNT Libraries Government Documents Department
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Studies of axial-leakage simulations for homogeneous and heterogeneous EBR-II core configurations

Description: When calculations of flux are done in less than three dimensions, leakage-absorption cross sections are normally used to model leakages (flows) in the dimensions for which the flux is not calculated. Since the neutron flux is axially dependent, the leakages, and hence the leakage-absorption cross sections, are also axially dependent. Therefore, to obtain axial flux profiles (or reaction rates) for individual subassemblies, an XY-geometry calculation delineating each subassembly has to be done a… more
Date: August 1, 1985
Creator: Grimm, K. N. & Meneghetti, D.
Partner: UNT Libraries Government Documents Department
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SUPERENERGY-2: a multiassembly, steady-state computer code for LMFBR core thermal-hydraulic analysis

Description: Core thermal-hydraulic design and performance analyses for Liquid Metal Fast Breeder Reactors (LMFBRs) require repeated detailed multiassembly calculations to determine radial temperature profiles and subchannel outlet temperatures for various core configurations and subassembly structural analyses. At steady-state, detailed core-wide temperature profiles are required for core restraint calculations and subassembly structural analysis. In addition, sodium outlet temperatures are routinely neede… more
Date: August 1, 1980
Creator: Basehore, K.L. & Todreas, N.E.
Partner: UNT Libraries Government Documents Department
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State-of-practice review of ultrasonic in-service inspection of Class I system piping in commercial nuclear power plants

Description: The Pacific Northwest Laboratory conducted a survey to determine the state of practice of ultrasonic in-service inspection of primary system piping in light water reactors. Personnel at four utilities, five inspection organizations, and three domestic reactor manufacturers were interviewed. The intention of the study was to provide a better understanding of the actual practices employed in in-service inspection of primary system piping and of the difficulties encountered.
Date: August 1, 1982
Creator: Morris, C. J. & Becker, F. L.
Partner: UNT Libraries Government Documents Department
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Cladding corrosion and hydriding in irradiated defected zircaloy fuel rods

Description: Twenty-one LWBR irradiation test rods containing ThO/sub 2/-UO/sub 2/ fuel and Zircaloy cladding with holes or cracks operated successfully. Zircaloy cladding corrosion on the inside and outside diameter surfaces and hydrogen pickup in the cladding were measured. The observed outer surface Zircaloy cladding corrosion oxide thicknesses of the test rods were similar to thicknesses measured for nondefected irradiation test rods. An analysis model, which was developed to calculate outer surface oxi… more
Date: August 1985
Creator: Clayton, J. C.
Partner: UNT Libraries Government Documents Department
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Nuclear fuel assembly identification using computer vision

Description: A new method of identifying fuel assemblies has been developed. The method uses existing in-cell TV cameras to read the notch-coded handling sockets of Fast Flux Test Facility (FFTF) assemblies. A computer looks at the TV image, locates the notches, decodes the notch pattern, and produces the identification number. A TV camera is the only in-cell equipment required, thus avoiding complex mechanisms in the hot cell. Assemblies can be identified in any location where the handling socket is visibl… more
Date: August 1, 1985
Creator: Moffett, S. D.
Partner: UNT Libraries Government Documents Department
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Analysis of core damage frequency: Peach Bottom, Unit 2 internal events appendices

Description: This document contains the appendices for the accident sequence analysis of internally initiated events for the Peach Bottom, Unit 2 Nuclear Power Plant. This is one of the five plant analyses conducted as part of the NUREG-1150 effort for the Nuclear Regulatory Commission. The work performed and described here is an extensive reanalysis of that published in October 1986 as NUREG/CR-4550, Volume 4. It addresses comments from numerous reviewers and significant changes to the plant systems and pr… more
Date: August 1, 1989
Creator: Kolaczkowski, A.M.; Cramond, W.R.; Sype, T.T.; Maloney, K.J.; Wheeler, T.A.; Daniel, S.L. (Science Applications International Corp., Albuquerque, NM (USA) et al.
Partner: UNT Libraries Government Documents Department
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Fuel debris assessment for Three Mile Island Unit 2 (TMI-2) reactor recovery by gamma-ray and neutron dosimetry

Description: As a result of the accident on March 28, 1979, fuel debris was dispersed into the primary coolant system of the Three Mile Island Unit 2 (TMI-2) reactor. Location and quantification of fuel debris is essential for TMI-2 recovery. TMI-2 fuel debris assessments can be carried out nondestructively by neutron and gamma-ray dosimetry. Efforts to date have been directed toward fuel debris characterization of the makeup and purification demineralizers, will maintain reactor coolant water purity. Two h… more
Date: August 22, 1983
Creator: Gold, R.; Roberts, J. H.; McNeece, J. P.; Kaiser, B. J.; Ruddy, F. H.; Preston, C. C. et al.
Partner: UNT Libraries Government Documents Department
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