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Tsai, T.-H.; Sasaki, Shinji; Maeda, Koji
Journal of Nuclear Science and Technology, 60(6), p.715 - 723, 2023/06
Times Cited Count：1 Percentile：16.36(Nuclear Science & Technology)Naoe, Takashi; Kinoshita, Hidetaka; Wakui, Takashi; Kogawa, Hiroyuki; Haga, Katsuhiro
JAEA-Technology 2022-018, 43 Pages, 2022/08
JAEA-Technology-2022-018.pdf:7.84MB
In the liquid mercury target system for the pulsed spallation neutron source of Materials and Life science experimental Facility (MLF) at the Japan in the Japan Proton Accelerator Research Complex (J-PARC), cavitation that is generated by the high-energy proton beam-induced pressure waves, resulting severe erosion damage on the interior surface of the mercury target vessel. The erosion damage is increased with increasing the proton beam power, and has the possibility to cause the leakage of mercury by the penetrated damage and/or the fatigue failure originated from erosion pits during operation. To achieve the long term stable operation under high-power proton beam, the mitigation technologies for cavitation erosion consisting of surface modification on the vessel interior surface, helium gas microbubble injection, double-walled beam window structure has been applied. The damage on interior surface of the vessel is never observed during the beam operation. Therefore, after the target operation term ends, we have cut out specimen from the target nose of the target vessel to inspect damaged surface in detail for verification of the cavitation damage mitigation technologies and lifetime estimation. We have developed the techniques of specimen cutting out by remote handling under high-radiation environment. Cutting method was gradually updated based on experience in actual cutting for the used target vessel. In this report, techniques of specimen cutting out for the beam entrance portion of the target vessel in high-radiation environment and overview of the results of specimen cutting from actual target vessels are described.
Naoe, Takashi; Kinoshita, Hidetaka; Kogawa, Hiroyuki; Wakui, Takashi; Wakai, Eiichi; Haga, Katsuhiro; Takada, Hiroshi
JPS Conference Proceedings (Internet), 28, p.081004_1 - 081004_6, 2020/02
The beam window of the mercury target vessel in J-PARC is severely damaged by the cavitation. The cavitation damage is a crucial factor to limit lifetime of the target because it increases with the beam power. Therefore, mitigating cavitation damage is an important issue to operate the target stably for long time at 1 MW. At J-PARC, to mitigate the cavitation damage: gas microbubbles are injected into mercury for suppressing pressure waves, and double-walled structure with a narrow channel of 2 mm in width to form high-speed mercury flow (
4m/s) has been adopted. After operation, the beam window was cut to inspect the effect of the cavitation damage mitigation on inner wall. We optimized cutting conditions through the cold cutting tests, succeeding in cutting the target No.2 (without damage mitigation technologies) smoothly in 2017, and target No.8 with damage mitigation technologies. In the workshop, progress of cavitation damage observation for the target vessel will be presented.
Shaimerdenov, A.*; Gizatulin, S.*; Kenzhin, Y.*; Dyussambayev, D.*; Ueta, Shohei; Aihara, Jun; Shibata, Taiju
Proceedings of 9th International Topical Meeting on High Temperature Reactor Technology (HTR 2018) (USB Flash Drive), 6 Pages, 2018/10
The Institute of Nuclear Physics of the Republic of Kazakhstan (INP) conducts an irradiation test and post-irradiation examinations (PIEs) of the high-temperature gas-cooled reactor (HTGR) fuel and materials to develop the extend burnup fuel up to 100 GWd/t-U collaboratively with the Japan Atomic Energy Agency (JAEA) under projects in a frame of the International Science and Technology Centre (ISTC). Cylindrical fuel compact specimens consisting of newly-designed TRISO (tri-structural isotropic) coated fuel particles and a matrix made of graphite material were manufactured in Japan. An irradiation test of the fuel specimens using a helium-gas swept capsule designed and constructed in the INP has been performed up to 100 GWd/t-U in the WWR-K research reactor by April 2015. In the next stage, PIEs with the irradiated fuel specimens have been started in February 2017 as a new ISTC project. Several PIE technologies by non-destructive and destructive techniques with irradiated fuel compacts were developed by the INP. This report presents the developed technologies and interim results of the PIE for high burning TRISO fuel.
Sasaki, Shinji; Maeda, Koji; Furuya, Hirotaka*
Journal of Nuclear Science and Technology, 55(3), p.276 - 282, 2018/03
Times Cited Count：4 Percentile：26.28(Nuclear Science & Technology)Mano, Akihiro; Yamaguchi, Yoshihito; Katsuyama, Jinya; Li, Y.
