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in subjects: "Containment"

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LOFT system structural response during subcooled blowdown

The Loss-of-Fluid Test (LOFT) facility is a highly instrumented, pressurized water reactor test system designed to be representative of large pressurized water reactors (LPWRs) for the simulation of loss-of-coolant accidents (LOCAs). Detailed structural analysis and appropriate instrumentation (accelerometers and strain gages) on the LOFT system provided information for evaluation of the structural response of the LOFT facility for loss-of-coolant experiment (LOCE) induced loads. In general, th… more

Date: January 1, 1978

Creator: Martinell, J. S.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Seismic response analysis for structures with non-orthogonal modes

For seismic analysis of structures in nuclear industry, the spectrum method of modal combination acceptable to US Nuclear Regulatory Commission is applicable only to systems with orthogonal natural modes. When a structure with frequency dependent boundary conditions is set in motion, its natural modes of vibration are generally non-orthogonal. Based on such a structure, the spectrum method of modal composition is generalized to include systems with non-orthogonal modes. The generalized expressi… more

Date: January 1, 1980

Creator: Yau, W. F.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Tritium in fusion reactor components

When tritium is used in a fusion energy experiment or reactor, several implications affect and usually restrict the design and operation of the system and involve questions of containment, inventory, and radiation damage. Containment is expected to be particularly important both for high-temperature components and for those components that are prone to require frequent maintenance. Inventory is currently of major significance in cases where safety and environmental considerations limit the expe… more

Date: January 1, 1980

Creator: Watson, J S; Fisher, P W & Talbot, J B

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Fifth in situ vitrification engineering-scale test of simulated INEL buried waste sites

In September 1990, an engineering-scale in situ vitrification (ISV) test was conducted on sealed canisters containing a combined mixture of buried waste materials expected to be present at the Idaho National Engineering Laboratory (INEL) Subsurface Disposal Area (SDA). The test was part of a Pacific Northwest Laboratory (PNL) program to assist INEL in treatability studies of the potential application of ISV to mixed transuranic wastes at the INEL SDA. The purpose of this test was to determine t… more

Date: June 1, 1992

Creator: Bergsman, T.M.; Shade, J.W. (Pacific Northwest Lab., Richland, WA (United States)) & Farnsworth, R.K. (EG and G Idaho, Inc., Idaho Falls, ID (United States))

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Mark I 1/5-scale boiling water reactor pressure suppression experiment facility report

An accurate Mark I /sup 1///sub 5/-scale, boiling water reactor (BWR), pressure suppression facility was designed and constructed at Lawrence Livermore Laboratory (LLL) in 11 months. Twenty-seven air tests using the facility are described. Cost was minimized by utilizing equipment borrowed from other LLL programs. The total value of borrowed equipment exceeded the program's budget of $2,020,000. Substantial flexibility in the facility was used to permit independent variation in the drywell pres… more

Date: October 11, 1977

Creator: Altes, R.G.; Pitts, J.H.; Ingraham, R.F.; Collins, E.K. & McCauley, E.W.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Empirical tuff equation of state models

No Description Available.

Date: April 24, 1975

Creator: Wight, L. H.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Reactor containment analysis for BWR suppression systems. Progress report, April--October 1977

Work during the period included analytic studies of vent clearing and jet-slug dynamics and computer simulations (HEMP and CHAMP codes) of pressure suppression experiments.

Date: December 14, 1977

Creator: Edwards, L.L.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Introduction to nuclear test engineering

The basic information in this report is from a vu-graph presentation prepared to acquaint new or prospective employees with the Nuclear Test Engineering Division (NTED). Additional information has been added here to enhance a reader's understanding when reviewing the material after hearing the presentation, or in lieu of attending a presentation.

Date: July 15, 1982

Creator: O'Neal, W.C. & Paquette, D.L.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

The hole of a blanket tritium system on the fusion fuel cycle

The requirements of tritium technology are centered in three main areas, i.e., (1) fuel processing, (2) breeder tritium extraction, and (3) tritium containment. The gaseous tritium stream from the breeder tritium extraction system is significantly different from the plasma exhaust stream and, therefore, may have important impact on the operation of the fuel processing system. For some blankets, such as aqueous solution blanket, the blanket tritium stream may dominate the fuel processing system … more

Date: February 1988

Creator: Sze, D.K.; Finn, P.; Clemmer, R.; Anderson, J.; Bartlit, J.; Naruse, Y. et al.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Characterization of the solid, airborne materials created when UF/sub 6/ reacts with moist air flowing in single-pass mode

