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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.

JAEA Reports

Assessment report of research and development on "Nuclear Safety Research" in FY2014 (Post- and pre-review report)

Kudo, Tamotsu; Onizawa, Kunio*; Nakamura, Takehiko

JAEA-Evaluation 2015-011, 209 Pages, 2015/11


Japan Atomic Energy Agency (JAEA) consulted an assessment committee, "Evaluation Committee of Research and Development (R&D) Activities for Nuclear Safety", for post- and pre-review assessment of R&D on nuclear safety research. In response to JAEA's request, the Committee assessed mainly the progress of the R&D project according to guidelines, which addressed the rationale behind the R&D project, the relevance of the project outcome and the efficiency of the project implementation during the period of the current and next plan. As a result, the Committee concluded that the progress of the R&D project is satisfactory. This report describes the results of evaluation by the Committee. In addition, the appendix of this report contains presentations used for the evaluation, and responses from JAEA on the comments from the member of the Committee.

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:52.49(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

Failure probability analyses for PWSCC in Ni-based alloy welds

Udagawa, Makoto; Katsuyama, Jinya; Onizawa, Kunio; Li, Y.

International Journal of Pressure Vessels and Piping, 131, p.85 - 95, 2015/07

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

A number of cracks due to primary water stress corrosion cracking (PWSCC) in pressurized water reactors and Ni-based alloy stress corrosion cracking (NiSCC) in boiling water reactors have been detected around Ni-based alloy welds. The causes of crack initiation and growth due to stress corrosion cracking include weld residual stress, operating stress, the materials, and the environment. We have developed the analysis code PASCAL-NP for calculating the failure probability and assessment of the structural integrity of cracked components on the basis of probabilistic fracture mechanics (PFM) considering PWSCC and NiSCC. This PFM analysis code has functions for calculating the incubation time of PWSCC and NiSCC crack initiation, evaluation of crack growth behavior considering certain crack location and orientation patterns, and evaluation of failure behavior near Ni-based alloy welds due to PWSCC and NiSCC in a probabilistic manner. Herein, actual plants affected by PWSCC have been analyzed using PASCAL-NP. Failure probabilities calculated by PASCAL-NP are in reasonable agreement with the detection data. Furthermore, useful knowledge related to leakage due to PWSCC was obtained through parametric studies using this code.

Journal Articles

Development of J-integral solutions for semi-elliptical circumferential cracked pipes subjected to internal pressure and bending moment

Udagawa, Makoto; Katsuyama, Jinya; Yamaguchi, Yoshihito; Li, Y.; Onizawa, Kunio

Proceedings of 2015 ASME Pressure Vessels and Piping Conference (PVP 2015) (Internet), 9 Pages, 2015/07

The J-integral solution for cracked pipes is a high important parameter in crack growth calculation and failure evaluation based on the elastic-plastic fracture mechanics. One of the most important crack types in structural integrity assessment for nuclear piping systems is circumferential semi-elliptical surface crack. Although several J-integral solutions have been provided, no solution was developed at both the deepest and the surface points of circumferential semi-elliptical surface cracks. In this study, the J-integral solutions of circumferential semi-elliptical surface cracks were developed by numerical finite element analyses. Moreover, in order to benefit users in practical applications, a pair of convenient J-integral estimation equations were developed. The accuracy and applicability of the convenient equations were confirmed by comparing with the provided stress intensity factor solutions in elastic region and with finite element analysis results in elastic-plastic region.

Journal Articles

Finite element analysis on the application of Mini-C(T) test specimens for fracture toughness evaluation

Takamizawa, Hisashi; Tobita, Toru; Otsu, Takuyo; Katsuyama, Jinya; Nishiyama, Yutaka; Onizawa, Kunio

Proceedings of 2015 ASME Pressure Vessels and Piping Conference (PVP 2015) (Internet), 7 Pages, 2015/07

