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Journal Articles

Recent verification activities on probabilistic fracture mechanics analysis code PASCAL4 for reactor pressure vessel

Lu, K.; Katsuyama, Jinya; Li, Y.; Miyamoto, Yuhei*; Hirota, Takatoshi*; Itabashi, Yu*; Nagai, Masaki*; Suzuki, Masahide*; Kanto, Yasuhiro*

Mechanical Engineering Journal (Internet), 7(3), p.19-00573_1 - 19-00573_14, 2020/06

Journal Articles

Improvements on evaluation functions of a probabilistic fracture mechanics analysis code for reactor pressure vessels

Lu, K.; Katsuyama, Jinya; Li, Y.

Journal of Pressure Vessel Technology, 142(2), p.021208_1 - 021208_11, 2020/04

 Times Cited Count:0 Percentile:100(Engineering, Mechanical)

Journal Articles

Guideline on probabilistic fracture mechanics analysis for Japanese reactor pressure vessels

Katsuyama, Jinya; Osakabe, Kazuya*; Uno, Shumpei*; Li, Y.; Yoshimura, Shinobu*

Journal of Pressure Vessel Technology, 142(2), p.021205_1 - 021205_10, 2020/04

 Times Cited Count:0 Percentile:100(Engineering, Mechanical)

no abstracts in English

JAEA Reports

Code-B-2.5.2 for stress calculation for SiC-TRISO fuel particle

Aihara, Jun; Goto, Minoru; Ueta, Shohei; Tachibana, Yukio

JAEA-Data/Code 2019-018, 22 Pages, 2020/01


Concept of Pu-burner high temperature gas-cooled reactor (HTGR) was proposed for purpose of more safely reducing amount of recovered Pu. In Pu-burner HTGR concept, coated fuel particle (CFP), with ZrC coated yttria stabilized zirconia (YSZ) containing PuO$$_{2}$$ (PuO$$_{2}$$-YSZ) small particle and with tri-structural isotropic (TRISO) coating, is employed for very high burn-up and high nuclear proliferation resistance. ZrC layer is oxygen getter. On the other hand, we have developed Code-B-2.5.2 for prediction of pressure vessel failure probabilities of SiC-tri-isotropic (TRISO) coated fuel particles for HTGRs under operation by modification of an existing code, Code-B-2. The main purpose of modification is preparation of applying code for CFPs of Pu-burner HTGR. In this report, basic formulae are described.

Journal Articles

Application of probabilistic fracture mechanics methodology for Japanese reactor pressure vessels using PASCAL4

Lu, K.; Katsuyama, Jinya; Li, Y.; Yoshimura, Shinobu*

Proceedings of 2019 ASME Pressure Vessels and Piping Conference (PVP 2019) (Internet), 9 Pages, 2019/07

Journal Articles

Effect of coolant water temperature of ECCS on failure probability of RPV

Katsuyama, Jinya; Masaki, Koichi; Lu, K.; Watanabe, Tadashi*; Li, Y.

Proceedings of 2019 ASME Pressure Vessels and Piping Conference (PVP 2019) (Internet), 7 Pages, 2019/07

For reactor pressure vessel (RPV) of pressurized water reactor, temperature of coolant water in emergency core cooling system (ECCS) may have influence on the structural integrity of RPV during pressurized thermal shock (PTS) events. Focusing on a mitigation measure to raise the coolant water temperature of ECCS for aged RPVs in order to reduce the effect of thermal shock due to PTS events, we performed thermal hydraulic analyses and probabilistic fracture mechanics analyses by using RELAP5 and PASCAL4, respectively. From the analysis results, it was shown that the failure probability of RPV was dramatically reduced when the coolant temperature in accumulator as well as high and low pressure injection systems (HPI/LPI) was raised, although raising the coolant temperature of HPI/LPI only did not cause reduction in the failure probability.

