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Suto, Masayoshi; Ichige, Satoshi; Isozaki, Kazunori
UTNL-R-0459, p.5_1 - 5_9, 2007/03
no abstracts in English
Isozaki, Kazunori; Ogawa, Toru; Nishino, Kazunari; Kaito, Yasuaki; Ichige, Satoshi; Sumino, Kozo; Suto, Masayoshi; Kawahara, Hirotaka; Suzuki, Toshiaki; Takamatsu, Misao; et al.
JNC TN9440 2005-003, 708 Pages, 2005/05
JNC-TN9440-2005-003.pdf:31.46MB
Periodic safety review (Review of the aging management) which consisted of Technical review on aging for the safety related structures, systems and components and Establish a long term maintenance program was carried out up to April 2005.1. Technical review on aging for the safety related structures, systems and components It was technically confirmed to prevent the loss of function of the safety related structures, systems and components due to aging phenomena, which (1) irradiation damage, (2) corrosion, (3) abrasion and erosion, (4) thermal aging, (5) creep and fatigue, (6) Stress Corrosion Cracking, (7) insulation deterioration and (8) general deterioration, under the periodic monitoring or renewal of them 2. Establish a long term maintenance program The long term maintenance program during JPY2005 to 2014 were established based on the technical review on aging for the safety related structures, systems and components. It was evaluated that the inspection or renewal based on the long term maintenance program, in addition to the spontaneous inspection long-term schedule of the long term voluntary inspection plan, could prevent the loss of function of the safety related structures, systems and components in future.
Sumino, Kozo; Ashida, Takashi; Kawahara, Hirotaka; Ichige, Satoshi; Isozaki, Kazunori; Nakai, Satoru
Proceedings of Operating Nuclear Facility Safety(2004ONFS),p204-216, p.204 - 216, 2004/11
None
Kawahara, Hirotaka; Isozaki, Kazunori; Ishii, Takayuki; Ichige, Satoshi; Nose, Shoichi; Sakaba, Hideo; Nakai, Satoru
JNC TN9410 2004-016, 106 Pages, 2004/06
JNC-TN9410-2004-016.pdf:8.47MB
A key part of the upgrade of the experimental fast reactor JOYO to the MK-III design was the replacement of the dump heat exchangers. MK-III function tests (SKS-1) of the new dump heat exchangers were carried out from August 27,2001 through September 13,2001. The major results of the function tests of the dump heat exchangers were as follows: (1) Air flow of the main blower with an inlet vane opening of 50% was confirmed to exceed the design rated flow of 7,700m3/min. It was also demonstrated that an inlet vane opening of 100% provides about 130% of the design rated flow. This is because the new DHX flow route has more low pressure loss than the design value. (2) Tests of the air flow of the main blower demonstrated that with a fully opened inlet damper a full opened outlet damper and an inlet vane opening of O% provides about 5% of the design rated flow. (3) Free flow coast down characteristics of the main blower achieved an inlet vane O% opening in an average of 7.9 seconds. Revolutions per minute of the main blower reached zero in an average of 8.7 seconds. The delay time from the opening of the vacuum contact breaker to the air flow decrease was approximately 1 second. This was a more conservative value than the 5 seconds assumed in design thermal transient analyses. (4) The loudest noise occurred with the main blower operating with a 25% inlet vane opening. At that time, the noise around the main blower was approximately 100dB, and in the surrounding monitoring area boundary, the noise was 50dB. This was confirmed to be within the standard of the Ibaraki prefectural ordinance. (5) Although the MK-III inlet vane and inlet damper drive unit was bigger than the MK-II unit, the accumulator tank was confirmed to provide sufficient volume during a compression air loss event.
