Research and examination of seismic safety evaluation and function maintenance for important equipment in nuclear facilities

Furuya, Osamu*; Fujita, Satoshi*; Muta, Hitoshi*; Otori, Yasuki*; Itoi, Tatsuya*; Okamura, Shigeki*; Minagawa, Keisuke*; Nakamura, Izumi*; Fujimoto, Shigeru*; Otani, Akihito*; et al.

Proceedings of ASME 2021 Pressure Vessels and Piping Conference (PVP 2021) (Internet), 6 Pages, 2021/07

Since the Fukushima accident, with the higher safety requirements of nuclear facilities in Japan, suppliers, manufacturers and academic societies have been actively considering the reconstruction of the safety of nuclear facilities from various perspectives. The Nuclear Regulation Authority has formulated new regulatory standards and is in operation. The new regulatory standards are based on defense in depth, and have significantly raised the levels of natural hazards and have requested to strengthen the countermeasures from the perspective of preventing the simultaneous loss of safety functions due to common factors. Facilities for dealing with specific serious accidents are required to have robustness to ensure functions against earthquakes that exceed the design standards to a certain extent. In addition, since the probabilistic risk assessment (PRA) and the safety margin evaluation are performed to include the range beyond the design assumption in the safety improvement evaluation, it is very important to extent the special knowledge in the strength of important equipment for seismic safety. This paper summarizes the research and examination results of specialized knowledge on the concept of maintaining the functions of important seismic facilities and the damage index to be considered by severe earthquakes. In the other paper, the study on reliability of seismic capacity analysis for important equipment in nuclear facilities will be reported.

Reliability enhancement of seismic risk assessment of NPP as risk management fundamentals, 3; Sensitivity analysis for the quantification of epistemic uncertainty on fragility assessment

Nishida, Akemi; Choi, B.; Itoi, Tatsuya*; Takada, Tsuyoshi*; Furuya, Osamu*; Muramatsu, Ken*

Transactions of 23rd International Conference on Structural Mechanics in Reactor Technology (SMiRT-23) (USB Flash Drive), 10 Pages, 2015/08

This study focused on uncertainty-assessment frameworks and the development of relevant software to improve the reliability of seismic probabilistic risk assessment (SPRA) for NPP and promote its further use. This research aimed at contributing to development of implementation guidelines on epistemic uncertainty. Some sensitivity analyses were performed using a three dimensional reactor-building model and a conventional evaluation model by using 3D vibration simulator for NPP of JAEA, and the results were provided to the experts for expert-opinion elicitation. The results of the sensitivity analyses were related to the uncertainty evaluation of the buildings and soil to evaluate the fragility of the equipment. In this paper, those results will be shown in comparison with a conventional evaluation model.

Reliability enhancement of seismic risk assessment of NPP as risk management fundamentals, 2; Quantifying epistemic uncertainty in fragility assessment using expert opinions

Takada, Tsuyoshi*; Itoi, Tatsuya*; Nishida, Akemi; Furuya, Osamu*; Muramatsu, Ken*

Transactions of 23rd International Conference on Structural Mechanics in Reactor Technology (SMiRT-23) (USB Flash Drive), 10 Pages, 2015/08

The assessment of seismic safety of nuclear power plant facilities has been performed by identifying and quantifying uncertainties in seismic probabilistic risk assessment (SPRA). For the evaluation process, all uncertainties are classified into either aleatory or epistemic uncertainty for their quantification. In this study, systematic evaluation of the epistemic uncertainty on the seismic fragility of structures and equipment is studied and implemented for a model NPP. There are two expert groups formed in this project: experts in the field of buildings and soil ground (CE experts) and experts in the field of pipe and equipment (ME experts). Along with ample results from relevant sensitivity analyses conducted, elicited opinions are carefully treated and classified into several specific areas and integrated into the form of knowledge tree technique (KTT), all of which can be utilized for future fragility estimation.

Review of JAEA activities on the IFMIF liquid lithium target in FY2006

Ida, Mizuho; Nakamura, Hiroo; Chida, Teruo*; Miyashita, Makoto; Furuya, Kazuyuki*; Yoshida, Eiichi; Hirakawa, Yasushi; Miyake, Osamu; Hirabayashi, Masaru; Ara, Kuniaki; et al.

JAEA-Review 2008-008, 38 Pages, 2008/03

JAEA-Review-2008-008.pdf:9.37MB

Engineering Validation Design and Engineering Design Activity (EVEDA) of the International Fusion Materials Irradiation Facility (IFMIF) is under going. IFMIF is an accelerator-based Deuterium-Lithium (D-Li) neutron source to produce intense high energy neutrons and a sufficient irradiation volume for testing candidate materials for fusion reactors. To realize such a condition, 40 MeV deuteron beam with a current of 250 mA is injected into high speed liquid Li flow with a speed of 20 m/s. In target system, nuclear heating due to neutron causes thermal stress especially on a back-wall of the target assembly. In addition, radioactive species such as beryllium-7, tritium and activated corrosion products are generated. In this report, thermal stress analyses of the back-wall, mechanical tests on weld specimen made of the back-wall material, estimations of beryllium-7 behavior and worker dose at the IFMIF Li loop and consideration on major EVEDA tasks are summarized.