Proceedings of Asian Symposium on Risk Assessment and Management 2017 (ASRAM 2017) (USB Flash Drive), 12 Pages, 2017/11
As a rational inspection methodology, risk informed in-service inspection (RI-ISI) has been widely utilized in in-service inspections of nuclear power plants (NPPs) in several countries. In some of NPPs, an RI-ISI methodology developed by Westinghouse Owners Group (WOG) was applied. As a part of RI-ISI process, extent of examination for important piping segments are determined through the comparisons of leak frequencies with its target value based on the industrial piping leak experiences. The leak frequencies for segments are used as a numerical factor for planning examination based on WOG methodology, and can be evaluated through analyses on the basis of probabilistic fracture mechanics (PFM). In Japan Atomic Energy Agency (JAEA), we have developed a PFM analysis code PASCAL-SP for evaluating leak and rupture probabilities or frequencies of welds in piping of light water reactors taking crack initiation and propagation due to aging degradation mechanisms such as fatigue into consideration. Also, evaluation models of probability of crack detection by non-destructive examination considering the crack type, crack depth and performance of examination team is incorporated in PASCAL-SP. In this study, we investigated the applicability of PASCAL-SP into planning of examination considering the effects of repair methodology, performance of inspection team, and examination time. On the basis of analysis results, it was found that examination plans can be reasonably determined by using PASCAL-SP under several conditions, and it was concluded that the PFM is very effective tools in RI-ISI.
Ueta, Shohei; Aihara, Jun; Shaimerdenov, A.*; Dyussambayev, D.*; Gizatulin, S.*; Chakrov, P.*; Sakaba, Nariaki
Proceedings of 8th International Topical Meeting on High Temperature Reactor Technology (HTR 2016) (CD-ROM), p.246 - 252, 2016/11
In order to examine irradiation performance of the new Tri-structural Isotropic (TRISO) fuel for the High Temperature Gas-cooled Reactor (HTGR) at the burnup around 100 GWd/t, a capsule irradiation test was conducted by WWR-K research reactor in the Institute of Nuclear Physics (INP) of Kazakhstan. The irradiated TRISO fuel was designed by Japan Atomic Energy Agency (JAEA) and fabricated in basis of the HTTR fuel technology in Japan. The fractional release of fission gas from the fuel during the irradiation shows good agreement with the predicted one released from as-fabricated failed TRISO fuel. It was suggested that unexpected additional fuel failure would not occur during the irradiation up to 100 GWd/t. In addition, the post-irradiation examination (PIE) with the irradiated fuel is planned to qualify TRISO fuel integrity and upgrade HTGR fuel design for further burnup extension.
Tanaka, Kosuke; Sasaki, Shinji; Katsuyama, Kozo; Koyama, Shinichi
Transactions of the American Nuclear Society, 113(1), p.619 - 621, 2015/10
In order to evaluate the microstructural change behavior of Am-MOX fuels at the initial stage of irradiation, detailed investigations using image analysis were performed on X-ray Computed Tomography (X-ray CT) images and on ceramographs from fuels irradiated in both B11 and B14.
Onodera, Junichi; Kurihara, Ryoichi; Seki, Yasushi
Gijutsushi Ichiji Shiken No Keiko To Taisaku; Denki Denshi, Joho Kogaku, Genshiryoku, Hoshasen Bumon Hen, p.137 - 178, 2005/08
In 2004, "Nuclear Power and Radiation Section" has been established in the National Qualification of Professional Engineer. A reference material to help those who will take the preliminary examination is produced.
Suyama, Kenya; Mochizuki, Hiroki*
Journal of Nuclear Science and Technology, 42(7), p.661 - 669, 2005/07
Times Cited Count：15 Percentile：68.99(Nuclear Science & Technology)Burnup is important value for criticality safety evaluation of spent nuclear fuel. Nd-148 method is one of most important method to evaluate the burnup of post irradiation examination (PIE) samples, and well known that it has good accuracy. However, the evaluated burnup values could be perturbed by the neutron capture reaction of Nd-147 and Nd-148. And in the analysis of PIE data from PWR, the calculation results of Nd-148 have approximately more than 1% deviation from experiment. In this study, the contribution of neutron capture reaction of Nd-147 and Nd-148 to Nd-148 amount are discussed. Especially for Nd-147 contribution, it is shown that the current evaluated cross section of Nd-147 is not supported and the new evaluation is consistent with the analysis of PIE data. Possible perturbed amount of Nd-148 by both reactions is less than 0.7% for normal reactor operation condition, and it is approximately 0.1% for 30 GWd/t (BWR) and 40 GWd/t (PWR). Finally, we confirm again that Nd-148 method is good evaluation method.