Fracture toughness evaluation by the Master Curve method using miniature compact tension (Mini-C(T)) specimens taken from the broken halves of surveillance Charpy specimens has been proposed. We performed finite element analysis (FEA) to examine the difference in constraint effect of the crack tip for the different size C(T) and precracked Charpy v-notch specimens. The constraint effect for Mini-C(T) specimens in terms of the T-stress and Q-parameter was similar to the larger C(T) specimens. To optimize the fatigue precracking conditions for the Mini-C(T) specimen fabrication, plastic zone distribution analysis was performed. We confirmed the fatigue precrack length and the availability of the mitigated crack shape for Mini-C(T). We also obtained the fracture toughness data for different sizes specimen. It was shown that To obtained from the Mini-C(T) specimens is in reasonably good agreement with that from others. We compared the fracture toughness data with T41J based fracture toughness curves proposed in recent study. All of the data were well enveloped by the proposed lower bound curve.

Journal Articles

Crack growth evaluation for cracked carbon and stainless steel pipes under large seismic cyclic loading

Yamaguchi, Yoshihito; Katsuyama, Jinya; Li, Y.; Onizawa, Kunio

Proceedings of 2015 ASME Pressure Vessels and Piping Conference (PVP 2015) (Internet), 8 Pages, 2015/07

Japanese nuclear power plants have recently experienced several large earthquakes beyond the previous design basis ground motion. In addition, cracks resulting from long-term operation have been detected in piping lines. Therefore, it is very important to establish a crack growth evaluation method for cracked pipes that are subjected to large seismic cyclic response loading. In our previous study, we proposed an evaluation method for crack growth during large earthquakes through experimental study using small specimens. In the present study, crack growth tests were conducted on pipes with a circumferential through-wall crack, considering large seismic cyclic response loading with complex wave forms. The predicted crack growth values are in good agreement with the experimental results for both stainless and carbon steel pipe specimens and the applicability of the proposed method was confirmed.

Journal Articles

Effects of plasticity on the stress intensity factor evaluation for underclad crack under pressurized thermal shock events

Katsuyama, Jinya; Huang, L.; Li, Y.; Onizawa, Kunio

Proceedings of 2015 ASME Pressure Vessels and Piping Conference (PVP 2015) (Internet), 8 Pages, 2015/07

When the structural integrity of reactor pressure vessel (RPV) under pressurized thermal shock (PTS) events is assessed, an underclad crack is postulated at the inner surface of RPV and the stress intensity factor (SIF) is evaluated for this crack. On the inner surface of RPV, cladding of stainless steel is overlay-welded as a means for corrosion protection. Because the cladding is a ductile material, it is important to evaluate the SIF considering the plasticity of cladding. A SIF evaluation method considering the effect of plasticity has been established in France. In this study, we examined the SIF evaluation method for underclad cracks during PTS transients. The elastic and elastic-plastic analyses based on the finite element method considering PTS events and inner pressure were performed using three-dimensional models including an underclad semi-elliptical crack with different geometry. We showed the conservativeness of plastic correction method based on the analysis results.

Journal Articles

Development of probabilistic evaluation models of fracture toughness K$$_{Ic}$$ and K$$_{Ia}$$ for Japanese RPV steels

Katsuyama, Jinya; Katsumata, Genshichiro; Onizawa, Kunio; Osakabe, Kazuya*; Yoshimoto, Kentaro*

Proceedings of 2015 ASME Pressure Vessels and Piping Conference (PVP 2015) (Internet), 9 Pages, 2015/07

Probabilistic fracture mechanics (PFM) analysis code PASCAL3 has been developed to apply the PFM analysis to the structural integrity assessment of domestic RPVs. In this paper, probabilistic evaluation models of fracture toughness KIc and KIa which have the largest scatter among the associated factors based on the database of Japanese RPV steels are presented. We developed probabilistic evaluation models for KIc and KIa based on the Weibull and lognormal distributions, respectively. The models are compared with the existing lower bound of fracture toughness in the Japanese code and probabilistic model in USA. As the results, the models established in present work satisfy lower bounds of fracture toughness in the Japanese code. The comparison in the models between present work and US showed significant differences that may have an influence on fracture probability of RPV.

