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JAEA Reports

Prototype fast breeder reactor Monju; Its history and achievements

Tsuruga Comprehensive Research and Development Center

JAEA-Technology 2019-007, 159 Pages, 2019/07

JAEA-Technology-2019-007.pdf:19.09MB
JAEA-Technology-2019-007-high-resolution1.pdf:42.36MB
JAEA-Technology-2019-007-high-resolution2.pdf:33.56MB
JAEA-Technology-2019-007-high-resolution3.pdf:38.14MB
JAEA-Technology-2019-007-high-resolution4.pdf:48.82MB
JAEA-Technology-2019-007-high-resolution5.pdf:37.61MB

This report summarizes the history and achievements of the prototype fast breeder reactor Monju. The development of Monju started in 1968 as a prototype reactor following the experimental fast reactor Joyo. The development covers all the activity related to the fast reactor; plant design, mockup tests, construction, operation, and plant management. This report summarizes the history and achievements for 11 technical areas: history and principal achievements, design and construction, operation test, plant safety, core physics, fuel, plant system, sodium technology, materials and mechanical design, plant management, and trouble management.

Journal Articles

Effect of local plastic component on crack opening displacement and on J-integral of a circumferential penetrated crack

Machida, Hideo*; Arakawa, Manabu*; Wakai, Takashi

Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 8 Pages, 2019/05

This paper describes the effect of local plastic component on J-integral and crack opening displacement (COD) evaluation of a circumferential penetrated crack, applicable to the leak before break (LBB) assessment for sodium cooled fast reactor (SFR) components. J-integral COD evaluation methods are generally formulated as a summation of elastic and plastic components, and so far many evaluation formulae based on these two components have been proposed. However, strictly, the plastic component consists of local plastic and fully plastic components. Many of the conventional evaluation methods often consider only the fully plastic component as the plastic component. The reason for this is that the effect of the local plastic component is much smaller than that of the fully plastic component excluding materials with extremely small work hardening. In contrast, for materials with high yield stress and small work hardening, such as modified 9Cr-1Mo steel which is one of the candidate materials for SFR piping, the effect of the local plastic component on J-integral and COD cannot be ignored. Therefore, the authors propose formulae taking the effect of local plastic component on J-integral and COD into account, based on finite element analysis (FEA) results, so that it is easy to apply to crack evaluation. The formulae will be employed in the guidelines on LBB assessment for SFR components published from Japan Society of Mechanical Engineers (JSME).

Journal Articles

Study on creep damage assessment method for Mod.9Cr-1Mo steel by sampling creep testing with thin plate specimen

Kanayama, Hideyuki*; Hiyoshi, Noritake*; Ogawa, Fumio*; Kawabata, Mie*; Ito, Takamoto*; Wakai, Takashi

Zairyo, 68(5), p.421 - 428, 2019/05

This study presents creep damage assessment method for Mod. 9Cr-1Mo steel by sampling creep testing with thin plate specimen. Tensile creep rupture tests were performed using three different sizes of specimen under two different test environments to verify the creep testing with the thin plate specimen. Time to rupture of Mod. 9Cr-1Mo steel using three different sizes were almost same. In addition, there was no effect of environment on time to rupture. Pre-damaged thin plate specimens were machined from a bulk specimen's gage section that pre-damage test was performed with. Pre-damage based on life fraction rule were 8%, 16% and 25%. No effect of the process of machining pre-damaged specimen on time to rupture was confirmed by verification tests in same test condition as pre-damage test. Stress acceleration creep rupture tests were performed to estimate creep damage assessment. Creep damage assessment by stress acceleration creep rupture tests was sufficiently accurate estimate. Creep damage assessments by Vickers hardness and lath width were compared with the assessment by stress acceleration creep rupture tests to study applicability of these methods.

Journal Articles

Ultra-high temperature creep rupture and transient burst strength of ODS steel claddings

Yano, Yasuhide; Sekio, Yoshihiro; Tanno, Takashi; Kato, Shoichi; Inoue, Toshihiko; Oka, Hiroshi; Otsuka, Satoshi; Furukawa, Tomohiro; Uwaba, Tomoyuki; Kaito, Takeji; et al.

