Event tree / fault tree assessment of explosion loads

Nishida, Akemi

Doboku Gakkai Dai-14-Kai Kozobutsu No Shogeki Mondai Ni Kansuru Shinpojiumu Rombunshu (Internet), 5 Pages, 2024/01

no abstracts in English

Improvement of DYANA; The Dynamic analysis program for event transition

Tamura, Kazuo*; Iriya, Yoshikazu*

JNC TJ9440 2000-004, 22 Pages, 2000/03

JNC-TJ9440-2000-004.pdf:2.35MB

In the probabilistic safety assessment(PSA), the fault tree/event tree technique has been widely used to evaluate accident sequence frequencies. However, event tansition which operators actually face can not be dynamically treated by the conventional technique. Therefore, we have made the dynamic analysis program(DYANA) for event transition for a liquid metal cooled fast breeder reactor. In the previous development, we made basic model for analysis. However, we have a probrem that calculation time is too long. At the current term, we made parallelization of DYANA usig MPI. So we got good performance on WS claster. It performance is close to ideal one.

Level-1 PSA on large fast breeder reactor (II); Evaluation of PLOHS frequency with the water steam system with decay heat removal capability

Hioki, Kazumasa

PNC TN9410 94-188, 160 Pages, 1994/05

PNC-TN9410-94-188.pdf:8.75MB

The Systems Analysis Section has been performing a probabilistic Safety Assessment (PSA) on a large fast breeder reactor (FBR) since JFY 1992. The objective of the study is to apply the PSA method to a plant in a conceptual design stage, develop system models, perform quantitative analyses and systematic evaluation, supply valuable insights to enhance reliability and safety, and reflect them to the basic design. The plant analyzed is a 600MWe class large FBR designed by the Plant Engineering Section in the "Large FBR design study" that has been performed since JFY 1990. The failure probability of the Decay Heat Removal System (DHRS) can be reduced approximately two orders if the Water Steam System (WSS) can remove the decay heat for the first 24 hours. The frequency of PLOHS, however, is not reduced to less than one third because the WSS cannot be used for some initiating events and the PLOHS frequency is dominated by the failure probability of DHRS without the WSS. The failure probability of DHRS is dominated by the common cause failures (CCFs) of vanes, dampers and valves around the air-coolers in the Auxiliary Cooling System (ACS). Therefore it is most important to eliminate the CCFs. Assuming that the CCFs have been eliminated by diversifying the components, the frequencies of PLOHS were evaluated. An analysis has shown that if the WSS can remove the decay heat alone, the PLOHS frequency is reduced approximately two orders. In this case the PLOHS frequency is dominated by the failure probability of the DHRS right after the reactor shutdown. The most effective way to reduce the PLOHS frequency is to increasc the redundancy of the DHRS for the first few hours after reactor shutdown. It is known through the experience of preceding plants that the success criteria can be relaxed to one loop natural circulation instead of forced circulation in the best estimate evaluation. It was shown that under such condition, the PLOHS frequency can be as low as 10 ...

ETAP users manual

Watanabe, Norio; *

JAERI-M 90-193, 55 Pages, 1990/11

JAERI-M-90-193.pdf:1.48MB

no abstracts in English

The plant thermohydraulic analysis for the monju PRA study; Recovery from PLOHS or LORL using the maintenance cooling system

*; *

PNC TN9410 88-055, 111 Pages, 1988/06

PNC-TN9410-88-055.pdf:5.87MB

In this study, decay heat removal capability of the Maintenance Cooling System (MCS) of Monju has been investigated with respect to protected accidents. The protected accidents of the Liquid Metal Fast Breeder Reactors (LMFBRs), such as Protected Loss-of-Heat-Sink (PLOHS) or Loss-of-Reactor-Level (LORL), are of great importance from the viewpoint of the annual frequency of core damage. The progression of the protected accidents is mild in general because reactor decay heat can be dispersed from the core by natural circulation. The decay heat for Monju is to be removed by the Intermediate Reactor Auxiliary Cooling system (IRACS). It is essential to keep the intactness of coolant flow path from the reactor core to the heat sink and the availability of heat sink itself. If the either of them is degraded, it is taken for granted d that protected slow meltdown follows. However, the reactor core can be prevented from any damage or meltdown if the decay heat can be removed through MCS. The plant thermohydraulics of the procected accidents is analyzed using SSC-L to develop success criteria in the decay heat removal by the MCS. Parametric calculations are performed with respect to: available heat capacity in the heat transport system, cooling time before the loss-of-heat-sink and MCS starting time. It has been found, for example, that (1)MCS can remove the decay heat immediately after the reactor shutdown if heat capacity of more than two main coolant loops is available; (2)after two hours cooling time by natual circulation, MCS can remove the decay heat even if no coolant flow is assumed in all the main heat transport system; (3)LORL caused by the failure in sodium make-up can be recovered by the MCS operation. In the PLOHS condition, the coolant temperature may exceed conservative design limit of the MCS piping. However, the conservativeness of the design limit and the method of qualification make compensation for the deterioration in structural strength. Finally, ...