Kawahara, Hirotaka; Kawahara, Hirotaka; Ichige, Satoshi; Isozaki, Kazunori; Nakai, Satoru
Proceedings of 12th International Conference on Nuclear Engineering (ICONE-12) (CD-ROM), 0 Pages, 2004/00
A recently completed major upgrade of the JOYO experimental sodium-cooled fast reactor, to the MK-III design, increased its irradiation capability approximately four times. 0ne major change was a 40% increase in thermal power to 140 MWt, which necessitated the replacement of the heat exchangers. Each of the two coolant loops includes an intermediate heat exchanger (IHX) and sodium pump in the primary system, and two dump heat exchangers (DHXs) and a pump in the secondary system. The heat transfer area of the finned tubes in each (air-cooled) DHX was doubled, compared to the old design, to achieve a 35 MWt rating, Major challenges in the replacement of secondary components, such as piping and DHX, were control of impurity ingress into the sodium system, and integrity assurance of the welding. Damage to existing components and systems was avoided during cutting and welding operations by taking measures to Prevent ingress of air into the sodium systems. The measures included use of seal b
Ishii, Takayuki; Isozaki, Kazunori; Ashida, Takashi; Minakawa, Satoru; Terakado, Tsuguo; Nogami, Hiroshi*; Kakurai, Katsuhiko*; Ueda, Soji*; Kawahara, Hirotaka; Ichige, Satoshi; et al.
JNC TN9410 2002-013, 86 Pages, 2002/11
JNC-TN9410-2002-013.pdf:68.29MB
The MK-III project has been proceeding to improve the irradiation capability of the experimental fast reactor JOY0. The MK-III project has three major purposes such as increase high neutron flux, improvement of plant availability factor and upgrading in irradiation techniques. Thermal output by core transformation for high neutron flux was increased from 100MWt to 140MWt. The main components in the cooling system such as IHX(Intermediate Heat Exchanger) and DHX(Dump Heat Exchanger) were replaced in MK-III modification in order to increase heat removal capability. Heat transfer capacity of IHX was increased from 50MWt/Unit to 70MWt/Unit and that of DHX was increased from 25MWt/Unit to 35MWt/Unit. These components replacement has been safety completed from october 30, 2000 through September 21, 2001. This report summarizes the way, results and safety measures about dismantling of no sodium adhered components and such as DHX blower and sodium cleaning of sodium components such as the DHX, the pipes connected with DHX and secondary side of IHX. Dismantling and sodium cleaning of secondary cooling system components were performed safely and efficiently as almost planned. The total amount of removed sodium was about 13.5kg.
Sumino, Kozo; Ichige, Satoshi; Fukami, Akihiro*; Maeda, Yukimoto; Suzuki, Soju
PNC TN9430 98-008, 40 Pages, 1998/09
The Raman Distributed Temperature Sensor (RDTS) based on the Raman Scattering Phenomena in the optical fiber is a system, which can easily measure the accurate temperature distribution. In order to evaluate the applicability of RDTS for FBR plant instrumentation, a temperature distribution measurement using RDTS was performed in the primary cooling system of JOYO. By using two optical fiber sensors, which were installed spirally around the primary piping, the temperature distribution on the primary piping was measured from the 30th through the 32nd duty cycle. In addition, the same test was carried out in the secondary cooling system of JOYO in order to test the measurement data from the primary cooling system. The main results were as follows; (1) The temperature data in the primary cooling system was acquired over 180EFPDs of operation at JOYO (accumulated dose : 3 
10
R). (2) The chracteristics of FTR in the high dose rate nuclear plant environment was confirmed. (3) The radiation induced temperature errors were calibrated succesfully by using thermocouple readings. The accuracy of the temperature after calibration was approximately 
3
C. (4) It was confirmed that diffent fiber and insulator settings on the piping cause temperature changes.
Suzuki, Kazuhisa; Yamanouchi, Sadamu; Tani, Satoshi; Ichige, Akio; Naito, Takeshi; Harada, Mamoru; Ito, Masahiko; Osugi, Shoichi*; Shibata, Kenichi*
PNC TN951 76-05, 70 Pages, 1976/03
The serpentine concentrate in which two kinds of cement were used as a binder was irradiated in JMTR to the fast neutron fluence -1.7
n/cm
(E
1Mev) and thermal neutron fluence -1.510
n/cm. Ave-rage irradiation temperature was estimated to be 200
. Prior to the postirradiation examination, effect of thermal history was investigated in out-of-pile test. The integrity of specimens were kept after irradiation. From the results of thermal history examination and the post-irradiation examination, it was clearly noticed that the changes occuring in dimensions, weight and compressive strength were caused mainly by the thermal effect. The change in the Young's modulus seemed to be caused by the effect of both heating and neutron irradiation. It wasconcluded that the ordinary portland cement with the serpentine concrete is better than that of almina cement.