Journal of Nuclear Materials, 516, p.347 - 353, 2019/04

 Times Cited Count:0 Percentile:100(Materials Science, Multidisciplinary)

9Cr-ODS steel claddings consisting of tempered martensitic matrix, showed prominent creep rupture strength at 1000 $$^{circ}$$C, which surpassed that of heat-resistant austenitic steels although creep rupture strength of tempered martensitic steels is generally lower than that of austenitic steels at high temperatures. The measured creep rupture strength of 9Cr-ODS steel claddings at 1000 $$^{circ}$$C was higher than that from extrapolated creep rupture trend curves formulated using data at temperatures from 650 to 850 $$^{circ}$$C. This superior strength seemed to be owing to transformation of the matrix from the $$alpha$$-phase to the $$gamma$$-phase. The transient burst strengths for 9Cr-ODS steel were much higher than those for 11Cr-ferritic/martensitic steel (PNC-FMS). Cumulative damage fraction analyses suggested that the life fraction rule can be used for the rupture life prediction of 9Cr-ODS steel and PNC-FMS claddings in the transient and accidental events with a certain accuracy.

Journal Articles

Improvement of penetrate crack length evaluation method for LBB assessment of sodium-cooled fast reactor components

Wakai, Takashi; Machida, Hideo*; Arakawa, Manabu*

Nippon Kikai Gakkai 2018-Nendo Nenji Taikai Koen Rombunshu (DVD-ROM), 5 Pages, 2018/09

According to the fitness for service code of Sodium-Cooled fast Reactor (SFR), the volumetric tests as in-service inspection can be replaced with continuous leak monitoring, where the Leak Before Break (LBB) is demonstrated, because the primary stress caused by internal pressure is not significant in SFR components. Basically, if the detectable crack length and the penetrated crack length are sufficiently smaller than the unstable critical crack length, it can be concluded that LBB is successfully demonstrated. The authors had already proposed a simplified method to calculate the penetrated crack length both of the circumferential and axial cracks in the pipe as a function of pipe geometry, fatigue crack growth characteristics and loading conditions. However, some problems in the method have been pointed out in the process of the reviewing by the JSME code committee. This study describes an improved method to calculate the penetrated crack length.

Journal Articles

Development of a crack opening displacement assessment procedure considering change of compliance at a crack part in thin wall pipes made of modified 9Cr-1Mo steel

Wakai, Takashi; Machida, Hideo*; Arakawa, Manabu*; Yanagihara, Seiji*; Suzuki, Ryosuke*; Matsubara, Masaaki*

Proceedings of 26th International Conference on Nuclear Engineering (ICONE-26) (Internet), 9 Pages, 2018/07

This paper studies crack opening displacement (COD) evaluation methods used in Leak-Before-Break (LBB) assessment of Sodium cooled Fast Reactor (SFR) pipe. For SFR pipe, the continuous leak monitoring will be adopted as an alternative to a volumetric test of the weld joints under conditions that satisfy LBB. The sodium pipes are made of ASME Gr.91 (modified 9Cr-1Mo steel). Thickness of the pipes is small, because the internal pressure is very small. Modified 9Cr-1Mo steel has a relatively large yield stress and small work hardening coefficient comparing to the austenitic stainless steels which are currently used in the conventional plants. In order to assess the LBB behavior of the sodium pipes made of modified 9Cr-1Mo steel, the coolant leak rate from a through wall crack must be estimated properly. Since the leak rate is strongly related to the crack opening displacement (COD), an appropriate COD assessment method must be established to perform LBB assessment. However, COD assessment method applicable for SFR pipes - having thin wall thickness and made of small work hardening material - has not been proposed yet. Thus, a COD assessment method applicable to such a pipe was proposed in this study. In this method, COD was calculated by classifying the components of COD; elastic, local plastic and fully plastic. In addition, the verification of this method was performed by comparing with the results of a series of four-point bending tests using modified 9Cr-1Mo steel pipe having a circumferential through wall notch. As a result, in some cases, COD were over-estimated especially for large cracks. Although the elastic component of COD is still over-estimated for large cracks, leak evaluation from small cracks is much more important in LBB assessment. Therefore, this study recommends that only the elastic component of COD should be adopted in LBB assessment of SFR pipes.

