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.

Refinement of a system configration control program (II); Development of the database function for cutsets of accident sequences and enhancement of GUI

*; *; *

JNC TJ9440 2000-003, 173 Pages, 1999/03

JNC-TJ9440-2000-003.pdf:19.86MB

In FY 1997, a program which evaluates the risk in each phase of the maintenance was developed to support the maintenance planning on an FBR plant. In FY 1998, the GUI (Graphical User Interface) of the program developed in FY 1997 was enhanced for its user-friendlyness including facilitation of data settings and interpretation of the computational results. Specifically, following functions were incorporated : (1)To call associated window displays (editing and reporting display) mutually. (2)To edit the combinations of the systems for their maintenance scheduling. Furthermore, some risk evaluation functions such as the database function for cutsets of accident sequences and tracking function of risk trends were developed and added to its analysis module. A series of test on the programs with the GUI and the analysis module was performed and it was verified that the progam worked correctly.

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 ...

Development of failure mitigation technologies for improving resilience of nuclear structures, 7; Effectiveness evaluation technology of the measures for improving resilience at ultra-high temperatures

Onoda, Yuichi; Nishino, Hiroyuki; Kurisaka, Kenichi; Yamano, Hidemasa; Demachi, Kazuyuki*

no journal, , 

In order to evaluate the effectiveness of the measures for improving resilience at ultra-high temperatures, a concept of evaluation focusing on core damage frequency was proposed. Assuming loss of heat removal systems event after reactor shutdown which may result in core damage in sodium-cooled fast reactors, the measures for improving resilience which enable to recover the safety functions at ultra-high temperatures are identified: one is to retain the primary coolant using failure mitigation technology, and the other is to add a heat removal system that can be used under ultra-high temperature conditions. The core damage frequencies were calculated preliminarily and their reduction effect was estimated by comparing them before and after the introduction of the measures for improving resilience.

Development of failure mitigation technologies for improving resilience of nuclear structures, 20; Effectiveness evaluation methodology of the measures for improving resilience against excessive earthquake

Kurisaka, Kenichi; Nishino, Hiroyuki; Yamano, Hidemasa

no journal, , 

This study aims to evaluate the effectiveness of the measures for improving resilience against excessive earthquake. Loss of heat removal systems (LORHS) after reactor shutdown is focused on. Implementation of those measures could prevent the seismic-induced LOHRS itself or core damage due to ultra-high temperature condition resulting from LOHRS. Based on this, we developed the methodology that the effectiveness is evaluated by the reduction of core damage frequency. As an example, its applicability to sodium-cooled fast reactors was examined.