Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Nakazawa, Osamu; Takiya, Hiroaki; Murakami, Masashi; Donomae, Yasushi; Meguro, Yoshihiro
JAEA-Review 2023-012, 6 Pages, 2023/08
JAEA-Review-2023-012.pdf:0.93MB
The selection of back-end technology development issues to be prioritized and their schedule of the Japan Atomic Energy Agency (JAEA) have been put together as the "Strategic Roadmap for Back-end Technology Development." The results of questionnaires on development technologies (seeds) and technical issues (needs) within JAEA conducted in FY2022 were reflected in the selection. The issues were extracted from among those that match the seeds and needs, from the perspective of early implementation in the work front and the perspective of common issues, and nine themes were selected. We will build a cross-organizational implementation framework within JAEA and aim to implement the development results in the work front as well as social implementation.
Okazaki, Hiro; Sumi, Mika; Abe, Katsuo; Kageyama, Tomio; Nakazawa, Hiroaki
Proceedings of INMM 53rd Annual Meeting (CD-ROM), 9 Pages, 2012/07
Accurate and precise measurement of Pu and U is critically important for safeguards and is the premise of safeguards inspection by IAEA at Nuclear Fuel handling facilities. Quality control section of Plutonium Fuel Development Center (PFDC-lab) has been analyzing the Pu and U isotopic compositions by MS as well as its content by IDMS of the MOX fuel pellet and its source materials for the accountancy purposes. To maintain and improve analysis quality and analysis reliability, PFDC-lab decided to have accreditation of ISO/IEC 17025 (International standard for the General requirements for the competence of testing and calibration laboratories). PFDC-lab established analysis quality management system, acquired the knowledge and information of uncertainty evaluation under cooperative study with US DOE's NBL, ensured our analysis traceability and validates our methods to meet technical requirements. In conclusion, we have accredited for ISO/IEC 17025 in March 1st, 2010.
Sumi, Mika; Okazaki, Hiro; Abe, Katsuo; Kageyama, Tomio; Nakazawa, Hiroaki
Proceedings of INMM 53rd Annual Meeting (CD-ROM), 9 Pages, 2012/07
PFDC has been developing various technologies for MOX fuel production, and Pu and U isotopic composition and content of these samples are analyzed at PFDC-lab. For the analysis of MOX fuel material including raw material powders, PFDC-lab is accredited as the testing laboratory of ISO/IEC 17025 scoped for (1) Pu and U isotopic composition measurement in the nuclear fuel materials by MS and (2) Pu and U content measurement in the nuclear fuel materials by IDMS. This accreditation was granted in March 2010. While preparing for accreditation, the PFDC-lab established the capability to estimate measurement uncertainty in accordance with the GUM Guide and developed calculation program combining GUM-Workbench and Excel. This paper presents details of our quality control system and measurement uncertainty evaluation systems.
Sumi, Mika; Abe, Katsuo; Kageyama, Tomio; Nakazawa, Hiroaki; Kurosawa, Akira; Yamamoto, Masahiko; Mason, P.*; Neuhoff, J.*; Doubek, N.*; Balsley, S.*; et al.
Proceedings of INMM 51st Annual Meeting (CD-ROM), 9 Pages, 2010/07
Large Size Dried (LSD) spikes are currently used in many facilities in Japan (and around the world) for U and Pu accountancy analysis by Isotope Dilution Mass Spectrometry (IDMS). Because of the large quantity of plutonium standard materials are needed to support Japanese facilities for nuclear fuel cycle and expected difficulties in the long term supply and transport of Pu reference materials, JAEA decided to evaluate the possibility of using MOX stored at Plutonium Fuel Development Center (PFDC) as a source of Pu standard material for LSD spike preparation. At PFDC, Pu nitrate solution was prepared from MOX and two types of the LSD spikes were prepared. The samples of each spike were distributed for verification measurements to international and domestic laboratories. Details of the Pu make-up value evaluation, the LSD spike preparation and the evaluation of the verification results will be presented.
