Filters
Results 1 - 10 of 17
Results 1 - 10 of 17.
Search took: 0.049 seconds
Sort by: date | relevance |
Yung, S.C.
Hanford Engineering Development Lab., Richland, WA (USA)1982
Hanford Engineering Development Lab., Richland, WA (USA)1982
AbstractAbstract
[en] The modeling of sodium pool fires constitutes an important ingredient in conducting LMFBR accident analysis. Such modeling capability has recently come under scrutiny at Westinghouse Hanford Company (WHC) within the context of developing CONACS, the Containment Analysis Code System. One of the efforts in the CONACS program is to model various combustion processes anticipated to occur during postulated accident paths. This effort includes the selection or modification of an existing model and development of a new model if it clearly contributes to the program purpose. As part of this effort, a new sodium pool fire model has been developed that is directed at removing some of the deficiencies in the existing models, such as SOFIRE-II and FEUNA
Original Title
LMFBR
Primary Subject
Secondary Subject
Source
19 Oct 1982; 13 p; American Nuclear Society winter meeting; Washington, DC (USA); 14-19 Nov 1982; CONF-821103--93; Available from NTIS, PC A02/MF A01 as DE83012153
Record Type
Report
Literature Type
Conference
Report Number
Country of publication
Reference NumberReference Number
INIS VolumeINIS Volume
INIS IssueINIS Issue
Yung, S.C.
Hanford Engineering Development Lab., Richland, WA (USA)1979
Hanford Engineering Development Lab., Richland, WA (USA)1979
AbstractAbstract
[en] The effects of different transient power histories on the intrasubassembly failure incoherencies in an unprotected Transient Overpower (TOP) Hypothetical Core Disruptive Accident (HCDA) are analyzed. The computational tool and other relevant points of present analysis are described. The computational results from the analysis are also discussed. The summary of the present study is stated
Primary Subject
Secondary Subject
Source
Sep 1979; 51 p; American Nuclear Society meeting; San Francisco, CA, USA; 12 - 16 Nov 1979; CONF-791103--64; Available from NTIS., PC A04/MF A01
Record Type
Report
Literature Type
Conference
Report Number
Country of publication
Reference NumberReference Number
INIS VolumeINIS Volume
INIS IssueINIS Issue
Yung, S.C.; Wilburn, N.P.
Hanford Engineering Development Lab., Richland, WA (USA)1979
Hanford Engineering Development Lab., Richland, WA (USA)1979
AbstractAbstract
[en] This study addresses the survival of peripheral pins within an LMFBR subassembly during an unprotected Transient Overpower (TOP) Hypothetical Core Disruptive Accident (HCDA) by consideration of intrasubassembly incoherencies. A continuous analysis was made vs time from the initiation of the accident up to the point where the power decreases to a quasi-steady state for a Beginning-of-Cycle-4 (BOC-4) core of Fast Test Reactor (FTR), 0.5$/sec ramp case. Blockage was assumed to be formed after fuel pin's failure and effects due to blockages were examined. The study concludes that most peripheral pins within an LMFBR subassembly indeed will survive
Original Title
LMFBR
Primary Subject
Secondary Subject
Source
Jul 1979; 19 p; International meeting on fast reactor safety technology; Seattle, WA, USA; 19 - 23 Aug 1979; CONF-790816--65; Available from NTIS., PC A02/MF A01
Record Type
Report
Literature Type
Conference
Report Number
Country of publication
Reference NumberReference Number
INIS VolumeINIS Volume
INIS IssueINIS Issue
Bian, S.H.; Budden, M.J.; Yung, S.C.
Pacific Northwest Lab., Richland, WA (USA); Rockwell International Corp., Richland, WA (USA). Rockwell Hanford Operations1987
Pacific Northwest Lab., Richland, WA (USA); Rockwell International Corp., Richland, WA (USA). Rockwell Hanford Operations1987
AbstractAbstract
[en] The GEOTHER/VT4 code is a modified and improved version of the GEOTHER code. It was applied to a two-dimensional simulation of a single waste package container in a high-level waste repository to predict the thermal-hydraulic environment where steam formation may occur. The groundwater and thermal conditions are important for waste package container corrosion, packing material swelling tests, and for evaluation of the near-field geochemical conditions. The waste package was assumed to be situated in the Cohassett flow of the Hanford Washington Site bounded by the flow top and flow bottom. The calculation indicates that the maximum steam formation occurs at about 10 years after waste package emplacement. The two-phase (steam and water) zone extends about 0.5 m above and below the waste package surface. After this period, the saturation profile stays essentially unchanged until 50 years after container emplacement. Then the two-phase zone condenses until resaturation at about 62 years after container emplacement
Primary Subject
Source
May 1987; 18 p; International conference on groundwater contamination; Amsterdam (Netherlands); 26-29 Oct 1987; CONF-8710141--1; Available from NTIS, PC A02/MF A01; 1 as DE88000086; Portions of this document are illegible in microfiche products.
Record Type
Report
Literature Type
Conference
Report Number
Country of publication
Reference NumberReference Number
INIS VolumeINIS Volume
INIS IssueINIS Issue
Armstrong, G.R.; Yung, S.C.; Schenter, R.E.
