2,094 Matching Results

Search Results

Advanced search parameters have been applied.

Supplement to Reactor Containment Design Study

Description: Summary: Report SL-1868 is a technical and economic feasibility study of four containment methods, i.e., standard containment, pressure relief, pressure supression and low pressure containment, for 44 MWe direct cycle and 180 and 300 MWe dual cycle boiling water reactor plants.
Date: April 23, 1962
Creator: Johnson, R. A. & Nelson, I.
Partner: UNT Libraries Government Documents Department

Reactor Containment Design Study

Description: Introduction: Sargent & Lundy was authorized on November 1, 1960, to make an economic and technical feasibility study of various reactor containment designs which are being utilized for several power plants now under construction.
Date: May 18, 1961
Creator: Johnson, R. A. & Nelson, I.
Partner: UNT Libraries Government Documents Department

REACTOR CONTAINMENT (INCLUDING A TECHNICAL PROGRESS REVIEW)

Description: An attempt is made to present available information pentinent to reactor containment. This is done directly, by summary and reference, or by reference alone. To provide a reference framework, the first review document must necessarily be handled differently from supplemental periodic reviews. The plan is to: (3) provide a detailed account of the problem and suggestions for work needed to yield adequate solutions; (2) present the accumulated knowledge and accomplishments; (3) give an account of experience in applying the containment concept; and (4) append extensive bibliographical material. An attempt is made in each case to indicate the significance of the information and its relation to the problems outlined. (A.C.)
Date: May 1, 1959
Creator: Brittan, R.O.
Partner: UNT Libraries Government Documents Department

Hanford prototype-barrier status report: FY 1995

Description: Surface barriers (or covers) have been proposed for use at the Hanford Site as a means to isolate certain waste sites that, for reasons of cost or worker safety or both, may not be exhumed. Surface barriers are intende to isolated the wastes from the accessible environment and to provide long-term protection to future populations that might use the Hanford Site. Currently, no ``proven`` long-term barrier system is available. For this reason, the Hanford Site Permanent Isolation Surface-Barrier Development Program (BDP) was organized to develop the technology needed to provide long-term surface barrier capability for the Hanford Site for the US Department of Energy (DOE). Designs have been proposed to meet the most stringent needs for long-term waste disposal. The objective of the current barrier design is to use natural materials to develop a protective barrier system that isolates wastes for at least 1000 years by limiting water, plant, animal, and human intrusion; and minimizing erosion. The design criteria for water drainage has been set at 0.5 mm/yr. While other design criteria are more qualitative, it is clear that waste isolation for an extended time is the prime objective of the design. Constructibility and performance. are issues that can be tested and dealt with by evaluating prototype designs prior to extensive construction and deployment of covers for waste sites at Hanford.
Date: November 1, 1995
Creator: Gee, G.W.; Ward, A.L.; Gilmore, B.G.; Ligotke, M.W. & Link, S.O.
Partner: UNT Libraries Government Documents Department

105-KE Basin isolation barrier leak rate test analytical development. Revision 1

Description: This document provides an analytical development in support of the proposed leak rate test of the 105-KE Basin. The analytical basis upon which the K-basin leak test results will be used to determine the basin leakage rates is developed in this report. The leakage of the K-Basin isolation barriers under postulated accident conditions will be determined from the test results. There are two fundamental flow regimes that may exist in the postulated K-Basin leakage: viscous laminar and turbulent flow. An analytical development is presented for each flow regime. The basic geometry and nomenclature of the postulated leak paths are denoted.
Date: May 9, 1995
Creator: Irwin, J. J.
Partner: UNT Libraries Government Documents Department

Studies of reactor containment structures : bi-monthly progress report, March 1, 1962 to February 28, 1962

