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[en] Goals: – Develop the fundamental scientific basis to understand, predict, and measure changes in materials and systems, structures and components (SSCs) as they age in environments; – Apply this knowledge to develop and demonstrate methods and technologies that support safe and economical long-term operation of existing reactors; – Research new technologies that enhance plant performance, economics, and safety.
[en] This report contains the safety analyses of the KALIMER-600 conceptual design which KAERI has been developing under the Long-term Nuclear R and D Program. The analyses have been performed reflecting the design developments during the second year of the 4th design phase in the program. The specific presentations are the key design features with the safety principles for achieving the safety objectives, the event categorization and safety criteria, and results on the safety analyses for the DBAs and ATWS events, the containment performance, and the channel blockages. The safety analyses for both the DBAs and ATWS events have been performed using SSC-K version 1.3., and the results have shown the fulfillment of the safety criteria for DBAs with conservative assumptions. The safety margins as well as the inherent safety also have been confirmed for the ATWS events. For the containment performance analysis, ORIGEN-2.1 and CONTAIN-LMR have been used. In results, the structural integrity has been acceptable and the evaluated exposure dose rate has been complied with 10 CFR 100 and PAG limits. The analysis results for flow blockages of 6-subchannels, 24-subchannels, and 54- subchannels with the MATRA-LMR-FB code, have assured the integrity of subassemblies
[en] This publication results from a technical meeting on phenomenology and technologies relevant to in-vessel melt retention (IVMR) and ex-vessel corium cooling (EVCC). The purpose of the publication is to capture the state of knowledge, at the time of that meeting, related to phenomenology and technologies as well as the challenges and pending issues relevant to IVMR and EVCC for water cooled reactors by summarizing the information provided by the meeting participants in a form useful to practitioners in Member States.
[en] The Nuclear Engineering Department at Israel Electric Company has been engaged for a number of years in a joint research agreement with the Technion Nuclear Reactor Research Group on various thermal-hydraulic aspects of reactor design and safety. Besides developing their own analytical models, the researchers rely heavily on the RELAPS computer code in their analyses. The RELAPS series are general purpose, thermal-hydraulic system codes, used to simulate system response (such as the RCS) to transients and accidents. They are based on solving the equations of conservation of mass, energy and momentum within the system being modeled, where the model is a series of control volumes connected by junctions. The equations are solved simultaneously in each volume and junction using a finite difference numerical scheme. As an example, a recent report refers to containment response to a large LOCA in an AP600-like advanced rector. This work has been performed using RELAPS/Mod2. Accidents like LOCA represent design base events necessary to verify the adequacy of the emergency core cooling system, the passive containment system and other safety systems. The validation of simulation results is therefore important to the IEC staff responsible for monitoring the research. (author); 3 refs
[en] Conclusions: Proposed V-1 ALS can cope with 2 x DN 500 DBA. Work on V-1 ALS able to cope with DN 200 DBA and to mitigate DN 500 BDBA will have been finished by April 1994. Further leaktightness improvements are going on on the V-1 HZ boundary during outages with the aim to reduce the existing leak rate by 50%
[en] Fuel cans of defined failure size and type were operated in the HSD loop of the FR2 reactor. The experiments were carried out with the purpose 1) of measuring the emission of fission products and fuel particles from failed fuel rods, 2) of determining the behaviour of failed fuel elements in constant-load and load-following operation, and 3) of determining the type of failure from the measuring signals. (orig./AK)
[de]Im Huellenschaden-Dampfkreislauf (HSD-Loop) des FR2 wurden Brennstaebe definierter Defektgroesse und -art betrieben. Ziel der Versuche war: 1) Messung der Freisetzung von Spaltprodukten und Brennstoffpartikeln aus beschaedigten Brennstaeben, 2) Bestimmung des Verhaltens defekter Brennelemente bei Konstantlast- und Lastfolgebetrieb, 3) Erkennen verschiedener Defektarten aus den Messsignalen. (orig./RW)
[en] The objective of this paper is to assess proposed transuranic waste accident analysis guidance and recent software improvements in a Windows-OS version of MACCS2 that allows the inputting of parameter uncertainty. With this guidance and code capability, there is the potential to perform a quantitative uncertainty assessment of unmitigated accident releases with respect to the 25 rem Evaluation Guideline (EG) of DOE-STD-3009-94 CN3 (STD-3009). Historically, the classification of safety systems in a U.S. Department of Energy (DOE) nuclear facility's safety basis has involved how subject matter experts qualitatively view uncertainty in the STD-3009 Appendix A accident analysis methodology. Specifically, whether consequence uncertainty could be larger than previously evaluated so the site-specific accident consequences may challenge the EG. This paper assesses whether a potential uncertainty capability for MACCS2 could provide a stronger technical basis as to when the consequences from a design basis accident (DBA) truly challenges the 25 rem EG
[en] FDS-I/DWT blanket Concepts and Designs optimization has been carrying out. Basic analysis models have been developed for neutronics/thermalhydraulics/magnetics analyses. Necessary codes and data libraries are ready for doing design optimization and static analyses, some more work is needed to prepare enough functions for doing practical transient analyses. Preliminary and tentative analyses on some hypothetical transients (LOCA, LOFA, Overpower etc.) have been performed