Journal Articles

Proposal of simplified J-integral evaluation method for a through wall crack in SFR pipe made of Mod.9Cr-1Mo steel

Wakai, Takashi; Machida, Hideo*; Arakawa, Manabu*; Kikuchi, Koichi*

Proceedings of ASME Symposium on Elevated Temperature Applications of Materials for Fossil, Nuclear, and Petrochemical Industries, 7 Pages, 2018/04

A simplified J-integral evaluation method applicable to unstable failure analysis in Leak Before Break (LBB) assessment of Sodium-cooled Fast Reactor (SFR) in Japan was proposed. Mod.9Cr-1Mo steel is supposed to be a candidate material for the coolant systems of SFR in Japan. This steel has relatively high yield strength and poor fracture toughness comparing to those of conventional austenitic stainless steels. In addition, SFR pipe has small thickness and large diameter. As a J-integral evaluation method for circumferential through-wall crack in a cylinder, EPRI has proposed a fully plastic solution method. However, the geometry of SFR pipe and material characteristics of Mod.9Cr-1Mo steel exceed the applicable range of EPRI's method. Therefore, a series of elastic, elasto-plastic and plastic finite element analyses (FEA) were performed for a pipe with a circumferential through-wall crack to propose a J-integral evaluation method applicable to such loading conditions. J-integrals obtained from the FEA were resolved into elastic, local plastic and fully plastic components. Each component was expressed as a function of analytical parameter, such as pipe geometries, crack size, material characteristics and so on. As a result, a simplified J-integral evaluation method was proposed. The method enables to conduct 2 parameter failure analysis using J-integral without any fracture mechanics knowledge.

Journal Articles

Study of the calculation method for the elastic follow-up coefficient by inelastic analysis

Watanabe, Sota*; Kubo, Koji*; Okajima, Satoshi; Wakai, Takashi

Nippon Kikai Gakkai M&M 2017 Zairyo Rikigaku Kanfarensu Koen Rombunshu (Internet), p.581 - 585, 2017/10

no abstracts in English

Journal Articles

Study on the seismic buckling evaluation method for Mod. 9Cr-1Mo steel cylindrical vessel

Okafuji, Takashi*; Miura, Kazuhiro*; Sago, Hiromi*; Murakami, Hisatomo*; Kubo, Koji*; Sato, Kenichiro*; Wakai, Takashi; Shimomura, Kenta

Nippon Kikai Gakkai M&M 2017 Zairyo Rikigaku Kanfarensu Koen Rombunshu (Internet), p.591 - 595, 2017/10

no abstracts in English

Journal Articles

Thermodynamic study of gaseous CsBO$$_{2}$$ by Knudsen effusion mass spectrometry

Nakajima, Kunihisa; Takai, Toshihide; Furukawa, Tomohiro; Osaka, Masahiko

Journal of Nuclear Materials, 491, p.183 - 189, 2017/08

 Times Cited Count:2 Percentile:78(Materials Science, Multidisciplinary)

One of the main chemical forms of cesium in the gas phase during severe accidents of light water reactor is expected to be cesium metaborate, CsBO$$_{2}$$, by thermodynamic equilibrium calculation considering reaction with boron. But accuracy of the thermodynamic data of gaseous metaborate, CsBO$$_{2}$$(g), has been judged as poor quality. Thus, Knudsen effusion mass spectrometric measurement of CsBO$$_{2}$$ was carried out to obtain reliable thermodynamic data. The evaluated values of standard enthalpy of formation of CsBO$$_{2}$$(g), $$Delta$$$$_{f}$$H$$^{circ}$$$$_{298}$$(CsBO$$_{2}$$,g), by the 2nd and 3rd law treatments are -700.7$$pm$$10.7 kJ/mol and -697.0$$pm$$10.6 kJ/mol, respectively, and agree with each other within the errors, which suggests our data are reliable. Further, it was found that the existing data of the Gibbs energy function and the standard enthalpy of formation agreed well with the values evaluated in this study, which indicates the existing thermodynamic data are also reliable.