Sumi, Mika; Abe, Katsuo; Kageyama, Tomio; Nakazawa, Hiroaki; Katchi, Tomokazu*; Murakami, Yoshiki*; Hishi, Tomoyuki*; Ai, Hironobu*
Proceedings of INMM 50th Annual Meeting (CD-ROM), 9 Pages, 2009/07
Currently, many laboratories who measure Pu and U concentration by isotope dilution mass spectrometry (IDMS) use Large Size Dried (LSD) spikes, which contain both Pu and U in individual vials. Plutonium Fuel Development Center (PFDC) prepared LSD spikes for MOX samples and has been used while LSD spikes prepared at inspection laboratories and commercially supplied are mainly aimed to measure input solution for reprocessing. Difficulties of importing reference materials are increasing though the needs of Pu reference materials are increasing. Stable securing Pu reference material is essential for facility operation and it is considered to be important to acquire technique to prepare domestic Pu reference material. Pu were prepared from MOX powder at PFDC and used for LSD spike preparation. Practical tests were performed with JNFL. Experience of preparation and utilization of LSD spike for MOX, consideration of certification method for MOX-Pu will be explained also in this paper.
Saito, Kosuke; Maeda, Seiichiro; Higuchi, Masashi*; Takano, Mitsuhiro*; Nakazawa, Hiroaki
JAEA-Technology 2006-035, 76 Pages, 2006/06
JAEA-Technology-2006-035.pdf:5.25MB
Because of the revision on the standardized strength of the ODS steel, the previous design study of MONJU demonstrative core has been obliged to be reconsidered. For economical advantages, only a 127 pins-bundle core was selected to be redesigned. For the sake of cladding endurance, the ratio of cladding thickness to outer diameter was reset incrementally followed by the determination of the basic specification of a pin. Notwithstanding some deterioration thanks to the reduction of a fuel volume fraction, the prospect in neutronics was obtained. Coolant flow distribution design which was based on power distribution was successfully carried out without overheating cladding. Average burn-up of 150 GWd/t and 380 days-long operational period per cycle are to be attained, and the designed core can thermally afford to receive test fuels. The study has necessity to be advanced extensively for the purpose of materialization according to the circumstances of MONJU in future.
Ikusawa,Yoshihisa; Kikuchi, Keiichi; Ozawa, Takayuki; Nakazawa, Hiroaki; Isozaki,Takao*; Nagayama, Masahiro*
JNC TN8410 2005-012, 113 Pages, 2005/08
JNC-TN8410-2005-012.pdf:15.2MB
The E09 fuel assembly was irradiated in the FUGEN from February 1990 to January 1997. The fuel assembly was the highest burn-up assembly in FUGEN and the pellet peak burn-up reached about 48 GWd/t. The E09 fuel assembly was transported to Japan Atomic Energy Research Institute (JAERI) Tokai in 2001. Post Irradiation Examinations (PIE) were started in July 2001, and all PIE items were completed by March 2005. The irradiation behavior of E09 MOX fuel was evaluated from the result of PIE. The major results are as follows; The integrity of E09 fuel assembly and fuel rods was confirmed. The corrosion behavior of ATR MOX fuel cladding was similar to that of LWR-UO2 fuel cladding. The central void was observed in outer ring samples irradiated with the maximum linear power over 45kW/m. A porous fine structure, similar to the rim structure seen in LWR-UO
pellet, was observed in the circumferential region of MOX pellet and around the plutonium-rich spots. The MOX fuel properties irradiated up to ~48 GWd/t, which are pellet swelling, thermal conductivity, pellet melting temperature and diffusivity of fission gas, were similar to LWR-UO fuel properties. These results will be used for CANDU-OPTION program, which is one of Russian surplus weapon plutonium disposition programs with AECL in Canada, and available for LWR plutonium recycle program in Japan.
