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AbstractAbstract
[en] Thirty-three papers were presented at this meeting on a) post accident heat removal research requirements; b) formation, cooling and remelting of particulate debris bed; c) molten pool cooling; d) melt front phenomena; e) materials, thermophysical properties and compatibility; f) PAHR system design concepts
Primary Subject
Source
4. Post-accident heat removal information exchange meeting; Varese, Italy; 10 - 12 Oct 1978; EUR--6595; Microfiche available from CEC, BP No. 1003, Luxembourg
Record Type
Journal Article
Literature Type
Conference
Journal
European Applied Research Reports. Nuclear Science and Technology Section; ISSN 0379-4229;
; v. 1(6); p. 1315-1712

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AbstractAbstract
[en] This EG and G Idaho, Inc. report reviews the technical specifications to establish the redundancy and the diversity of systems available for the removal of decay heat at the Kewaunee Nuclear Power Plant
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Source
Apr 1982; 22 p; Available from NTIS., PC A02/MF A01 as DE82015455
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Report
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AbstractAbstract
[en] The American National Standard for Decay Heat Power in Light Water Resources (ANSI/ANS-5.1) was issued in 1979 and has been used extensively for the past decade. Since the standard was issued, there have been new decay-heat measurements, and new and improved calculational capabilities have been developed. In this article the standard is compared with these new data and with calculations based on improved methods. Three foreign standards or proposed standards have also been developed, and their contents are compared with the present American standard. Proposals for improving the standard are presented and discussed. 27 refs., 9 figs., 3 tabs
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Secondary Subject
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Journal Article
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Chi, H.W.; Chung, H.S.; Shenoy, A.
General Atomic Co., San Diego, CA (USA)1980
General Atomic Co., San Diego, CA (USA)1980
AbstractAbstract
[en] The residual heat removal (RHR) capability for providing lines of protection (LOPS) 1 and 2 of the gas-cooled fast breeder reactor (GCFR) demonstration plant is described. Included are design criteria and system descriptions for the RHR cooling systems and the portion of the plant protection system that is related to initiation of the RHR system operation. The design features of these systems provide inherently redundant and diverse means of core cooling for the GCFR. The hierarchy in the selection of the RHR systems and the application of the systems to key transient events are discussed. Methods of RHR system operation, dynamic responses of the GCFR plant, and margins of safety in RHR operations are also presented
Primary Subject
Source
May 1980; 31 p; GCFRP program technical review meeting; San Diego, CA, USA; 4 - 6 Jun 1980; CONF-800648--6; Available from NTIS., PC A03/MF A01
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Report
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Conference
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AbstractAbstract
No abstract available
Primary Subject
Source
Israel Nuclear Society, Yavne; Israel Health Physics Society; Radiation Research Society of Israel; Israel Society of Medical Physics; Transactions; v. 11; 301 p; 1983; p.61; Nuclear Societies of Israel joint annual meeting; Haifa (Israel); 21-22 Dec 1983; Published in summary form only.
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Miscellaneous
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AbstractAbstract
[en] Sample calculations were done for KORI-1 to develop a better understanding of what happens after very small LOCA (< or approx.0.05 ftsup(2)). For a water-side break with the break size larger than 0.006ftsup(2), fluid-loss through break exceeds the makeup. If the break sizeis larger than 0.008ftsup(2), decay heat can be completely removed through break. Based on these results, it was concluded that KORI-1 is fairly safe for the whole spectrum of sizes in very small LOCA. However, for the reactor with 9000MWe or 1200MWe, a certain spectrum of sizes in very small LOCA should be carefully considered. In the accident sequence the transition from natural circulation to pool boiling or from pool boiling to natural circulation may be troublesome to the operator or in the safety analysis. Operator's intervension was discussed; primary pump shutoff, HPI pump shutoff, break isolation, and opening relief valve. It was proved that continuous operation of HPI pumps after shutdown will not threaten the integrity of the primary system. (Author)
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Record Type
Journal Article
Journal
Journal of the Korean Nuclear Society; ISSN 0372-7327;
; v. 16(1); p. 11-17

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Kayser, G.; Abdon, R.; Arslan, M.; Bergeonneau, P.; Gesi, E.