Department of JMTR
JAERI-Review 2004-029, 100 Pages, 2005/01
JAERI-Review-2004-029.pdf:19.03MB
no abstracts in English
Okumura, Keisuke; Oki, Shigeo*; Yamamoto, Munenari*; Matsumoto, Hideki*; Ando, Yoshihira*; Tsujimoto, Kazufumi; Sasahara, Akihiro*; Katakura, Junichi; Matsumura, Tetsuo*; Aoyama, Takafumi*; et al.
JAERI-Research 2004-025, 154 Pages, 2005/01
JAERI-Research-2004-025.pdf:19.46MB
This report summarizes the activity (FY2000-2003) of Working Group (WG) on Evaluation of Nuclide Generation and Depletion under Subcommittee on Nuclear Fuel Cycle of Japanese Nuclear Data Committee. In the WG, analyses of Post Irradiation Examinations have been carried out for UO
and MOX fuels irradiated in PWRs, BWRs and FBRs, and for actinide samples irradiated in fast reactors, by using ORIGEN or more detailed calculation codes with their libraries based on JENDL-3.2, JENDL-3.3 and other foreign nuclear data files. From these results, current prediction accuracy and problems for evaluation of nuclide generation and depletion are discussed. Furthermore, this report covers other products of our activity; development of the ORIGEN libraries for PWR, BWR and FBR based on JENDL-3.3, study on introduction of neutron spectrum index to ORIGEN calculations, and results of questionnaire survey on desirable accuracy of ORIGEN calculations.
Onozawa, Atsushi; Kushida, Teruo; Kanazawa, Hiroyuki
JAERI-Tech 2004-061, 39 Pages, 2004/11
JAERI-Tech-2004-061.pdf:8.64MB
The swelling observed on irradiated fuels is caused by the accumulation of fission products and irradiation defects. The swelling ratio is changed along with radius region in the pellet due to burn up difference caused by that of neutron flux. To investigate the swelling behavior at the small area of the pellet, it is needed to measure the density of fuel fragments picked from an irradiated pellet. In this circumstance, once-through type densitometer was developed to measure the density of the small irradiated specimen precisely and to handle the samples easily with remote control systems. Several kinds of metallic and ceramic standard specimens are prepared to investigate the dependence of the sample weight, density and porosity on the accuracy. The results of characteristic examination using these specimens indicate that this densitometer has enough accuracy. In addition, some parts of this apparatus are controlled by motor drive units, which made it possible to measure the density full-automatically.
Onizawa, Kunio; Shibata, Katsuyuki; Kato, Daisuke*; Li, Y.*
JSME International Journal, Series A, 47(3), p.486 - 493, 2004/07
The probabilistic fracture mechanics (PFM) analysis code PASCAL has been developed in JAERI. This code can evaluate the conditional probabilities of crack initiation and fracture of a reactor pressure vessel (RPV) under transient conditions such as pressurized thermal shock (PTS). Based on the temperature and stress distributions in the vessel wall for four PTS sequences in a typical 3-loop PWR, parametric PFM analyses are performed using PASCAL on the variables such as pre-service inspection model, crack geometry, fracture toughness curve and irradiation embrittlement prediction equation. The results showed that the good perfomance inspection model had a significant effect on the fracture probability and reduced it by more than 3 orders of magnitude. The fracture probability calculated by the fracture toughness estimation method in Japan was about 2 orders of magnitude lower than that by the USA method. It was found that the treatment of a semi-elliptical crack in PASCAL reduced the conservatism in a conventional method that it is transformed into an infinite length crack.
Suyama, Kenya; Mochizuki, Hiroki*; Okuno, Hiroshi; Miyoshi, Yoshinori
Proceedings of International Conference on Physics of Fuel Cycles and Advanced Nuclear Systems; Global Developments (PHYSOR 2004) (CD-ROM), 10 Pages, 2004/04
This paper provides validation results of SWAT2, the revised version of SWAT, which is a code system combining point burnup code ORIGEN2 and continuous energy Monte Carlo code MVP, by the analysis of post irradiation examinations (PIEs). Some isotopes show differences of calculation results between SWAT and SWAT2. However, generally, the differences are smaller than the error of PIE analysis that was reported in previous SWAT validation activity, and improved results are obtained for several important fission product nuclides. This study also includes comparison between an assembly and a single pin cell geometry models.
