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Journal Articles

Mitigation of cavitation damage in J-PARC mercury target vessel

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 ($$sim$$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.

Journal Articles

Investigation of irradiated properties of extended burnup TRISO fuel

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.

Journal Articles

Irradiation performance of sodium-bonded control rod for the fast breeder reactor

Sasaki, Shinji; Maeda, Koji; Furuya, Hirotaka*

Journal of Nuclear Science and Technology, 55(3), p.276 - 282, 2018/03

 Times Cited Count:3 Percentile:45.99(Nuclear Science & Technology)

Journal Articles

Irradiation test and post irradiation examination of the high burnup HTGR fuel

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.

Journal Articles

Early-in-life fuel restructuring behavior of Am-bearing MOX fuels

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.

Journal Articles

Effect of neutron induced reactions of neodymium-147 and 148 on burnup evaluation

Suyama, Kenya; Mochizuki, Hiroki*

Journal of Nuclear Science and Technology, 42(7), p.661 - 669, 2005/07

 Times Cited Count:14 Percentile:70.37(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.

JAEA Reports

Annual report of JMTR, No.18, FY2003; April 1, 2003 - March 31, 2004

Department of JMTR

JAERI-Review 2004-029, 100 Pages, 2005/01

JAERI-Review-2004-029.pdf:19.03MB

no abstracts in English

JAEA Reports

Study on the prediction accuracy of nuclide generation and depletion with JENDL

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$$_{2}$$ 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.

JAEA Reports

Development of once-through type densitometer

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.

Journal Articles

Validation of integrated burnup code system SWAT2 by the analyses of isotopic composition of spent nuclear fuel

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.

JAEA Reports

Development of pellet melting temperature measuring technique; Melting temperature measuring technique for small sample

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$$_{2}$$ pellet.

JAEA Reports

Annual report of JMTR, No.17, FY2002; April 1, 2002 - March 31, 2003

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 $$gamma$$-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.

Journal Articles

Improvements to SFCOMPO; A Database on isotopic composition of spent nuclear fuel

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.

JAEA Reports

Irradiation test of high burnup coated fuel particles for High Temperature Gas-cooled Reactor; 91F-1A Sweap gas capsule irradiation test

Sawa, Kazuhiro; Tobita, Tsutomu*; Takahashi, Masashi; Saito, Takashi; Iimura, Katsumichi; Yokouchi, Iichiro; Serizawa, Hiroyuki; Sekino, Hajime; Ishikawa, Akiyoshi

JAERI-Research 2001-043, 52 Pages, 2001/09

JAERI-Research-2001-043.pdf:14.92MB

no abstracts in English

Journal Articles

Melting temperature of high burn-up UO$$_{2}$$ pellet

Harada, Katsuya; Nakata, Masahito; Yasuda, Ryo; Nishino, Yasuharu; Amano, Hidetoshi

HPR-356, 11 Pages, 2001/00

no abstracts in English

JAEA Reports

Revised SWAT; The Integrated burnup calculation code system

Suyama, Kenya; Kiyosumi, Takehide*; Mochizuki, Hiroki*

JAERI-Data/Code 2000-027, 88 Pages, 2000/07

JAERI-Data-Code-2000-027.pdf:4.08MB

no abstracts in English

Journal Articles

Irradiation experiment of ZrC-coated fuel particles for high-temperature gas-cooled reactors

Minato, Kazuo; Ogawa, Toru; Sawa, Kazuhiro; Ishikawa, Akiyoshi; Tomita, Takeshi; Iida, Shozo; Sekino, Hajime

Nuclear Technology, 130(3), p.272 - 281, 2000/06

 Times Cited Count:63 Percentile:96.39(Nuclear Science & Technology)

no abstracts in English

JAEA Reports

Annual report of JMTR, No.13; FY1998 (April 1, 1998 - March 31, 1999)

Department of JMTR

JAERI-Review 2000-007, p.94 - 0, 2000/05

JAERI-Review-2000-007.pdf:4.32MB

no abstracts in English

JAEA Reports

Annual report of JMTR, No.12; JMTR operation and technical development, April 1, 1997 - March 31, 1998

Department of JMTR

JAERI-Review 99-006, 79 Pages, 1999/03

JAERI-Review-99-006.pdf:3.72MB

no abstracts in English

JAEA Reports

Annual report of JMTR, FY1997, April 1, 1997 - March 31, 1998

Department of JMTR

JAERI-Review 99-005, 79 Pages, 1999/03

JAERI-Review-99-005.pdf:3.6MB

no abstracts in English

40 (Records 1-20 displayed on this page)