Description: A report about analyzing the responses of containment vessels, the effect of openings on the strength of containment vessels, bulging experiments on pressurized cylinders of finite length, and the development of the containment handbook.
Date: 1962
Creator: Weil, N. A.; Chiapetta, R. L.; Costantino, C. J.; Hodge, Philip Gibson, 1920-; Morse, Stearns A. & Salmon, M. A.
Partner: UNT Libraries Government Documents Department

Instrumentation of a prestressed concrete containment vessel model

Description: A series of static overpressurization tests of scale models of nuclear containment structures is being conducted by Sandia National Laboratories for the Nuclear Power Engineering Corporation of Japan and the U.S. Nuclear Regulatory Commission. At present, two tests are being planned: a test of a model of a steel containment vessel (SCV) that is representative of an improved, boiling water reactor (BWR) Mark II design; and a test of a model of a prestressed concrete containment vessel (PCCV). This paper discusses plans and the results of a preliminary investigation of the instrumentation of the PCCV model. The instrumentation suite for this model will consist of approximately 2000 channels of data to record displacements, strains in the reinforcing steel, prestressing tendons, concrete, steel liner and liner anchors, as well as pressure and temperature. The instrumentation is being designed to monitor the response of the model during prestressing operations, during Structural Integrity and Integrated Leak Rate testing, and during test to failure of the model. Particular emphasis has been placed on instrumentation of the prestressing system in order to understand the behavior of the prestressing strands at design and beyond design pressure levels. Current plans are to place load cells at both ends of one third of the tendons in addition to placing strain measurement devices along the length of selected tendons. Strain measurements will be made using conventional bonded foil resistance gages and a wire resistance gage, known as a {open_quotes}Tensmeg{close_quotes}{reg_sign} gage, specifically designed for use with seven-wire strand. The results of preliminary tests of both types of gages, in the laboratory and in a simulated model configuration, are reported and plans for instrumentation of the model are discussed.
Date: September 1995
Creator: Hessheimer, M. F.; Rightley, M. J. & Matsumoto, T.
Partner: UNT Libraries Government Documents Department

Topological Galleries: A High Level User Interface for Topology Controlled Volume Rendering

Description: Existing topological interfaces to volume rendering are limited by their reliance on sophisticated knowledge of topology by the user. We extend previous work by describing topological galleries, an interface for novice users that is based on the design galleries approach. We report three contributions: an interface based on hierarchical thumbnail galleries to display the containment relationships between topologically identifiable features, the use of the pruning hierarchy instead of branch decomposition for contour tree simplification, and drag-and-drop transfer function assignment for individual components. Initial results suggest that this approach suffers from limitations due to rapid drop-off of feature size in the pruning hierarchy. We explore these limitations by providing statistics of feature size as function of depth in the pruning hierarchy of the contour tree.
Date: June 30, 2011
Creator: MacCarthy, Brian; Carr, Hamish & Weber, Gunther H.
Partner: UNT Libraries Government Documents Department

Analysis of main steam isolation valve leakage in design basis accidents using MELCOR 1.8.6 and RADTRAD.

Description: Analyses were performed using MELCOR and RADTRAD to investigate main steam isolation valve (MSIV) leakage behavior under design basis accident (DBA) loss-of-coolant (LOCA) conditions that are presumed to have led to a significant core melt accident. Dose to the control room, site boundary and LPZ are examined using both approaches described in current regulatory guidelines as well as analyses based on best estimate source term and system response. At issue is the current practice of using containment airborne aerosol concentrations as a surrogate for the in-vessel aerosol concentration that exists in the near vicinity of the MSIVs. This study finds current practice using the AST-based containment aerosol concentrations for assessing MSIV leakage is non-conservative and conceptually in error. A methodology is proposed that scales the containment aerosol concentration to the expected vessel concentration in order to preserve the simplified use of the AST in assessing containment performance under assumed DBA conditions. This correction is required during the first two hours of the accident while the gap and early in-vessel source terms are present. It is general practice to assume that at {approx}2hrs, recovery actions to reflood the core will have been successful and that further core damage can be avoided. The analyses performed in this study determine that, after two hours, assuming vessel reflooding has taken place, the containment aerosol concentration can then conservatively be used as the effective source to the leaking MSIV's. Recommendations are provided concerning typical aerosol removal coefficients that can be used in the RADTRAD code to predict source attenuation in the steam lines, and on robust methods of predicting MSIV leakage flows based on measured MSIV leakage performance.
Date: October 1, 2008
Creator: Salay, Michael (United States Nuclear Regulatory Commission, Washington, D.C.); Kalinich, Donald A.; Gauntt, Randall O. & Radel, Tracy E.
Partner: UNT Libraries Government Documents Department