Journal Articles

Verification of a probabilistic fracture mechanics code PASCAL4 for reactor pressure vessels

Lu, K.; Katsuyama, Jinya; Li, Y.; Miyamoto, Yuhei*; Hirota, Takatoshi*; Itabashi, Yu*; Nagai, Masaki*; Suzuki, Masahide*; Kanto, Yasuhiro*

Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 9 Pages, 2019/05

Journal Articles

Development of remote sensing technique using radiation resistant optical fibers under high-radiation environment

Ito, Chikara; Naito, Hiroyuki; Ishikawa, Takashi; Ito, Keisuke; Wakaida, Ikuo

JPS Conference Proceedings (Internet), 24, p.011038_1 - 011038_6, 2019/01

A high-radiation resistant optical fiber has been developed in order to investigate the interiors of the reactor pressure vessels and the primary containment vessels at the Fukushima Daiichi Nuclear Power Station. The tentative dose rate in the reactor pressure vessels is assumed to be up to 1 kGy/h. We developed a radiation resistant optical fiber consisting of a 1000 ppm hydroxyl doped pure silica core and 4 % fluorine doped pure silica cladding. We attempted to apply the optical fiber to remote imaging technique by means of fiberscope. The number of core image fibers was increased from 2000 to 22000 for practical use. The transmissive rate of infrared images was not affected after irradiation of 1 MGy. No change in the spatial resolution of the view scope by means of image fiber was noted between pre- and post-irradiation. We confirmed the applicability of the probing system, which consists of a view scope using radiation-resistant optical fibers.

JAEA Reports

Mechanical properties database of reactor pressure vessel steels related to fracture toughness evaluation

Tobita, Toru; Nishiyama, Yutaka; Onizawa, Kunio

JAEA-Data/Code 2018-013, 60 Pages, 2018/11


Mechanical properties of materials including fracture toughness are extremely important for evaluating the structural integrity of reactor pressure vessels (RPVs). In this report, the published data of mechanical properties of nuclear RPVs steels, including neutron irradiated materials, acquired by the Japan Atomic Energy Agency (JAEA), specifically tensile test data, Charpy impact test data, drop-weight test data, and fracture toughness test data, are summarized. There are five types of RPVs steels with different toughness levels equivalent to JIS SQV2A (ASTM A533B Class 1) containing impurities in the range corresponding to the early plant to the latest plant. In addition to the base material of RPVs, the mechanical property data of the two types of stainless overlay cladding materials used as the lining of the RPV are summarized as well. These mechanical property data are organized graphically for each material and listed in tabular form to facilitate easy utilization of data.

Journal Articles

Development of crack evaluation models for probabilistic fracture mechanics analyses of Japanese reactor pressure vessels

Lu, K.; Masaki, Koichi; Katsuyama, Jinya; Li, Y.

Proceedings of 2018 ASME Pressure Vessels and Piping Conference (PVP 2018), 8 Pages, 2018/07

Journal Articles

Development of probabilistic fracture mechanics analysis code PASCAL Version 4 for reactor pressure vessels

Lu, K.; Masaki, Koichi; Katsuyama, Jinya; Li, Y.; Uno, Shumpei*

Proceedings of 2018 ASME Pressure Vessels and Piping Conference (PVP 2018), 10 Pages, 2018/07

JAEA Reports

Confirmation tests for Warm Pre-stress (WPS) effect in reactor pressure vessel steel (Contract research)

Chimi, Yasuhiro; Iwata, Keiko; Tobita, Toru; Otsu, Takuyo; Takamizawa, Hisashi; Yoshimoto, Kentaro*; Murakami, Takeshi*; Hanawa, Satoshi; Nishiyama, Yutaka

JAEA-Research 2017-018, 122 Pages, 2018/03


Warm pre-stress (WPS) effect is a phenomenon that after applying a load at a high temperature fracture does not occur in unloading during cooling, and then the fracture toughness in reloading at a lower temperature increases effectively. Engineering evaluation models to predict an apparent fracture toughness in reloading are established using experimental data with linear elasticity. However, there is a lack of data on the WPS effect for the effects of specimen size and surface crack in elastic-plastic regime. In this study, fracture toughness tests were performed after applying load-temperature histories which simulate pressurized thermal shock transients to confirm the WPS effect. The experimental results of an apparent fracture toughness tend to be lower than the predictive results using the engineering evaluation models in the case of a high degree of plastic deformation in preloading. Considering the plastic component of preloading can refine the engineering evaluation models.

