15 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

Studies of reactor containment : bi-monthly progress report, July to August 1962

Description: A report about 3 tasks which are undergone in this project. These tasks pertain to preparing computer codes for calculations of wave transmission in layered composites of materials, the performance of gas shock impingement experiments, and analyzing previously obtained experimental data using a liquid driven shock tube.
Date: 1962
Creator: Lieberman, P. & Zaker, T. A.
Partner: UNT Libraries Government Documents Department

Studies of reactor containment structures : bi-monthly progress report, July-August 1962

Description: A report that aims to resolve three problem areas. These solutions involve extending the incremental method to apply to the solution of cylinders with a rigid base and a hemispherical head, investigating shield-structure interaction, and the determination of the dynamic response of vessels with various end conditions.
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

Prestressed Concrete Reactor Vessel Model 1

Description: From introduction: The goal of engineers associated with nuclear power plants is the achievement of safe plants with low generating costs.One possible means of lowering costs is to increase the power generating capability for a single generating unit. To accomplish this, the sizes of nuclear reactors have been increased.
Date: October 25, 1966
Partner: UNT Libraries Government Documents Department

Effect of Radiation Damage on SM-1, SM-1A and PM-2A Reactor Vessels

Description: Report describing the status of the SM-1, SM-1A, and PM-2A reactors, specifically regarding the effects "of irradiation on nil-ductility transition temperature and the associated problem of brittle fracture." (p. iii)
Date: October 14, 1961
Creator: McLaughlin, D. W.; Rowekamp, B. J.; Chittum, R. A.; Coombe, J. R.; Kelleman, R. W.; Bobe, P. E. et al.
Partner: UNT Libraries Government Documents Department

TSOAK-M1 : a Computer Code to Determine Tritium Reaction/Adsorption/Release Parameters from Experimental Results of Air-Detritiation Tests

Description: A computer code has been developed which permits the determination of tritium reaction (T2 to HTO)/adsorption/release and instrument correction parameters from enclosure (building) - detritiation test data. The code is based on a simplified model which treats each parameter as a normalized time-independent constant throughout the data-unfolding steps. Because of the complicated four-dimensional mathematical surface generated by the resulting differential equation system, occasional local-minima effects are observed, but these effects can be overcome in most instances by selecting a series of trial guesses for the initial parameter values and observing the reproducibility of final parameter values for cases where the best overall fit to experimental data is achieved. The code was then used to analyze existing small-cubicle test data with good success, and the resulting normalized parameters were employed to evaluate hypothetical reactor-building detritiation scenarios. It was concluded from the latter evaluation that the complications associated with moisture formation, adsorption, and release, particularly in terms of extended cleanup times, may not be as great as was previously thought. It is recommended that the validity of the TSOAK-M1 model be tested using data from detritiation tests conducted on large experimental enclosures (5 to 10 cu cm) and, if possible, actual facility buildings.
Date: 1979?
Creator: Land, Robert H.; Land, Robert H.; Maroni, V. A. & Minkoff, Michael
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

Studies of reactor containment structures : bi-monthly progress report, September-October 1962

Description: A report about a project which aims to resolve three problem areas. The first problem pertains to a solution for a rigid-strain hardening material obeying the Mises yield condition, the second area relates to preparing the FORTRAN program for a solution of the simplified shield structure model, and the third area relates to computer programs for the response of spheres and infinite cylinders have been rewritten in FORTRAN.
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

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