Journal Articles

Thermal fatigue test on dissimilar welded joint between Gr.91 and 304SS

Wakai, Takashi; Kobayashi, Sumio; Kato, Shoichi; Ando, Masanori; Takasho, Hideki*

Proceedings of 25th International Conference on Nuclear Engineering (ICONE-25) (CD-ROM), 7 Pages, 2017/07

This paper describes a thermal fatigue test on a structural model with a dissimilar welded joint. In the present design of JSFR, there may be dissimilar welded joints between ferritic and austenitic steels especially in IHX and SG. Creep-fatigue is one of the most important failure modes in JSFR components. However, the creep-fatigue damage evaluation method has not been established for dissimilar welded joint. To investigate the evaluation method, structural test will be needed for verification. Therefore, a thermal fatigue test on a thick-wall cylinder with a circumferential dissimilar welded joint between Mod.9Cr-1Mo steel and 304SS was performed. Since the coefficients of thermal expansion of these steels were significantly different, buttering layer of Ni base alloy was installed between them. After the completion of the test, deep cracks were observed at the HAZ in 304SS, as well as at HAZ in Mod.9Cr-1Mo steel. There were many tiny surface cracks in BM of 304SS. According to the fatigue damage evaluation based on the finite element analysis results, the largest fatigue damage was calculated at HAZ in 304SS. Large fatigue damage was also estimated at BM of 304SS. Fatigue cracks were observed at HAZ and BM of 304SS in the test, so that analytical results are in a good agreement with the observations. However, though relatively small fatigue damage was estimated at HAZ in Mod.9Cr-1Mo steel, deep fatigue cracks were observed in the test. To identify the cause of such a discrepancy between the test and calculations, we performed a series of finite element analyses. Some metallurgical investigations were also performed.

Journal Articles

Preparation for a new experimental program addressing core-material-relocation behavior during severe accident with BWR design conditions; Conduction of preparatory tests applying non-transfer-type plasma heating technology

Abe, Yuta; Sato, Ikken; Ishimi, Akihiro; Nakagiri, Toshio; Nagae, Yuji

Proceedings of 24th International Conference on Nuclear Engineering (ICONE-24) (DVD-ROM), 7 Pages, 2016/06

A new experimental program using non-transfer type plasma heating is under consideration in JAEA to clarify the uncertainty on core-material relocation (CMR) behavior of BWR. In order to confirm the applicability of this new technology, authors performed preparatory plasma heating tests using small-scale test pieces (107 mm $$times$$ 107 mm $$times$$ 222 mmh). Based on these preliminary results, an excellent perspective in terms of applicability of the non-transfer plasma heating technology to the SA (Severe Accident) experimental study was obtained. Furthermore, JAEA is preparing for the next step intermediate-scale preparatory tests in 2016 using ca. 50 rods and a control blade that would not only confirm its technical applicability, but also some insights relevant to the issue on CMR itself.

Journal Articles

Collapse evaluation of double notched stainless pipes subjected to combined tension and bending

Suzuki, Ryosuke*; Matsubara, Masaaki*; Yanagihara, Seiji*; Morijiri, Mitsugu*; Omori, Atsushi*; Wakai, Takashi

Procedia Materials Science, 12, p.24 - 29, 2016/00

 Times Cited Count:2 Percentile:28.79

In this study, the plastic collapse strength of asymmetry multiple circumferential notched stainless steel pipes subjected to combined axial tension and bending is investigated experimentally and is compared with the theoretical plastic collapse strength. In addition, the potential is discussed for the simplification of structural integrity evaluation of multiple cracked piping. The integrity of the asymmetry multiple circumferential notched stainless steel pipes subjected to combined axial tension and bending can be evaluated conservatively using the theoretical plastic collapse strength for the pipe with multiple notches calculated based on the elastic-perfectly plastic model.