Abe, Tomoyuki; Maeda, Seiichiro; Nakazawa, Hiroaki
Proceedings of 13th International Conference on Nuclear Engineering (ICONE-13), 0 Pages, 2005/05
Japan Nuclear Cycle Development Institute (JNC) developed plutonium and uranium mixed oxide (MOX) fuels for an advanced thermal reactor (ATR) for a flexible utilization of plutonium. JNC made endeavors to obtain well-homogenized MOX pellets by a ball mill mixing method with a variety of raw powders, including MOX powder by a microwave-heating denitration process. A total of 772 MOX fuel assemblies were utilized in the ATR prototype reactor
Ikusawa,Yoshihisa; Kikuchi, Keiichi; Ozawa, Takayuki; Nakazawa, Hiroaki; Abe, Tomoyuki; Isozaki,Takao*; Nagayama, Masahiro*
JNC TN8410 2004-008, 106 Pages, 2004/10
JNC-TN8410-2004-008.pdf:25.96MB
The "E09" was irradiated in the FUGEN from February 1990 to January 1997, and its average burn-up reached 37.7GWd/t at the end of irradiation. In order to be irradiated up to high burn-up, this fuel assembly had the design improved by applying the fissile content with axial distribution, four UO- GdO
fuel rods located with MOX fuel rods and so on. The E09 fuel assembly had been cooled in the FUGEN spent fuel pool for four years after irradiation.After that, it was transported to Japan Atomic Energy Research Institute (JAERI) Tokai in 2001.Post Irradiation Examinations (PIE) were started in July 2001 at Reactor Fuel Examination Facility in JAERI, and a part of destructive examinations(Puncture examination, Ceramography, Metallography and alpha-autoradiography) were completed in March 2003. The destructive examinations will be completed by December 2004.In this report, the data obtained from destructive examinations completed in March 2003 were summarized, and the evaluation results of irradiation performance of MOX fuel and cladding were discussed. Consequently, the MOX fuel rod integrity during irradiation was confirmed from the result of the destructive PIE. These results will be used for CANDU-OPTION program, which is one of Russian surplus weapon plutonium disposition programs with AECL in Canada, and available for LWR plutonium recycle program in Japan.
Tachibana, Yukio; Sawahata, Hiroaki; Iyoku, Tatsuo; Nakazawa, Toshio
Nuclear Engineering and Design, 233(1-3), p.89 - 101, 2004/10
Times Cited Count：10 Percentile：55.72(Nuclear Science & Technology)no abstracts in English
Saito, Kenji; Nakagawa, Shigeaki; Hirato, Yoji; Kondo, Makoto; Sawahata, Hiroaki; Tsuchiyama, Masaru*; Ando, Toshio*; Motegi, Toshihiro; Mizushima, Toshihiko; Nakazawa, Toshio
JAERI-Tech 2004-042, 26 Pages, 2004/04
JAERI-Tech-2004-042.pdf:1.16MB
The reactor control system of HTTR is composed of the reactor power control system, the reactor inlet coolant temperature control system, the primary coolant flow rate control system and so on. The reactor control system of HTTR achieves reactor power 30MW, reactor outlet coolant temperature 850
C, reactor inlet coolant temperature 395C under the condition that primary coolant flow rate is fixed. In the Rise-to-Power Test, the performance test of the reactor inlet coolant temperature control system was carried out in order to confirm the control capability of this control system. This report shows the test results of performance test. As a result, the control parameters, which can control the reactor inlet coolant temperature stably during the reactor operation, were successfully selected. And it was confirmed that the reactor inlet coolant temperature control system has the capability of controlling the reactor inlet coolant temperature stably against any disturbances on the basis of operational condition of HTTR.
Hirato, Yoji; Saito, Kenji; Kondo, Makoto; Sawahata, Hiroaki; Motegi, Toshihiro; Tsuchiyama, Masaru*; Ando, Toshio*; Mizushima, Toshihiko; Nakazawa, Toshio
JAERI-Tech 2004-037, 33 Pages, 2004/04
JAERI-Tech-2004-037.pdf:4.08MB
HTTR (High Temperature Engineering Test Reactor) was operated from May 6th, 2003 to June 18th, 2003 to obtain operation data in parallel loaded operation mode and in safety demonstration tests. Operated with the reactor power at 60% of the rated power on May 21st, HTTR was automatically scrammed by a signalof "Primary coolant flow rate of the Primary Pressurized Water Cooler (PPWC): Low". The cause of the shutdown was the primary gas circulator (A) automatically stopped. The primary coolant flow rate of the PPWC decresed and reached the scram set value due to the gas circulator stop. As a result of investigation, it became clear that the cause of the gas circulator stop was malfunction of an auxiliary relay which monitored electric power of a circuit breaker in power line of the gas circulator. The cause of malfunction was deterioration of the relay under high temperature condition because the relay was installed beside an electric part which was heated up by electricity.