Shipment of Small Quantities of Unspecified Radioactive Material in Chalfant Packagings

Description: In the post 6M era, radioactive materials package users are faced with the disciplined operations associated with use of Certified Type B packagings. Many DOE, commercial and academic programs have a requirement to ship and/or store small masses of poorly characterized or unspecified radioactive material. For quantities which are small enough to be fissile exempt and have low radiation levels, the materials could be transported in a package which provides the required containment level. Because their Chalfant type containment vessels meet the highest standard of containment (helium leak-tight), the 9975, 9977, and 9978 are capable of transporting any of these contents. The issues associated with certification of a high-integrity, general purpose package for shipping small quantities of unspecified radioactive material are discussed and certification of the packages for this mission is recommended.
Date: June 12, 2009
Creator: Smith, Allen; Abramczyk, Glenn; Nathan, Steven & Bellamy, Steve
Partner: UNT Libraries Government Documents Department

Wildfire Suppression Spending: Background, Issues, and Legislation in the 115th Congress

Description: This report provides background information and analysis of federal funding for wildfire suppression operations on federal lands or for fires that began on federal land. The report provides a discussion of the issues facing Congress and concludes by summarizing several legislative proposals under consideration by the 115th Congress.
Date: October 5, 2017
Creator: Hoover, Katie & Lindsay, Bruce R.
Partner: UNT Libraries Government Documents Department

Sample push out fixture

Description: This invention generally relates to the remote removal of pelletized samples from cylindrical containment capsules. V-blocks are used to receive the samples and provide guidance to push out rods. Stainless steel liners fit into the v-channels on the v-blocks which permits them to be remotely removed and replaced or cleaned to prevent cross contamination between capsules and samples. A capsule holder securely holds the capsule while allowing manual up/down and in/out movement to align each sample hole with the v-blocks. Both end sections contain identical v-blocks; one that guides the drive out screw and rods or manual push out rods and the other to receive the samples as they are driven out of the capsule.
Date: February 22, 2000
Creator: Biernat, John L.
Partner: UNT Libraries Government Documents Department

Surface Environmental Surveillance Procedures Manual

Description: Environmental surveillance data are used in assessing the impact of current and past site operations on human health and the environment, demonstrating compliance with applicable local, state, and federal environmental regulations, and verifying the adequacy of containment and effluent controls. SESP sampling schedules are reviewed, revised, and published each calendar year in the Hanford Site Environmental Surveillance Master Sampling Schedule. Environmental samples are collected by SESP staff in accordance with the approved sample collection procedures documented in this manual.
Date: September 20, 2000
Creator: Hanf, RW & Poston, TM
Partner: UNT Libraries Government Documents Department

Technical Assistance in Review of Source Term-Related Issues of Advanced Reactors

Description: The distribution of iodine in containment during an AP-600 design-basis accident was evaluated using models in the "TRENDS" code. The AP-6003BE accident sequmce calculations showed that a pH >7 was maintained for at least 30 days. Because the pH was maintained at this level, most of the iodine was in the form of iodid~ only 3 x 10-3% was present as aqueous 12, and only 1 x 10< `/0 was present as J in the vapor phase.
Date: October 1, 1998
Creator: Beahm, E.C.; Dillow, T.A. & Weber, C.F.
Partner: UNT Libraries Government Documents Department