[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

[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

[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

No abstract available

[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

[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

[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

[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

[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)

[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