Journal Articles

Study on application of PFM analysis method to Japanese code for RPV integrity assessment under PTS events

Osakabe, Kazuya*; Masaki, Koichi*; Katsuyama, Jinya; Katsumata, Genshichiro; Onizawa, Kunio; Yoshimura, Shinobu*

Proceedings of 2015 ASME Pressure Vessels and Piping Conference (PVP 2015) (Internet), 8 Pages, 2015/07

A probabilistic fracture mechanics (PFM) analysis method for pressure boundary components is useful to evaluate the structural integrity in a quantitative way. This is because the uncertainties related to influence parameters can be rationally incorporated in PFM analysis. From this viewpoint, the probabilistic approach evaluating through-wall cracking frequencies (TWCFs) of reactor pressure vessels (RPVs) has already been adopted as the regulation on fracture toughness requirements against PTS events in the U.S. As a study of applying PFM analysis to the integrity assessment of domestic RPVs, JAEA has been preparing input data and analysis models to calculate TWCFs using PFM analysis code PASCAL3. In this paper, activities have been introduced such as preparing input data and models for domestic RPVs, verification of PASCAL3, and formulating guideline on general procedures of PFM analysis for the purpose of utilizing PASCAL3. In addition, TWCFs for a model RPV evaluated by PASCAL3 are presented.

Journal Articles

International round robin test on Master Curve reference temperature evaluation utilizing Miniature C(T) specimen

Yamamoto, Masato*; Onizawa, Kunio; Yoshimoto, Kentaro*; Ogawa, Takuya*; Mabuchi, Yasuhiro*; Valo, M.*; Lambrecht, M.*; Viehrig, H.-W.*; Miura, Naoki*; Soneda, Naoki*

Small Specimen Test Techniques; 6th Volume (ASTM STP 1576), p.53 - 69, 2015/05

In order to ensure the robustness of the Master Curve technique, round-robin tests were performed using 0.16 inch-thick Mini-CT specimens by different investigators to see if consistent $$T$$$$_{0}$$ values can be obtained. All the specimens used were machined and pre-cracked by one fabricator from unique Japanese RPV material. Seven institutes participated in this exercise, and obtained valid $$T$$$$_{0}$$ values according to the ASTM E1921 standard. The scatter of $$T$$$$_{0}$$ values obtained was well within the uncertainty range defined in the standard, indicating the robustness of the Mini-CT specimen test technique. Throughout this activity, we could obtain 182 $$K$$$$_{Jc}$$ for a single material. We investigated the statistics of this large database, and found that there is no remarkable difference not only in the $$T$$$$_{0}$$ values but also in the fracture toughness distribution between the Mini-CT specimen and the standard size 1T-C(T) specimen results.

Journal Articles

Fracture evaluation of reactor pressure vessel steel based on local approach

Takamizawa, Hisashi; Katsuyama, Jinya; Yamaguchi, Yoshihito; Nishiyama, Yutaka; Li, Y.; Onizawa, Kunio

Yosetsu Kozo Shimpojiumu 2014 Koen Rombunshu, p.97 - 100, 2014/12

no abstracts in English

Journal Articles

Effect of cyclic loading on the relaxation of residual stress in the butt-weld joints of nuclear reactor piping

Katsuyama, Jinya; Yamaguchi, Yoshihito; Li, Y.*; Onizawa, Kunio

Nuclear Engineering and Design, 278, p.222 - 228, 2014/10

 Times Cited Count:5 Percentile:48.21(Nuclear Science & Technology)

Weld residual stress is one of the most important factors in stress corrosion cracking (SCC) of nuclear reactor piping. To assess the integrity of piping, it is necessary to understand the effects of excessive cyclic loading, caused by earthquake, on residual stress. In this study, finite element analyses were performed using a model of a 250A pipe butt weld of Type 316L stainless steel. The accuracy of the welding simulation was verified by comparing the calculated results with the experimental measurements. After conducting the welding simulation and residual stress analysis, several loading patterns for the axial cyclic loadings were applied to the model by varying the amount of maximum load, in order to study the effect of excessive cyclic loading. The analysis indicated that higher loading caused a greater relaxation of the weld residual stress near the piping welds. It was thus concluded that excessive cyclic loading on piping butt welds affects the suppression of SCC growth.