Journal Articles

A Study on evaluation method of penetrate crack length for LBB assessment of fast reactor pipes

Wakai, Takashi; Machida, Hideo*; Sato, Kenichiro*

Nippon Kikai Gakkai M&M 2015 Zairyo Rikigaku Kanfarensu Koen Rombunshu (Internet), 3 Pages, 2015/11

This paper describes a through-wall crack length evaluation procedure applicable to Leak Before Break (LBB) assessment of Japan Sodium cooled Fast Reactor (JSFR) pipes made of Mod.9Cr-1Mo steel. In LBB assessment of JSFR pipes, it is required to calculate virtual through-wall crack length, though the crack growth is quite small under design condition. Generally, it is known that the through-wall crack length depends on loading condition, namely the load ratio between tensile and bending and that the length under pure bending load condition is largest. This study proposes a simplified method to evaluate the through-wall crack length both for axial and circumferential cracks as a function of load ratio and fatigue crack growth characteristics. Using the method, through-wall crack length can be predicted as far as we know the loading condition and material properties.

Journal Articles

J-integral evaluation method for a through wall crack in thin-walled large diameter pipes made of Mod.9Cr-1Mo steel

Wakai, Takashi; Machida, Hideo*; Arakawa, Manabu*; Sato, Kenichiro*

Nippon Kikai Gakkai 2015-Nendo Nenji Taikai Koen Rombunshu (DVD-ROM), 5 Pages, 2015/09

This paper describes a J-integral evaluation procedure applicable to unstable failure analysis for a circumferential through wall crack in a pipe. Japan Sodium cooled Fast Reactor (JSFR) pipes are made of Mod.9Cr-1Mo steel. The fracture toughness of the material is inferior to that of conventional austenitic stainless steels. In addition, JSFR pipe has small thickness and large diameter and displacement controlled load is predominant. Therefore, the load balance in such piping system changes by crack extension and 2 parameter method using J-integral is applicable to unstable failure analysis for the pipes under such conditions. As a J-integral evaluation method for circumferential through wall crack in a cylinder, EPRI has proposed a fully plastic solution method. However, the geometry of JSFR pipe and material characteristics of Mod.9Cr-1Mo steel exceed the applicable range of EPRI's method. Therefore, a series of elastic, elastoplastic and plastic finite element analyses (FEA) were performed for a pipe with a circumferential through-wall crack to establish a J-integral evaluation method applicable to such conditions. J-integrals obtained from the FEA were resolved into elastic, local plastic and fully plastic components. Each component was expressed as a function of analytical parameter, such as pipe geometries, crack size, material characteristics and so on. As a result, a simplified J-integral evaluation method was proposed.

Journal Articles

Requirements for fracture toughness to satisfy LBB behavior of a pipe made of high chromium steel

Machida, Hideo*; Wakai, Takashi; Sato, Kenichiro*

Nippon Kikai Gakkai 2015-Nendo Nenji Taikai Koen Rombunshu (DVD-ROM), 5 Pages, 2015/09

The volumetric test for piping in a sodium cooled fast reactor (SFR) is difficult from the poor accessibility. Detection of a crack, therefore, is difficult before its penetration of a pipe wall, an SFR has a strategy to detect sodium leakage from a through wall crack before fracture of a pipe. Plant safety is ensured by shutting down a plant as soon as possible to detect small quantity of sodium leakage even if a crack penetrates a pipe wall. Consequently, it is important to ensure establishment of leakage-before-break (LBB) in this strategy. Effects of fracture resistance curve on fracture strength of a cracked pipe made of high chromium steel (Mod. 9Cr-1Mo steel), which is one of the candidates for fast reactor piping material, are evaluated in this study; and requirements for fracture resistance curve to achieve the LBB were proposed.

Journal Articles

Development of electromagnetic non-destructive testing method for the inspection of heat exchanger tubes of Japan Sodium-cooled Fast Reactor, 2; Detection of flaws on the inner surface using electromagnetic waves

Sasaki, Kota*; Yusa, Noritaka*; Wakai, Takashi; Hashizume, Hidetoshi*

Electromagnetic Nondestructive Evaluation (XVIII), p.244 - 251, 2015/06

 Times Cited Count:0 Percentile:100

Journal Articles

Creep-fatigue tests of double-end notched bar made of Mod.9Cr-1Mo steel

Shimomura, Kenta; Kato, Shoichi; Wakai, Takashi; Ando, Masanori; Hirose, Yuichi*; Sato, Kenichiro*

Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 9 Pages, 2015/05

This paper describes experimental and analytical works to confirm that the design standard for SFR components sufficiently covers possible failure mechanisms. Creep-fatigue damage evaluation method in JSME design standard for SFR components has been constructed based on experiments and/or numerical analyses of conventional austenitic stainless steels, such as 304SS. Since the material characteristics of Mod.9Cr-1Mo steel are substantially different from those of austenitic stainless steels, it is required to verify the applicability of the design standards to the SFR components made of Mod.9Cr-1Mo steel. A series of uni-axial creep-fatigue tests were conducted using double-ended notch bar specimens made of Mod.9Cr-1Mo steel under displacement controlled condition with 30 minute holding. The curvature radii of the specimens were 1.6mm, 11.2mm and 40.0mm. The specimen having 1.6mm notch and 11.2mm notch failed from outer surface but the specimen having 40.0mm notch showed obvious internal crack nucleation. In addition, though total duration time of the creep-fatigue test was only 2,000 hours, a lot of creep voids and inter granular crack growth were observed. To clarify the cause of such peculiar failure, some additional experiments were performed, as well as some numerical analyses. We could point out that such a peculiar failure aspect might result from corresponding stress distribution in the cross section. As a result of a series of investigations, possible causes of such peculiar failure could be narrowed down. A future investigation plan was proposed to clarify the most significant cause.

Journal Articles

Thermal transient test and strength evaluation of a tubesheet structure made of Mod.9Cr-1Mo steel, 1; Test model design and experimental results

Ando, Masanori; Hasebe, Shinichi; Kobayashi, Sumio; Kasahara, Naoto*; Toyoshi, Akira*; Omae, Takahiro*; Enuma, Yasuhiro*

Nuclear Engineering and Design, 275, p.408 - 421, 2014/08

AA2013-0395.pdf:2.65MB

 Times Cited Count:1 Percentile:87.59(Nuclear Science & Technology)

To clarify the failure mode of a semispherical tubesheet structure originally designed for SG in the JSFR, a cyclic thermal loading test was performed using a tubesheet model test structure. The tubesheet model made of Mod.9Cr-1Mo steel was subjected to 1,873 cycles of severe thermal transient loads using a large-scale sodium loop, in which sodium heated to 600$$^{circ}$$C and 250$$^{circ}$$C was flowed repeatedly with periods for each transient of 2 and 1 h, respectively. After the test, the test model was inspected by PT. Then, observation using a SEM and hardness testing were performed. A thermal-hydraulic analysis was also performed to validate the measured temperature history during the thermal transient. Through these examinations and evaluation with thermal-hydraulic analysis, the manner of failure in the tubesheet under cyclic thermal loading is discussed.

Journal Articles

Thermal transient test and strength evaluation of a tubesheet structure made of Mod.9Cr-1Mo steel, 2; Creep-fatigue strength evaluation

Ando, Masanori; Hasebe, Shinichi; Kobayashi, Sumio; Kasahara, Naoto*; Toyoshi, Akira*; Omae, Takahiro*; Enuma, Yasuhiro*

Nuclear Engineering and Design, 275, p.422 - 432, 2014/08

AA2013-0396.pdf:1.44MB

 Times Cited Count:6 Percentile:43.72(Nuclear Science & Technology)

In this study, the strength of a tubesheet test model simulating a semispherical tubesheet structure subjected to cyclic thermal transients was evaluated using the finite element analysis (FEA). A test model made of Mod.9Cr-1Mo steel was subjected to 1,873 cycles of severe thermal transient loading using a large-scale sodium loop, in which elevated-temperature sodium at 600$$^{circ}$$C and 250$$^{circ}$$C was flowed repeatedly and kept at the final temperature for 2 and 1 h, respectively. Heat transfer analysis and stress analysis were performed using the sodium temperature data measured during the test. Then, the elastic and inelastic stress analysis results were used to investigate the failure mechanism by creep-fatigue damage and evaluate the failure strength. The evaluation based on the results of inelastic analysis estimated the number of cycles to failure within a factor of 3.

43 (Records 1-20 displayed on this page)