A series of experiments has been performed in which UF/sub 6/ was released into flowing air in order to characterize the solid particulate material produced under non-static conditions. In two of the experiments, the aerosol was allowed to stagnate in a static chamber after release and examined further but in the other experiments characterization was done only on material collected a few seconds after release. Transmission electron microscopy and mass measurement by cascaded impactor were used… more

Date: October 1, 1985

Creator: Pickrell, P. W.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Eulerian method for large-displacement fluid-structure interaction in reactor containments

An Eulerian method for analyzing large-displacement fluid-structure interaction in reactor containments is presented. The emphasis is on the development of a generalized hydrodynamic scheme to treat the irregular cells created by the movement of the structure with respect to the fixed Eulerian coordinates. A relaxation equation is derived from the boundary condition at the fluid-structure interface for the solution of the pressure at the interface. By combining this with the Poisson equation a … more

Date: January 1, 1977

Creator: Chang, Y. W. & Wang, C. Y.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Data report of a pretest analysis of soil-structure interaction and structural response in low-amplitude explosive testing (50 KG) of the heissdampfreaktor (HDR)

This report describes a three-dimensional nonlinear TRANAL finite element analysis of a nuclear reactor subjected to ground shaking from a buried 50 kg explosive source. The analysis is a pretest simulation of a test event which was scheduled to be conducted in West Germany on 3 November 1979.

Date: November 29, 1979

Creator: Vaughan, D.K.; Sandler, I.; Rubin, D.; Isenberg, J. & Nikooyeh, H.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Physical Model of Lean Suppression Pressure Oscillation Phenomena: Steam Condensation in the Light Water Reactor Pressure Suppression System (PSS)

Using the results of large scale multivent tests conducted by GKSS, a physical model of chugging is developed. The unique combination of accurate digital data and cinematic data has provided the derivation of a detailed, quantified correlation between the dynamic physical variables and the associated two-phase thermo-hydraulic phenomena occurring during lean suppression (chugging) phases of the loss-of-coolant accident in a boiling water reactor pressure suppression system.

Date: April 1, 1980

Creator: McCauley, E. W.; Holman, G. S.; Aust, E.; Schwan, H. & Vollbrandt, J.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Experimental results of direct containment heating by high-pressure melt ejection into the Surtsey vessel: The DCH-3 and DCH-4 tests

Two experiments, DCH-3 and DCH-4, were performed at the Surtsey test facility to investigate phenomena associated with a high-pressure melt ejection (HPME) reactor accident sequence resulting in direct containment heating (DCH). These experiments were performed using the same experimental apparatus with identical initial conditions, except that the Surtsey test vessel contained air in DCH-3 and argon in DCH-4. Inerting the vessel with argon eliminated chemical reactions between metallic debris … more

Date: August 1, 1991

Creator: Allen, M.D.; Pilch, M.; Brockmann, J.E.; Tarbell, W.W. (Sandia National Labs., Albuquerque, NM (United States)); Nichols, R.T. (Ktech Corp., Albuquerque, NM (United States)) & Sweet, D.W. (AEA Technology, Winfrith (United Kingdom))

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Fire protection countermeasures for containment ventilation systems

The goal of this project is to find countermeasures to protect High Efficiency Particulate Air (HEPA) filters, in exit ventilation ducts, from the heat and smoke generated by fire. Initially, methods were developed to cool fire-heated air by fine water spray upstream of the filters. It was recognized that smoke aerosol exposure to HEPA filters could also cause disruption of the containment system. Through testing and analysis, several methods to partially mitigate the smoke exposure to the HEPA… more

Date: August 25, 1980

Creator: Alvares, N.; Beason, D.; Bergman, V.; Creighton, J.; Ford, H. & Lipska, A.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Quarterly report on the strontium heat source development program, Advanced Nuclear Systems and Projects Division for April-June 1980

Oak Ridge National Laboratory completed metallographic examination of the metal specimens from the 30,000-h compatibility tests with /sup 90/SrF/sub 2/. Electron microprobe analysis of the specimens is now under way. Results show that chemical attack of the 30,000-h specimens was not much greater than that observed in the 6000- to 20,000-h tests. Work continues on qualification testing of the as-fabricated prototype outer capsules. As-fabricated Hastelloy S and Hastelloy C-4 outer capsules pass… more