Detailed design of new material nitric acid recovery evaporator

Yokoyama, Hiroomi*; Kanazawa, Toshio*; Fukuma, Tadashi*; Tamekiyo, Kozo*; Yanagida, Koji*; Furuya, Takashi*; Kono, Hiroshi*; Ito, Keiji*; Shirakura, Takao*; Kashiwara, Shinichiro*; et al.

PNC TN8410 87-086VOL1, 1037 Pages, 1986/09

PNC-TN8410-87-086VOL1.pdf:34.39MB

A detailed design for a New Material Nitric Acid Recovery Evaporator was carried out with a plan to use it to replace an already constructed stainless steel Nitric Acid Recovery Evaporator at the Power Reactor and Nuclear Fuel Development Corporation's Tokai Works. Most of the original Evaporator's conditions such as compliance with applicable laws, standards, structure, treatment performance and operating conditions were maintained when designing the new machine. The material is titanium with the addition of 5% tantalum. The Evaporator was designed with an operational life expectance of 10 years. We have calculated that the new Evaporator will have sufficient strength (including a seismatic design) and have the same evaporative performance as the already constructed one. During design, we referred to the results of already completed basic designs (Phase 1) of New Material Nitric Acid Recovery Evaporators, design and production of small-scale test equipment units, and the development of successful joints between different materials. We also considered manufacturing, installation, trial runs, maintenance, and the specifications for materials used for manufacturing, installation, piping and operation of the new Evaporator.

Study on reliability enhancement of seismic risk assessment of NPP, 1; Program plan and development of new framework of probabilistic models

Muramatsu, Ken*; Furuya, Osamu*; Fujimoto, Shigeru*; Hirano, Mitsumasa*; Muta, Hitoshi*; Takada, Tsuyoshi*; Itoi, Tatsuya*; Nishida, Akemi; Uchiyama, Tomoaki*

no journal, , 

This project, focusing on uncertainty assessment framework and utilization of expertise, and finally developing relevant computer codes to improve reliability of seismic probabilistic risk assessment (SPRA) and to promote its further use of the SPRA, develops methodology for quantification of uncertainty associated with final results from SPRA in the framework of risk management of NPP facilities. The following scopes are set. (1) Development of framework of probabilistic models for uncertainty quantification and Computer codes (2) Aggregation of expert opinion on structure/equipment fragility estimation and development of implementation guidance on epistemic uncertainty (modeling uncertainty) (3) Study on applicability of SPRA to model plant. In this report, program plan of our research and new probabilistic models and treatment of epistemic uncertainty will be explained as part 1.

Sensitivity analysis for quantification of uncertainty in fragility evaluation

Nishida, Akemi; Itoi, Tatsuya*; Takada, Tsuyoshi*; Furuya, Osamu*; Muramatsu, Ken*

no journal, , 

This research develops methodology for quantification of uncertainty associated with final results from seismic probabilistic risk assessment (SPRA) in the framework of risk management of NPP facilities. In this research, we take a role of aggregation of expert opinion on structure/equipment fragility estimation and development of implementation guidance on epistemic uncertainty (modeling uncertainty). In this report, a trial evaluation result of quantification of uncertainty using analytical results of seismic response analysis and sensitivity analysis for a model plant are shown.

Reliability enhancement of seismic risk assessment of NPP as risk management fundamentals, 13; Evaluation of piping and components

Furuya, Osamu*; Fujimoto, Shigeru*; Muramatsu, Ken*; Muta, Hitoshi*; Nishida, Akemi

no journal, , 

This project, focusing on uncertainty assessment framework and utilization of expertise, and finally developing relevant computer codes to improve reliability of seismic probabilistic risk assessment (SPRA) and to promote its further use of the SPRA, develops methodology for quantification of uncertainty associated with final results from SPRA in the framework of risk management of NPP facilities. The following scopes are set. (1) Development of framework of probabilistic models for uncertainty quantification and Computer codes, (2) Aggregation of expert opinion on structure/equipment fragility estimation and development of implementation guidance on epistemic uncertainty (modeling uncertainty), (3) Study on applicability of SPRA to model plant. In this report, epistemic uncertainty of the equipment and piping responses on the same floor is estimated by analytical point of view.

Behaviour with damage on steel pipe joints subjected to large-scale deformation

Nakamura, Izumi*; Takito, Kiyotaka; Okuda, Yukihiko; Furuya, Osamu*

no journal, , 

no abstracts in English