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Advanced nuclear reactor public opinion project

Description: This Interim Report summarizes the findings of our first twenty in-depth interviews in the Advanced Nuclear Reactor Public Opinion Project. We interviewed 6 industry trade association officials, 3 industry attorneys, 6 environmentalists/nuclear critics, 3 state officials, and 3 independent analysts. In addition, we have had numerous shorter discussions with various individuals concerned about nuclear power. The report is organized into the four categories proposed at our April, 1991, Advisory G… more
Date: July 25, 1991
Creator: Benson, B.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

LMFBR safety program. Quarterly technical progress report, October-December 1976

Description: Information related to sodium fires and fission products in LMFBR type reactors is presented concerning SOMIX code development; sodium jet dispersal tests; aerosol leakage; high temperature-concentration aerosol tests; aerosol source term size; and properties of high temperature fuel mixtures.
Date: February 25, 1977
Creator: Heisler, M. P.; Johnson, R. P.; Nelson, C. T.; Vaughan, E. U.; Guderjahn, C. & Eytel, L.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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HWMA/RCRA CLOSURE PLAN FOR THE MATERIALS TEST REACTOR WING (TRA-604) LABORATORY COMPONENTS VOLUNTARY CONSENT ORDER ACTION PLAN VCO-5.8 D REVISION2

Description: This Hazardous Waste Management Act/Resource Conservation and Recovery Act closure plan was developed for the laboratory components of the Test Reactor Area Catch Tank System (TRA-630) that are located in the Materials Test Reactor Wing (TRA-604) at the Reactor Technology Complex, Idaho National Laboratory Site, to meet a further milestone established under Voluntary Consent Order Action Plan VCO-5.8.d. The TRA-604 laboratory components addressed in this closure plan were deferred from the TRA-… more
Date: February 25, 2008
Creator: Winterholler, Kirk
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Advanced Simulation and Computing FY08-09 Implementation Plan Volume 2 Revision 0

Description: The Stockpile Stewardship Program (SSP) is a single, highly integrated technical program for maintaining the safety and reliability of the U.S. nuclear stockpile. The SSP uses past nuclear test data along with current and future nonnuclear test data, computational modeling and simulation, and experimental facilities to advance understanding of nuclear weapons. It includes stockpile surveillance, experimental research, development and engineering programs, and an appropriately scaled production … more
Date: April 25, 2007
Creator: McCoy, M; Kusnezov, D; Bikkel, T & Hopson, J
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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SSTAR: The U.S. Lead-Cooled Fast Reactor (LFR)

Description: It is widely recognized that the developing world is the next area for major energy demand growth, including demand for new and advanced nuclear energy systems. With limited existing industrial and grid infrastructures, there will be an important need for future nuclear energy systems that can provide small or moderate increments of electric power (10-700 MWe) on small or immature grids in developing nations. Most recently, the Global Nuclear Energy Partnership (GNEP) has identified, as one of … more
Date: September 25, 2007
Creator: Smith, C. F.; Halsey, W. G.; Brown, N. W.; Sienicki, J. J.; Moisseytsev, A. & Wade, D. C.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Study of safety relief valve operation under ATWS conditions. [Super critical flow]

Description: In March 1979, ETEC published as ETEC-TDR-78-19 a search which updated earlier NRC studies of saturated or subcooled water flow through relief valves, under ATWS conditions. This Supplement expands upon that search to include supercritical steam-water flow. No applicable data for the supercritical conditions was found, nor were any newer data on saturated or subcooled conditions uncovered. The Supplement also updated a look for facilities currently capable of simultaneously imposing all ATWS co… more
Date: July 25, 1979
Creator: Hutmacher, E. S. & Whitten, S. D.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Study of safety relief valve operation under ATWS conditions. [PWR]

Description: A literature survey and analysis project has been performed to determine if recent (since mid-1975) data has been reported which could influence the current approach to predicting PWR relief valve capacity under ATWS conditions. This study was conducted by the Energy Technology Engineering Center for NRC. Results indicate that the current relief valve capacity model tends to predict less capacity than actually obtains; however, no experimental verification at PWR ATWS conditions was found. Othe… more
Date: June 25, 1979
Creator: Hutmacher, E. S.; Nesmith, B. J. & Brukiewa, J. B.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Effect of torus wall flexibility on hydro-structural interaction in BWR containment system