Journal Articles

Effects of thermal aging on microstructure and hardness of stainless steel weld-overlay claddings of nuclear reactor pressure vessels

Takeuchi, Tomoaki; Kakubo, Yuta*; Matsukawa, Yoshitaka*; Nozawa, Yasuko*; Toyama, Takeshi*; Nagai, Yasuyoshi*; Nishiyama, Yutaka; Katsuyama, Jinya; Yamaguchi, Yoshihito; Onizawa, Kunio; et al.

Journal of Nuclear Materials, 452(1-3), p.235 - 240, 2014/09

 Times Cited Count:22 Percentile:5.64(Materials Science, Multidisciplinary)

Microstructures and hardness of stainless steel weld overlay cladding of reactor pressure vessels subjected to the thermal aging at 400 $$^{circ}$$C for 100-10,000 h were investigated using atom probe tomography and nanoindentation technique. The Cr concentration fluctuation in the $$delta$$-ferrite phase caused by spinodal decomposition rapidly progressed by the 100 h aging while NiSiMn clusters increased in number density at 2,000 h and coarsened at 10,000 h. The hardness of the $$delta$$-ferrite phase also rapidly increased at the short aging time. The Cr concentration fluctuation and the hardness were in good correlation with the degree of the Cr concentration fluctuation rather than the formation of the NiSiMn clusters. These results strongly suggested that the dominant factor of the hardening of the $$delta$$-ferrite phase by the thermal aging was Cr spinodal decomposition.

Journal Articles

Effect of neutron irradiation on the mechanical properties of weld overlay cladding for reactor pressure vessel

Tobita, Toru; Udagawa, Makoto; Chimi, Yasuhiro; Nishiyama, Yutaka; Onizawa, Kunio

Journal of Nuclear Materials, 452(1-3), p.61 - 68, 2014/09

 Times Cited Count:4 Percentile:56.06(Materials Science, Multidisciplinary)

To investigate the changes in the mechanical properties of cladding materials irradiated with high neutron fluence, two types of cladding materials were fabricated using the submerged-arc welding and electroslag welding methods. The tensile tests, Charpy impact tests and fracture toughness tests were conducted before and after neutron irradiation with a fluence of 10$$^{20}$$ n/cm$$^{2}$$ at 290 $$^{circ}$$C. With neutron irradiation, the yield strength and ultimate strength increased, and the total elongation decreased. The Charpy upper-shelf energy was reduced and the ductile-to-brittle transition temperature was increased with neutron irradiation. There was no obvious decrease in the elastic-plastic fracture toughness (J$$_{Ic}$$) of the cladding materials at high neutron fluence. The tearing modulus decreased with neutron irradiation, and considerable low J$$_{Ic}$$ values were observed at high temperatures submerged-arc-welded cladding materials.

Journal Articles

Study on structural integrity assessment of reactor pressure vessel based on three-dimensional thermal-hydraulics and structural analyses

Katsuyama, Jinya; Katsumata, Genshichiro; Onizawa, Kunio; Watanabe, Tadashi*; Nishiyama, Yutaka

Proceedings of 2014 ASME Pressure Vessels and Piping Conference (PVP 2014) (DVD-ROM), 9 Pages, 2014/07

For structural integrity assessment on reactor pressure vessel (RPV) of pressurized water reactor during the pressurized thermal shock (PTS) events, temperature of coolant water and heat transfer coefficient between coolant water and RPV are dominant factors. These values can be determined on the basis of thermal-hydraulics (TH) analysis simulating PTS events. Using these values, structural integrity assessment of RPV is performed by thermal-structural analysis, e.g. loading that affects the crack initiation and propagation is evaluated. In this study, we performed the TH and thermal-structural analyses using three-dimensional model of cold-leg, downcomer and RPV to assess loading conditions during the PTS more accurate. We obtained the loading histories at the reactor core region of RPV where a crack is postulated in the structural integrity assessment. Through the comparison between analysis results and current evaluation method, conservatism of current method will be discussed.