Journal Articles

Benchmark analyses using probabilistic fracture mechanics analysis codes for reactor pressure vessels

Arai, Kensaku*; Katsuyama, Jinya; Li, Y.

Proceedings of Asian Symposium on Risk Assessment and Management 2017 (ASRAM 2017) (USB Flash Drive), 8 Pages, 2017/11

Probabilistic fracture mechanics (PFM) analysis code PASCAL has been developed to assess structural integrity of aged reactor pressure vessels (RPVs) of light water nuclear power plants by Japan Atomic Energy Agency (JAEA). PASCAL is able to obtain failure frequency such as through-wall cracking frequency (TWCF) of RPVs under several transients including pressurized thermal shock (PTS) event. On the other hand, FAVOR was developed to perform almost the same analysis by Oak Ridge National Laboratory (ORNL) under United States Nuclear Regulatory Commission (USNRC) funding and has been utilized in the US nuclear regulation. To improve the reliability of PFM analysis results of PASCAL, benchmark analyses between PASCAL and FAVOR were performed. This paper provides results of the benchmark analyses using analysis conditions and parameters of the US 3-loop pressurized water reactor (PWR) nuclear power plant. Furthermore, sensitivity analyses relating to differences of analysis models (ex. Embrittlement correlation model) between Japan and the US were also conducted.

Journal Articles

Probabilistic fracture mechanics analysis models for Japanese reactor pressure vessels

Lu, K.; Katsuyama, Jinya; Uno, Shumpei; Li, Y.

Proceedings of 2017 ASME Pressure Vessels and Piping Conference (PVP 2017) (CD-ROM), 8 Pages, 2017/07

Journal Articles

Fracture toughness evaluation of reactor pressure vessel steels by master curve method using miniature compact tension specimens

Tobita, Toru; Nishiyama, Yutaka; Otsu, Takuyo; Udagawa, Makoto; Katsuyama, Jinya; Onizawa, Kunio

Journal of Pressure Vessel Technology, 137(5), p.051405_1 - 051405_8, 2015/10

 Times Cited Count:6 Percentile:56.09(Engineering, Mechanical)

We conducted a series of fracture toughness tests based on the Master curve method for several specimen size and shapes, such as 0.16T-CT, pre-cracked Charpy type, 0.4T-CT and 1T-CT specimens, in commercially manufactured 5 kinds of A533B class1 steels with different impurity contents and fracture toughness levels. The reference temperature ($$T_{o}$$) values determined from the 0.16T-CT specimens were overall in good agreement with those determined from the 1T-CT specimens. The scatter of the 1T-equivalent fracture toughness values obtained from the 0.16T-CT specimens was equivalent to that obtained from the other larger specimens. The higher loading rate gave rise to a slightly higher $$T_{o}$$, and this dependency was almost the same for the larger specimens. We suggested an optimum test temperature on the basis of the Charpy transition temperature for determining $$T_{o}$$ using the 0.16T-CT specimens.

Journal Articles

Development of Stress intensity factor coefficients database for a surface crack of an RPV considering the stress discontinuity between cladding and base metal

Onizawa, Kunio; Shibata, Katsuyuki*; Suzuki, Masahide

Proceedings of 2005 ASME/JSME Pressure Vessels and Piping Division Conference (PVP 2005), 12 Pages, 2005/07

Under a transient loading like pressurized thermal shock (PTS), the stress discontinuity near the interface between cladding and base metal of a reactor pressure vessel (RPV) is caused by the difference in their thermal expansion factors. So the stress intensity factor (SIF) of a surface crack which the deepest point exceeds the interface should be calculated by taking account of the stress discontinuity. Many SIF calculations are performed in Monte Carlo simulation of the probabilistic fracture mechanics (PFM) analysis. To avoid the time consuming process from the SIF calculation in the PFM analysis, the non-dimensional SIF coefficients corresponding to the stress distributions in the cladding and base metal were developed. The non-dimensional SIF coefficients database were obtained from 3D FEM analyses. The SIF value at the surface was determined by linear extrapolation of SIF value near the surface. Using the SIF coefficients database, the SIF values at both surface and deepest points of a surface crack can be evaluated precisely and in a reasonable time.