Harada, Katsuya; Nakata, Masahito; Harada, Akio; Nihei, Yasuo; Yasuda, Ryo; Nishino, Yasuharu
JAERI-Tech 2004-034, 13 Pages, 2004/03
JAERI-Tech-2004-034.pdf:0.69MB
The Department of Hot Laboratories has been aiming the establishment of the melting temperature measuring technique for small samples obtained from the micro-region of irradiated fuel pellet. Due to the modification of the shape of tungsten capsule contained sample and the improvement of the detection method for melting temperature from indistinct thermal arrest point owing to small sample, it is possible to determine the melting temperature of small sample and to utilize effectively for the irradiated fuel pellet by using the existing apparatus. This paper describes the technique of the melting temperature measurement for small sample and the experimental results by using tantalum, molybdenum, hafnium oxide and un-irradiated UO pellet.
Department of JMTR
JAERI-Review 2003-041, 89 Pages, 2004/01
JAERI-Review-2003-041.pdf:6.32MB
During the FY2002 (April 2002 to March 2003), the Japan Materials Testing Reactor (JMTR) was operated in three operation cycles for studies on the irradiation-assisted stress corrosion cracking (IASCC) of internal structures of light water reactors, development of the fusion blanket, basic materials research, radioisotope production, and so on. The reactor was twice subjected to unscheduled shutdowns, once due to an error signal of the control-rod drive mechanism and afterwards due to leakage of primary coolant. Recovery and corrective actions were taken properly, such as improvements in related facilities and revision of operation manuals. Besides, the beryllium reflector frames were replaced, which are internal structures of the JMTR. Concerning development on techniques for irradiation utilization, noticeable progress was made in an evaluation method of neutron and 
-ray heating rates of materials in JMTR irradiation, in development and installation of a monitoring and control system of specimen temperatures for IASCC irradiation examinations, and in successful performance tests by a post-irradiation crack-growth testing device, which brought a prospect of comparative studies between in-situ and post-irradiation examinations. In studies on the fusion blanket, a basic process was established for production of tritium breeder pebbles, and irradiation techniques were developed for in-pile functional tests of a simulated blanket module. In addition, joint seminars were held with the Korea Atomic Energy Research Institute (KAERI) on techniques for irradiation and post-irradiation examinations. The present report summarizes the JMTR operation and related technical development in the Department of JMTR during the FY2002.
Suyama, Kenya; Nouri, A.*; Mochizuki, Hiroki*; Nomura, Yasushi*
JAERI-Conf 2003-019, p.890 - 892, 2003/10
Isotopic composition is one of the most relevant data to be used in the calculation of burnup of irradiated nuclear fuel. Since autumn 2002, the Organisation for Economic Co-operation and Development/Nuclear Energy Agency OECD/NEA) has operated a database of isotopic composition; SFCOMPO, initially developed in Japan Atomic Energy research Institute. This paper describes latest version of SFCOMPO and the future development plan in OECD/NEA.
Yachi, Shigeyasu; Sato, Tadashi; Suga, Shinichi*; Komuro, Yuichi; Uchida, Masaaki; Nakajima, Kunihisa; Nakamura, Jinichi; Amezawa, Hiroo; Omura, Hideaki*; Minato, Kazuo; et al.
JAERI-Review 2003-025, 162 Pages, 2003/09
JAERI-Review-2003-025.pdf:5.96MB
The report contains example of answers to the Problems of 31st(1999) to 35th(2003) Examinations for the Chief Engineer of Nuclear Fuel which were conducted as a national qualification examination. Brief explanations or references are given to some answers.
Nakahira, Masataka
Journal of Nuclear Science and Technology, 40(9), p.687 - 694, 2003/09
Times Cited Count：3 Percentile：25.38(Nuclear Science & Technology)The ITER vacuum vessel is a double-walled torus with large-sized quadrilateral ports and is required to provide a quite high degree of vacuum for deutrium - tritium fusion reaction. The vacuum vessel is required to install after assembled with toroidal field coils. From a radiological safety aspect, the vacuum vessel is functioned as a physical barrier to enclose radioactive materials. Therefore, construction of the vacuum vessel needs application of newly developed technologies on design, fabrication and examination. The technologies include design approach by finite element analysis, and partial penetration T welded joints to join ribs to outer shell. Several issues have to be resolved for applying those technologies to the vacuum vessel. This paper describes several newly developed technologies and key issues for applying to the vacuum vessel and then their resolutions.