International meeting on fast reactor safety technology1979
International meeting on fast reactor safety technology1979
AbstractAbstract
[en] The studies are related to a commercial pool-type LMFBR of the SUPER PHENIX type. The most important sequences result from a non-equilibrium between power and heat removal associated with a total lack of rod drop. It has been shown that this reactor type presents a good safety against the first type of melt, that is by sodium boiling, due to the important neutronic feed-back which diminishes the neutronic power as temperature rises. 2 refs
Primary Subject
Source
Anon; p. 1932-1941; 1979; p. 1932-1941; American Nuclear Society; LaGrange Park, IL; International meeting on fast reactor safety technology; Seattle, WA, USA; 19 - 23 Aug 1979
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Book
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Conference
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Hwang, I. G.; Lee, D. Y.; Cha, K. H.; Park, J. C.; Sim, Y. S.
Proceedings of the KNS spring meeting2000
Proceedings of the KNS spring meeting2000
AbstractAbstract
[en] In process instrumentation systems of such as nuclear plants, response time information is very important in most temperature transient measurements. Generally the response time of thermocouples is measured at a laboratory by using a plunge method. However, it is not easy to use the plunge testing method when a response time measurement of an installed thermocouple is required. A measurement system was developed to measure the response time of a thermocouple installed in a process by using the Loop Current Step Response(LCSR) testing method. This device heats a thermocouple by providing an electrical current, and then it measures the thermocouple output as the temperature of the thermocouple measurement junction returns to ambient temperature. The time constant of the thermocouple is determined from the transient curve of the thermocouple output indicating the temperature difference between the reference junction and measurement junction of the thermocouple. The device is designed to heat a middle point to reduce the temperature error caused by residual heat of thermocouple wire
Primary Subject
Source
Korean Nuclear Society, Taejon (Korea, Republic of); [CDROM]; May 2000; [9 p.]; 2000 spring meeting of the KNS; Kori (Korea, Republic of); 26-27 May 2000; Available from KNS, Taejon (KR); 8 refs, 9 figs, 1 tab
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Miscellaneous
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Torri, A.; Taniguchi, T.; Raabe, P.H.
General Atomic Co., San Diego, CA (USA)1979
General Atomic Co., San Diego, CA (USA)1979
AbstractAbstract
[en] Reliability of decay heat removal is an important safety consideration in the gas-cooled fast breeder reactor (GCFR). The design evolution of the residual heat removal (RHR) systems over the past few years has been markedly aided by system reliability analyses to the point where there is confidence that loss of coolable core geometry can be classified as a beyond-design-basis accident. This evolution proceeded in three steps. First, the reliability-limiting features in the total combination of RHR systems were investigated and the need for improvements in the reliability of the main loop cooling system for RHR as well as in the physical separation of RHR support systems between the main loops and the core auxiliary cooling systems (CACS) was established. Secondly, a wide range of RHR options for the main loop cooling system were investigated resulting in the adoption of a new reference concept for the main loop RHR system. The third and last step then consisted of an evaluation of the reliability aspects of natural circulation decay heat removal in an upflow GCFR design. The major conclusion from this study is that decay heat removal can be reliable in the GCFR. Furthermore, the current limitations of natural circulation RHR reliability have been identified, and means to optimally exploit natural circulation have been defined
Primary Subject
Source
Jul 1979; 21 p; Helium Breeder Associates/US Dept of Energy GCFR technical meeting; San Diego, CA, USA; 30 May - 1 Jun 1979; CONF-790572--8; CONF-790384--3; Available from NTIS., PC A02/MF A01
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Report
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AbstractAbstract
[en] 16 improved main loop residual heat removal options were identified and analyzed for their potential to improve heat removal train reliability. Ten shutdowns and three reactor trips per year with a plant availability data of 80% formed the basis of the analysis. The results of quantitative reliability analyses using the evaluated reliability data bank for gas-cooled reactors showed that several of the systems analyzed had the potential for substantial improvements in the main cooling system reliability. On the basis of this work, a new interim design of the main loop residual heat removal system has been adopted. 2 refs
Primary Subject
Source
Anon; p. 2192-2201; 1979; p. 2192-2201; American Nuclear Society; La Grange Park, IL; International meeting on fast reactor safety technology; Seattle, WA, USA; 19 - 23 Aug 1979
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Book
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Conference
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