Journal Articles

Estimation of through-wall cracking frequency of RPV under PTS events using PFM analysis method for identifying conservatism included in current Japanese code

Osakabe, Kazuya*; Masaki, Koichi*; Katsuyama, Jinya; Katsumata, Genshichiro; Onizawa, Kunio

Proceedings of 2014 ASME Pressure Vessels and Piping Conference (PVP 2014) (DVD-ROM), 7 Pages, 2014/07

The structural integrity of reactor pressure vessel (RPV) during pressurized thermal shock events is judged to be maintained unless the stress intensity factors at the crack tip is smaller than fracture toughness $$K$$$$_{Ic}$$ based on deterministic approach in the current Japanese code. Application of a probabilistic fracture mechanics (PFM) analysis method for the structural reliability assessment of RPVs has become attractive recently, because uncertainties of several parameters can be incorporated rationally. According to the PFM analysis method in the U.S., through-wall cracking frequencies (TWCFs) are estimated. In this study, in order to identify the conservatism in the current code, PFM analyses on TWCF have been performed for certain model of RPVs. The result shows that the current assumption in JEAC 4206-2007 is conservative as compared with realistic conditions. Effects of variation of PTS transients on crack initiation frequency and TWCF have been also discussed.

Journal Articles

A Round robin propgram of master curve evaluation using miniature C(T) specimens, 3; Comparison of $$T_{0}$$ under various selections of temperature conditions

Yamamoto, Masato*; Kimura, Akihiko*; Onizawa, Kunio; Yoshimoto, Kentaro*; Ogawa, Takuya*; Mabuchi, Yasuhiro*; Viehrig, H.-W.*; Miura, Naoki*; Soneda, Naoki*

Proceedings of 2014 ASME Pressure Vessels and Piping Conference (PVP 2014) (DVD-ROM), 7 Pages, 2014/07

The Master Curve (MC) approach for the fracture toughness reference temperature To is expected to be a powerful tool to ensure the reliability of long-term used RPV steels. In order to get sufficient number of data for the MC approach related to the present surveillance program for RPVs, the use of miniature specimens is important. The test technique for the miniature specimens (Mini-CT) of 4 mm thick had been verified the basic applicability of MC approach by means of Mini-CT for the determination of fracture toughness of typical Japanese RPV steels. A round robin (RR) program was organized to assure the robustness of the technique. As the third step of RR program, blinded tests were carried out. Precise material information was not provided to the participants. From the results obtained, the scatter range in $$T_{0}$$ was within the acceptable scatter range specified in the testing standard. The selection of testing temperature seems to give limited effect like that in larger specimens.

Journal Articles

Benchmark analysis on probabilistic fracture mechanics analysis codes concerning multiple cracks and crack initiation in aged piping of nuclear power plants

Li, Y.; Osakabe, Kazuya*; Katsumata, Genshichiro; Katsuyama, Jinya; Onizawa, Kunio; Yoshimura, Shinobu*

Proceedings of 2014 ASME Pressure Vessels and Piping Conference (PVP 2014) (DVD-ROM), 8 Pages, 2014/07

Multiple cracks in the same welded joints have been detected in piping systems of nuclear power plants. Therefore, structural integrity assessments considering multiple cracks and crack initiation in aged piping have been important. Probabilistic fracture mechanics (PFM) is a rational methodology in structural integrity assessment of aged piping in nuclear power plants. Two PFM codes, PASCAL-SP and PRAISE-JNES, have been improved or developed in Japan for the structural integrity assessment considering the age related degradation mechanisms of pipes. In this paper, a benchmark analysis was conducted considering multiple cracks and crack initiation, in order to confirm their reliability and applicability. Based on the numerical investigation in consideration of important influence factors such as crack number, crack location, crack distribution and crack detection probability of in-service inspection, it was concluded that the analysis results of these two codes are in good agreement.

Journal Articles

Improvement of probabilistic fracture mechanics analysis code for reactor piping considering large earthquakes

Yamaguchi, Yoshihito; Katsuyama, Jinya; Udagawa, Makoto; Onizawa, Kunio; Nishiyama, Yutaka; Li, Y.*

Proceedings of 22nd International Conference on Nuclear Engineering (ICONE-22) (DVD-ROM), 8 Pages, 2014/07

272 (Records 1-20 displayed on this page)