Date: July 1, 1980

Creator: Fullam, H. T.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Comparison of aerosol behavior during sodium fires in CSTF with the HAA-3B code. [LMFBR]

Four large-scale tests using sodium fire aerosol sources have been carried out in the Containment System Test Facility (CSTF). Two of the tests employed pool fires and two used spray fires as the aerosol source. Because the CSTF containment vessel is approximately half-scale (20.3 m in height) of a typical reactor building, the CSTF results have provided a large-scale proof test of the HAA-3B Code. For the two pool fire tests, the measured and predicted airborne concentrations were in good agre… more

Date: March 1, 1980

Creator: Postma, A.K. & Owen, R.K.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Effects of a hypothetical loss-of-coolant accident on a Mark I Boiling Water Reactor pressure-suppression system

A loss-of-coolant accident (LOCA) in a boiling-water-reactor (BWR) power plant has never occurred. However, because this type of accident could be particularly severe, it is used as a principal theoretical basis for design. A series of consistent, versatile, and accurate air-water tests that simulate LOCA conditions has been completed on a /sup 1///sub 5/-scale Mark I BWR pressure-suppression system. Results from these tests are used to quantify the vertical-loading function and to study the as… more

Date: December 22, 1977

Creator: Pitts, J.H. & McCauley, E.W.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Identification and assessment of containment and release management strategies for a BWR Mark I containment

This report identifies and assesses accident management strategies which could be important for preventing containment failure and/or mitigating the release of fission products during a severe accident in a BWR plant with a Mark 1 type of containment. Based on information available from probabilistic risk assessments and other existing severe accident research, and using simplified containment and release event trees, the report identifies the challenges a Mark 1 containment could face during t… more

Date: September 1, 1991

Creator: Lin, C. C. & Lehner, J. R.

Item Type: Report

Partner: UNT Libraries Government Documents Department

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Severe accident testing of electrical penetration assemblies

This report describes the results of tests conducted on three different designs of full-size electrical penetration assemblies (EPAs) that are used in the containment buildings of nuclear power plants. The objective of the tests was to evaluate the behavior of the EPAs under simulated severe accident conditions using steam at elevated temperature and pressure. Leakage, temperature, and cable insulation resistance were monitored throughout the tests. Nuclear-qualified EPAs were produced from D. … more

Date: November 1, 1989

Creator: Clauss, D.B. (Sandia National Labs., Albuquerque, NM (USA))

Item Type: Report

Partner: UNT Libraries Government Documents Department

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Tritium permeability of structural materials and surface effects on permeation rates

Tritium management in any system always will include containment such that tritium release rates will be less than established limits and also will be as low as practical. The well known properties of hydrogen to permeate through most materials make the complete containment of tritium an impossible task. However, tritium release rates from a given system can be minimized by two primary efforts. First is the selection of a compatible containment material which frequently also will be the structu… more

Date: January 1, 1980

Creator: Bell, J T; Redman, J D & Bittner, H F

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Waste management units: Savannah River Site

This report indexes every waste management unit of the Savannah River Site. They are indexed by building number and name. The waste units are also tabulated by solid waste units receiving hazardous materials with a known release or no known release to the environment. It also contains information on the sites which has received no hazardous waste, and units which have received source, nuclear, or byproduct material only. (MB)

Date: September 1, 1991

Creator: Molen, G.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Supplemental Task A-2,A-2A, Design of robust waste cannister

The original design for High Level Nuclear Waste cannisters anticipated a dry and oxidizing environment and stable geological conditions. For these conditions, a welded light walled cannister of type 316L austenitic stainless steel was adequate. More recently, the possibility of geological activity, has been raised, and with it the occurrence of mechanical loading, water immersion, and reducing conditions. In order to meet these new design conditions, a multi barrier containment system is propo… more

Date: January 1, 1991

Creator: Skaggs, R.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

A database of information on technologies for hazardous waste site remediation

A personal-computer-based database and user interface has been developed for retrieving and reviewing information on technologies applicable to the environmental remediation of hazardous waste sites. This system and its information represent a useful source of technology information for people preparing, reviewing, and approving site remediation plans or evaluating remediation technologies. The system includes a variety of information for approximately 90 distinct remedial action technologies. … more

Date: April 1, 1992

Creator: Holter, G. M.; White, M. K. & Byrant, J. L.

Item Type: Article

Partner: UNT Libraries Government Documents Department