Description: The MARK I boiling water reactor (BWR) containment system is comprised of a light-bulb-shaped reactor compartment connected through vent pipes to a torus-shaped and partially water-filled pressure suppression chamber, or the wetwell. During either a normally occurring safety relief valve (SRV) discharge or a hypothetical loss-of-coolant accident (LOCA), air or steam is forced into the wetwell water pool for condensation and results in hydrodynamically induced loads on the torus shell. An analyt… more
Date: April 25, 1979
Creator: Lu, S.C.H.; McCauley, E.W. & Holman, G.S.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Linear variable differential transformer and its uses for in-core fuel rod behavior measurements

Description: The linear variable differential transformer (LVDT) is an electromechanical transducer which produces an ac voltage proportional to the displacement of a movable ferromagnetic core. When the core is connected to the cladding of a nuclear fuel rod, it is capable of producing extremely accurate measurements of fuel rod elongation caused by thermal expansion. The LVDT is used in the Thermal Fuels Behavior Program at the U.S. Idaho National Engineering Laboratory (INEL) for measurements of nuclear … more
Date: June 25, 1979
Creator: Wolf, J.R.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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US NRC/LLL liaison with the Federal Republic of Germany for the GKSS-PSS steam condensation tests. Progress report No. 1. [BWR]

Description: This progress report for the USNRC/LLL Liaison program with the Federal Republic of Germany regarding boiling water reactor containment multivent steam condensation tests being conducted by GKSS (Gesellschaft fuer Kernenergieverwertung in Schiffbau und Schiffahrt) address program activity during the period of May-June 1979. During this period the program scope was defined, initial contacts between LLL and GKSS were established, and the first trip by an LLL representative to the GKSS test facili… more
Date: June 25, 1979
Creator: McCauley, E.W.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Instrumenting a pressure suppression experiment for a Mark I boiling water reactor: another measurements engineering challenge

Description: A /sup 1///sub 5/-scale test facility of a pressure-suppression system from a Mark I boiling water reactor was instrumented with seven types of transducers to obtain high-accuracy, dynamic loading data during a hypothetical loss-of-coolant accident. A total of 27 air tests have been completed with an average of 175 transducers recorded for each test. An end-to-end calibration of the total measurement system was run to establish accuracy of the data. The instrumentation verified the analysis of … more
Date: April 25, 1978
Creator: Shay, W.M.; Brough, W.G. & Miller, T.B.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Final safety analysis report for the Galileo mission: Volume 3 (Book 2), Nuclear risk analysis document: Appendices: Revision 1

Description: It is the purpose of the NRAD to provide an analysis of the range of potential consequences of accidents which have been identified that are associated with the launching and deployment of the Galileo mission spacecraft. The specific consequences analyzed are those associated with the possible release of radioactive material (fuel) of the Radioisotope Thermoelectric Generators (RTGs). They are in terms of radiation doses to people and areas of deposition of radioactive material. These consequen… more
Date: January 25, 1989
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

In-reactor corrosion of X-8001 and 1245 aluminum: Analysis of data from PT IP-42A

Description: Failures of standard fuel element jackets (1245 aluminum alloy) after relatively short Hanford reactor exposures are occurring at high reactor power levels. Metallographic examination of fuel element jackets from short exposure ruptures shows the cause of failure to be rapid intergranular corrosion. This type of attack has been reproduced in laboratory tests by exposure to water at temperatures above 200 C, or to steam at temperatures higher than 300 C. Flow disruptions due to misalignment of t… more
Date: June 25, 1958
Creator: Dillon, R. L. & Richman, R. B.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Production Test No. 105-522-E, Examination of pile process tubes

