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Nakamura, Hironobu; Shimizu, Yasuyuki; Makino, Risa; Mukai, Yasunobu; Ishiyama, Koichi; Kurita, Tsutomu; Ikeda, Atsushi*; Yamaguchi, Katsuhiro*
Proceedings of INMM 57th Annual Meeting (Internet), 9 Pages, 2016/07
Regarding the Integrated Safeguards (IS) in Japan, the implementation of IS has been started on September 2004, and the concept has been introduced to the JNC-1 facilities since August 2008. Then, random interim inspection with short notice and reducing person-days of inspection (PDI) was introduced instead of traditional scheduled IIV in order to improve deterrence of the nuclear material diversion with timeliness goal. And it was agreed that it should be evaluated and reviewed because RII was designed when inter-campaign. In JAEA, we decided to restart PCDF campaign to reduce potential safety risks of reprocessing facilities. To adopt the RII scheme to the process operation in campaign, JAEA proposed a new scheme to JSGO and IAEA without increasing PDI and reducing detection probability. As a result of the discussion, it was agreed and successfully introduced since March 2014. The new scheme for PCDF consists of scheduled inspection (fixed-day RII), reduction of estimated material for the verification, implementation of remote monitoring with data provision, improvement of operational status check list, introduction of NRTA and MC&A data declaration with timeliness. Though the operator's workloads for information provision were increased, we could manage to balance IS requirement with implementation of our operation successfully. This contribution was helped to safeguards implementation and our operation for 2 years.
Makino, Risa; Swinhoe, M. T.*; Suzuki, Hisanori; Ikeda, Atsushi*; Menlove, H. O.*; Shimizu, Yasuyuki; Nakamura, Hironobu
Kaku Busshitsu Kanri Gakkai (INMM) Nihon Shibu Dai-35-Kai Nenji Taikai Rombunshu (Internet), 9 Pages, 2015/01
The Inventory Verification Sample system (INVS) is a non-destructive assay system for samples to quantify the Pu amount in Pu nitrate solutions and MOX with 31.3% of counting efficiency. It has been used for IAEA verification measurement for many years at the Plutonium Conversion Development Facility for the samples taken at the timing of PIV etc. as a partial defects verification system (uncertainty: about 3-5%). If the measurement uncertainty can be improved (to 
1%), it is expected that the usage can be extended to the operator's own measurements in MC&A to reduce the number of destructive analyses. In order to improve the measurement uncertainty for solution samples, after optimization of detector parameter and sample position, we conducted 3 different types of calibration method that is passive calibration curve, known-
and multiplicity method to achieve the target uncertainty. To perform calibration and control the measurement quality, MOX fuel pellets with known Pu amount are fabricated and used. In the range of concentration of typical solution samples, we could confirm good correlations between measured doubles and 
Pu effective mass in the three methods. Especially, it was confirmed that the conventional calibration curve method could meet our target uncertainty (1%).
Makino, Risa; Ishiyama, Koichi; Kimura, Takashi; Yamazaki, Katsuyuki; Nakamura, Hironobu; Ikeda, Atsushi*; Yamaguchi, Katsuhiro*
Kaku Busshitsu Kanri Gakkai (INMM) Nihon Shibu Dai-33-Kai Nenji Taikai Rombunshu (Internet), 9 Pages, 2012/10
The Integrated Safeguards (IS) applied to JNC-1 site began from August, 2008. At that time, the inspection scheme was replaced from Interim Inventory Verification (IIV) with scheduled day to Random Interim Inspection (RII) with short notice in order to strengthen deterrent against diversion of nuclear materials. To satisfy the requirements for IS in Tokai reprocessing facilities including Tokai Reprocessing Plant and Plutonium Conversion Development Facility, we have cooperated inspectorates to establish remote monitoring systems, and to introduce RII smoothly. Though the new inspection scheme requires additional efforts which contain keeping a certain number of operators every day and declaration of interim inventory lists in a short time, the whole inspection days per year during inter-campaign could be decreased to about 60% by comparing with the conventional IIV. This paper reports the effects after introduction of RII and the future tasks to be discussed regarding RII during campaign from the operator's standpoint.
WAKASUGI, Keiichiro; Miyahara, Kaname; Makino, Hitoshi; Ishiguro, Katsuhiko; Sawamura, Hidenori*; Neyama, Atsushi*; Nishimura, Kazuya*
JNC TN8400 2003-022, 84 Pages, 2003/11
JNC-TN8400-2003-022.pdf:9.58MB
The radiation dose from the vitrified waste which is the same specification set in the Reference Case of the second progress report (H12 report) was evaluated quantitatively taking into account of the shield of the canister and the overpack. In order to understand the feature of radiation dose from the vitrified waste in terms of shielding, the thickness of the concrete shield to decrease less than safety standard for a radiation controlled zone was evaluated. Main results are summarized as follows. (1)The effective dose rates in the case considering the vitrified waste and the canister decrease approximate 45 orders of magnitude during the period of 1,000 years after vitrification due to decay of short half-life radionuclides. The effective dose rate doesn't decrease from 1,000 to 10,000 years. (2)The effective dose rates at the outside of overpack in the case considering the vitrified waste, canister and overpack are smaller than those inside of overpack approximate 5 orders of magnitude during the period of 100 years due to shielding effect of the overpack. However this difference is relatively small after 100 years since the contribution of 
radiation to total effective dose rates decrease due to decay of fission products. (3)Excepting a few cases, the result using the old law (dose equivalent rate) is larger than the result using the new law (effective dose rate). However the difference between these results is less than factor of 1.2. (4)The thickness of the concrete shield required to attenuate the effective dose during the period of 50 years less than safety standard for a radiation controlled zone is calculated as approximate 0.8m1.5m. The important factors to determine the thickness of the concrete shield are the radiation in the case of vitrified waste and the canister, and the neutron radiation in the case of vitrified waste, canister and overpack.
Yui, Mikazu; Makino, Hitoshi; Ashida, Takashi; ; Ishiguro, Katsuhiko; Neyama, Atsushi*
PNC TN8410 92-161, 177 Pages, 1992/09
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Thermal Conduct., 18, p.95 - 104, 1985/00
no abstracts in English