Hanford Engineering Development Lab., Richland, WA (USA)1986
Hanford Engineering Development Lab., Richland, WA (USA)1986
AbstractAbstract
[en] The purpose of this paper is to study the implications of more correctly treating the impact of decay heat on concrete
Primary Subject
Secondary Subject
Source
Mar 1986; 7 p; Conference on the science and technology of fast reactor safety; Channel Islands (UK); 12-16 May 1986; CONF-860501--16; Available from NTIS, PC A02/MF A01; 1 as DE86015607; Portions of this document are illegible in microfiche products.
Record Type
Report
Literature Type
Conference
Report Number
Country of publication
Reference NumberReference Number
INIS VolumeINIS Volume
INIS IssueINIS Issue
Bian, S.H.; Budden, M.J.; Bartley, C.L.; Yung, S.C.
Pacific Northwest Lab., Richland, WA (USA); Westinghouse Hanford Co., Richland, WA (USA)1988
Pacific Northwest Lab., Richland, WA (USA); Westinghouse Hanford Co., Richland, WA (USA)1988
AbstractAbstract
[en] The objective of the work is to evaluate the GEOTHER code and peform necessary improvements to make it specifically suitable for predicting the environmental conditions of the waste package for the Basalt Waste Isolation Project (BWIP); and to perform resaturation analyses, that is, the analyses of steam formation and condensation, for the repository and waste package using the improved GEOTHER code. This is a progress report to BWIP documenting the status of GEOTHER code testing, evaluation, and improvements. The computational results documented in this report reflect the current condition of the code and the condition before code improvements. The test cases used are intended for examining the code features in sufficient detail and are not intended to be taken as final conclusions for BWIP applications
Primary Subject
Secondary Subject
Source
Mar 1988; 391 p; Available from NTIS, PC A17/MF A01; 1 as DE88007620; Portions of this document are illegible in microfiche products.
Record Type
Report
Literature Type
Progress Report
Report Number
Country of publication
Reference NumberReference Number
INIS VolumeINIS Volume
INIS IssueINIS Issue
Bian, S.H.; Budden, M.J.; Bartley, C.L.; Yung, S.C.
Pacific Northwest Lab., Richland, WA (USA); Westinghouse Hanford Co., Richland, WA (USA)1988
Pacific Northwest Lab., Richland, WA (USA); Westinghouse Hanford Co., Richland, WA (USA)1988
AbstractAbstract
[en] The GEOTHER/VT4 code has been developed at Pacific Northwest Laboratory for the Basalt Waste Isolation Project (BWIP). This code is a modified version of the GEOTHER code developed by the US Geological Survey and later modified by Battelle's Office of Nuclear Waste Isolation (ONWI) for nuclear waste repository simulation. The two-equation model of the original GEOTHER was modified by adding a conduction equation to the model. Other changes were made to the code to make it suitable for simulation of waste repositories. This report gives the detailed derivation of the three-equation model, the numerical solution method, code verification, and input description. Input listings for the benchmark cases used to verify the code are presented. The twelve new subroutines added to the code are also described. These descriptions are followed by a sample output, a discussion of graphics programs for the code, program redimensioning, and bit packing. The current version is suitable only for an environment where noncondensable gases are absent. An improved version is under development to account for the noncondensable gases. 13 refs., 8 figs., 8 tabs
Primary Subject
Secondary Subject
Source
Mar 1988; 175 p; Available from NTIS, PC A08/MF A01; 1 as DE88008436
Record Type
Report
Report Number
Country of publication
Reference NumberReference Number
INIS VolumeINIS Volume
INIS IssueINIS Issue
AbstractAbstract
No abstract available
Original Title
LMFBR
Primary Subject
Secondary Subject
Source
American Nuclear Society meeting; San Francisco, CA, USA; 12 - 16 Nov 1979; CONF-791103--; Published in summary form only.
Record Type
Journal Article
Literature Type
Conference
Journal
Transactions of the American Nuclear Society; ISSN 0003-018X; ; v. 33 p. 536-538
Country of publication
Reference NumberReference Number
INIS VolumeINIS Volume
INIS IssueINIS Issue
AbstractAbstract
No abstract available
Primary Subject
Secondary Subject
Source
ANS winter meeting; San Francisco, CA, USA; 27 Nov 1977; See CONF-771109--. Published in summary form only.
Record Type
Journal Article
Literature Type
Conference
Journal
Transactions of the American Nuclear Society; v. 27 p. 536-537
Country of publication
Reference NumberReference Number
INIS VolumeINIS Volume
INIS IssueINIS Issue
AbstractAbstract
No abstract available
Primary Subject
Secondary Subject
Source
ANS annual meeting; San Diego, CA, USA; 18 Jun 1978; See CONF-780622--. Published in summary form only.
Record Type
Journal Article
Literature Type
Conference
Journal
Transactions of the American Nuclear Society; v. 28 p. 475-476
Country of publication
Reference NumberReference Number
INIS VolumeINIS Volume
INIS IssueINIS Issue
1 | 2 | Next |