Ikusawa,Yoshihisa; Kikuchi, Keiichi; Nakazawa, Hiroaki; Abe, Tomoyuki; Isozaki,Takao*; Nagayama, Masahiro*
JNC TN8410 2003-015, 251 Pages, 2004/01
JNC-TN8410-2003-015.pdf:16.07MB
The FUGEN High Burn-up MOX Fuel Assembly E09 was developed for high burn-up fuel of DATR. The E09 MOX fuel assembly was irradiated at the FUGEN from February 1990 to January 1997, and its average burn-up reached 37.7GWd/t. In order to be irradiated up to high burn-up, they had the design improved by applying the fissile content with axial distribution, four UO2-Gd2O3 fuel rods and so on. The E09 fuel assembly had been cooled in the FUGEN spent fuel pool for four years after irradiation. After that, it was transported to Japan Atomic Energy Research Institute (JAERI) Tokai Research Establishment in 2001. Post Irradiation Examinations (PIE) were started in July 2001 at Reactor Fuel Examination Facility in JAERI, and a part of destructive examinations(Puncture examination ,Metallography and Alpha Autoradiography) were completed in March 2003. In this report, the data from destructive examinations will be summarized, and evaluation results of irradiation performance will be discussed. The integrity of fuel assembly during irradiation was confirmed in the destructive PIE.
Ozawa, Takayuki; Nakazawa, Hiroaki; Abe, Tomiyuki
JNC TY8410 2003-002, 40 Pages, 2003/06
JNC-TY8410-2003-002.pdf:63.99MB
None
; Nakazawa, Hiroaki; ; Nagayama, Masahiro*
JNC TY8410 2003-001, 47 Pages, 2003/04
None
Ikusawa,Yoshihisa; ; Nakazawa, Hiroaki; ; *; Nagayama, Masahiro*
JNC TN8410 2003-004, 233 Pages, 2003/02
JNC-TN8410-2003-004.pdf:24.5MB
The irradiation examination on the fuel that was fabricated for the demonstration advanced thermal reactor (DATR), had been conducted in SGHWR in the U.K. and in FUGEN. The FUGEN High Burn-up MOX Fuel Assembly "E09" was developed for high burn-up fuel of DATR. The "E09" MOX fuel assembly was irradiated at the FUGEN from February 1990 to January 1997, and its average burn-up reached 37.7GWd/t. In order to be irradiated up to high burn-up, they had the design improved by applying the fissile content with axial distribution, four UO-GdO fuel rods and so on. The E09 fuel assembly had been cooled in the FUGEN spent fuel pool for four years after irradiation. After that, it was transported to Japan Atomic Energy Research Institute (JAERI) Tokai Research Establishment in 2001. Post Irradiation Examinations (PIE) were started in July 2001 at Reactor Fuel Examination Facility in JAERI, and the nondestructive examinations were completed in March 2002. The destructive examinations have been started in April 2002, and will be completed by March 2004. In this report, the data from nondestructive examinations will be summarized, and evaluation results of irradiation performance will be discussed. The integrity of fuel assembly during irradiation was confirmed in the nondestructive PIE. These results will be available for LWR MOX fuel program in Japan and for CANDU-OPTION program, which is one of Russian surplus weapon plutonium disposition programs with AECL in Canada.
Maeda, Seiichiro; Abe, Tomoyuki; Nakazawa, Hiroaki
GENES4/ANP2003, CD-ROM, Paper1150, 8p., 8 Pages, 2003/00
None
; Nakazawa, Hiroaki; ; Nagayama, Masahiro*
JNC TY8410 2002-002, 46 Pages, 2002/10
JNC-TY8410-2002-002.pdf:0.18MB
None
Suzuki, Masaaki*; Sekiguchi, Hidetoshi*; Akatsuka, Hiroshi*; Goto, Takanobu*; Osugi, Takeshi*; Kobayashi, Hiroaki; Nakazawa, Osamu
JNC TY8400 2002-016, 158 Pages, 2002/03
JNC-TY8400-2002-016.pdf:15.93MB
None
Ozawa, Takayuki; Nakazawa, Hiroaki; ; Nagayama, Masahiro*
JNC TY8410 2001-002, 46 Pages, 2002/02
JNC-TY8410-2001-002.pdf:4.62MB
None