Journal Articles

Reactor pressure vessel design of the high temperature engineering test reactor

Tachibana, Yukio; Nakagawa, Shigeaki; Iyoku, Tatsuo

Nuclear Engineering and Design, 233(1-3), p.103 - 112, 2004/10

 Times Cited Count:1 Percentile:89.27(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Importance of fracture criterion and crack tip material characterization in probabilistic fracture mechanics analysis of an RPV under a pressurized thermal shock

Shibata, Katsuyuki; Onizawa, Kunio; Li, Y.*; Kato, Daisuke*

International Journal of Pressure Vessels and Piping, 81(9), p.749 - 756, 2004/09

 Times Cited Count:5 Percentile:63.05(Engineering, Multidisciplinary)

The paper describes the procedure to evaluate the ductile crack extension, where an increase in fracture resistance by a ductile crack extension is considered. Two standard ${it J}$-resistance curves are prepared for applying the elasto-plastic fracture criterion. Case studies concerning the effect of elasto-plastic fracture criterion were carried out using a severe PTS transient. The introduction of the elasto-plastic fracture criterion significantly contributes to remove the over-conservatism in applying the linear elastic fracture criterion. It was also found that the algorithm of the re-evaluation of crack tip characterization also has a significant effect on the failure probability.

Journal Articles

Design and fabrication of reactor pressure vessel for High Temperature Engineering Test Reactor (HTTR)

Tachibana, Yukio; Nakagawa, Shigeaki; Iyoku, Tatsuo

Elevated Temperature Design and Analysis, Nonlinear Analysis, and Plastic Components, 2004 (PVP-Vol.472), p.39 - 44, 2004/07

The reactor pressure vessel (RPV) of the HTTR is 5.5m in inside diameter, 13.2m in inside height, and 122mm and 160mm in wall thickness of the body and the top head dome, respectively. Because the reactor inlet temperature of the HTTR is higher than that of LWRs, 2 1/4Cr-1Mo steel is chosen for the RPV material. Fluence of the RPV is estimated to be less than 1$$times$$10$$^{17}$$n/cm$$^{2}$$(E$$>$$1 MeV), and so irradiation embrittlement is presumed to be negligible, but temper embrittlement is not. For the purpose of reducing embrittlement, content of some elements is limited on 2 1/4 Cr-1 Mo steel for the RPV using embrittlement parameters, J-factor and X-bar. In this paper design, fabrication procedure, and in-service inspection technique of the RPV for the HTTR are described.

Journal Articles

Correlation between cleavage fracture toughness and charpy impact properties in the transition temperature range of reactor pressure vessel steels

Onizawa, Kunio; Suzuki, Masahide

JSME International Journal, Series A, 47(3), p.479 - 485, 2004/07

In the structural integrity assessment of reactor pressure vessel, fracture toughness values are estimated by assuming that the radiation effect on fracture toughness is equivalent to that on Charpy properties. Therefore, it is necessary to establish the correlation between both properties especially on irradiation embrittlement. In this paper, we present the fracture toughness data obtained by applying the master curve approach that was adopted recently in the ASTM test method. Materials used in this study are five ASTM A533B class 1 steels and one weld metal. Neutron irradiation for Charpy-size specimens as well as standard Charpy-v specimens was carried out at the Japan Materials Testing Reactor. The shifts of the reference temperature on fracture toughness due to neutron irradiation are evaluated. Correlation between the fracture toughness reference temperature and Charpy transition temperature is established. Based on the correlation, the optimum test temperature for fracture toughness testing and the method to determine a lower bound fracture toughness curve are discussed.

44 (Records 1-20 displayed on this page)