Description: The objective of the production test described in this report is to provide the authority and mechanism to remove process tubes from the piles for inspection. The test is intended to provide a more positive tube inspection program than was previously provided under production test number 105-9-P, ``Corrosion of Slugs and Tubes``, document number 7-3928.
Date: May 25, 1953
Creator: Falkoski, R. E.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Draft of physics sections to 100-N technical manual

Description: The information presented is a collection of most of the physics information available for the New Production Reactor. The details of some of the physics information, particularly those dealing with exposure and temperature effects, are by no means to be considered the final word since there has been no experimental verification of these effects. However, the gross physics characteristics described are felt to be reasonable representations of the expected physics behavior of N Reactor and shoul… more
Date: January 25, 1963
Creator: Nilson, R.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Recommendations to apply the ``square pile`` total control concept

Description: It is recommended that the ``square pile`` concept be adopted for all disaster total control calculations, and that the basic reactor constants listed in HW-62884, except for Ball 3X local strength at the DR Reactor, be used in applying this method. Curves are included for each reactor type, indicating allowable enrichment based on appropriate local control strengths. (The reactors whose operating methods are affected by disaster total control requirements are B, D, F, and DR Reactors; the rema… more
Date: February 25, 1960
Creator: Bowers, C. E.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Hazard analysis for 300 Area N Reactor Fuel Fabrication and Storage Facilty

Description: This hazard analysis (HA) has been prepared for the 300 Area N Reactor Fuel Fabrication and Storage Facility (Facility), in compliance with the requirements of Westinghouse Hanford Company (Westinghouse Hanford) controlled manual WHC-CM-4-46, Nonreactor Facility Safety Analysis Manual, and to the direction of WHC-IP-0690, Safety Analysis and Regulation Desk Instructions, (WHC 1992). An HA identifies potentially hazardous conditions in a facility and the associated potential accident scenarios. … more
Date: January 25, 1994
Creator: Johnson, D. J. & Brehm, J. R.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Loading and operating conditions for charge of five NIE-1 fuel tubes and two target assemblies in KER-1 or KER-2 Production Test IP-584-D, Supplement B

Description: Production Test IP-584-D, Supplement B authorized the irradiation of target assemblies in KER-1 and KER-2. Furthermore, the production test authorized that these target assemblies be irradiated with)other approved production tests; for this test, N Reactor inner fuel elements (NIE-1) will be irradiated. The production test required that specific operating conditions and loading patterns would be approved by the managers of Testing, Process and Reactor Development, and Process Technology Subsect… more
Date: May 25, 1964
Creator: Deobald, T. L.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Modified tubular support patterns: Ribbed and ribless process tubes

Description: Reactor support charge designs and patterns have been numerous and varied over the past few years, the basis for acceptance of a particular support loading being determined from a consideration of fabrication costs and the hydraulic characteristics of the proposed dummy load as it applies to a particular reactor. Since the first three or four dummy pieces downstream of the fuel column (and upstream of the column in a spline tube) are discarded after the irradiation process, it is desirable to u… more
Date: August 25, 1964
Creator: Angle, C. W.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Advanced nuclear reactor public opinion project. Interim report

Description: This Interim Report summarizes the findings of our first twenty in-depth interviews in the Advanced Nuclear Reactor Public Opinion Project. We interviewed 6 industry trade association officials, 3 industry attorneys, 6 environmentalists/nuclear critics, 3 state officials, and 3 independent analysts. In addition, we have had numerous shorter discussions with various individuals concerned about nuclear power. The report is organized into the four categories proposed at our April, 1991, Advisory G… more
Date: July 25, 1991
Creator: Benson, B.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Proposal for charging the fifth rupture fuel experiment: GEH-10, 34, 35

Description: The objective of this irradiation is to further verify the corrosion rate of tubular-type fuel elements under conditions of high specific power and central core temperatures. This fuel will be the inner tube only of an NPR fuel assembly. As in previous tests, this inner tube rupture will be used to further substantiate the rupture detection instrumentation that is being used in the development of the NPR. Previously unirradiated fuel will be used in this test. The reactor is to operate at full … more
Date: August 25, 1960
Creator: Call, R. L. & Kaulitz, D. C.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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