Intruder configurations in Ne at the transition into the island of inversion; Detailed structure study of Ne

Wang, H.*; Yasuda, Masahiro*; Kondo, Yosuke*; Nakamura, Takashi*; Tostevin, J. A.*; Ogata, Kazuyuki*; Otsuka, Takaharu*; Poves, A.*; Shimizu, Noritaka*; Yoshida, Kazuki; et al.

Physics Letters B, 843, p.138038_1 - 138038_9, 2023/08

https://doi.org/10.1016/j.physletb.2023.138038

Detailed -ray spectroscopy of the exotic neon isotope Ne has been performed using the one-neutron removal reaction from Ne. Based on an analysis of parallel momentum distributions, a level scheme with spin-parity assignments has been constructed for Ne and the negative-parity states are identified for the first time. The measured partial cross sections and momentum distributions reveal a significant intruder p-wave strength providing evidence of the breakdown of the N = 20 and N = 28 shell gaps. Only a weak, possible f-wave strength was observed to bound final states. Large-scale shell-model calculations with different effective interactions do not reproduce the large p-wave and small f-wave strength observed experimentally, indicating an ongoing challenge for a complete theoretical description of the transition into the island of inversion along the Ne isotopic chain.

Impact assessment for internal flooding in HTTR (High temperature engineering test reactor)

Tochio, Daisuke; Nagasumi, Satoru; Inoi, Hiroyuki; Hamamoto, Shimpei; Ono, Masato; Kobayashi, Shoichi; Uesaka, Takahiro; Watanabe, Shuji; Saito, Kenji

JAEA-Technology 2021-014, 80 Pages, 2021/09

JAEA-Technology-2021-014.pdf:5.87MB

In response to the new regulatory standards established in response to the accident at TEPCO's Fukushima Daiichi Nuclear Power Station in March 2011, measures and impact assessments related to internal flooding at HTTR were carried out. In assessing the impact, considering the characteristics of the high-temperature gas-cooled reactor, flooding due to assumed damage to piping and equipment, flooding due to water discharge from the system installed to prevent the spread of fire, and flooding due to damage to piping and equipment due to an earthquake. The effects of submersion, flooding, and flooding due to steam were evaluated for each of them. The impact of the overflow of liquids containing radioactive materials outside the radiation-controlled area was also evaluated. As a result, it was confirmed that flooding generated at HTTR does not affect the safety function of the reactor facility by taking measures.

Thermal fatigue test on dissimilar welded joint between Gr.91 and 304SS

Wakai, Takashi; Kobayashi, Sumio; Kato, Shoichi; Ando, Masanori; Takasho, Hideki*

Proceedings of 25th International Conference on Nuclear Engineering (ICONE-25) (CD-ROM), 7 Pages, 2017/07

https://doi.org/10.1115/ICONE25-66099

This paper describes a thermal fatigue test on a structural model with a dissimilar welded joint. In the present design of JSFR, there may be dissimilar welded joints between ferritic and austenitic steels especially in IHX and SG. Creep-fatigue is one of the most important failure modes in JSFR components. However, the creep-fatigue damage evaluation method has not been established for dissimilar welded joint. To investigate the evaluation method, structural test will be needed for verification. Therefore, a thermal fatigue test on a thick-wall cylinder with a circumferential dissimilar welded joint between Mod.9Cr-1Mo steel and 304SS was performed. Since the coefficients of thermal expansion of these steels were significantly different, buttering layer of Ni base alloy was installed between them. After the completion of the test, deep cracks were observed at the HAZ in 304SS, as well as at HAZ in Mod.9Cr-1Mo steel. There were many tiny surface cracks in BM of 304SS. According to the fatigue damage evaluation based on the finite element analysis results, the largest fatigue damage was calculated at HAZ in 304SS. Large fatigue damage was also estimated at BM of 304SS. Fatigue cracks were observed at HAZ and BM of 304SS in the test, so that analytical results are in a good agreement with the observations. However, though relatively small fatigue damage was estimated at HAZ in Mod.9Cr-1Mo steel, deep fatigue cracks were observed in the test. To identify the cause of such a discrepancy between the test and calculations, we performed a series of finite element analyses. Some metallurgical investigations were also performed.

Verification of alternative dew point hygrometer for CV-LRT in MONJU; Short- and long-term verification for capacitance-type dew point hygrometer (Translated document)

Ichikawa, Shoichi; Chiba, Yusuke; Ono, Fumiyasu; Hatori, Masakazu; Kobayashi, Takanori; Uekura, Ryoichi; Hashiri, Nobuo*; Inuzuka, Taisuke*; Kitano, Hiroshi*; Abe, Hisashi*

JAEA-Research 2017-001, 40 Pages, 2017/03

JAEA-Research-2017-001.pdf:5.19MB

In order to reduce the influence on a plant schedule of the MONJU by the maintenance of dew point hygrometers, The JAEA examined a capacitance type dew point hygrometer as an alternative dew point hygrometer for a lithium-chloride type dew point hygrometer which had been used at the CV-LRT in the MONJU. As a result of comparing a capacitance type dew point hygrometer with a lithium-chloride type dew point hygrometer at the CV-LRT (Atmosphere: nitrogen, Testing time: 24 hours), there weren't significant difference between a capacitance type dew point hygrometer and a lithium-chloride type dew point hygrometer. As a result of comparing a capacitance dew point hygrometer with a high-mirror-surface type dew point hygrometer for long term verification (Atmosphere: air, Testing time: 24 months), the JAEA confirmed that a capacitance type dew point hygrometer satisfied the instrument specification (2.04C) required by the JEAC4203-2008.

Investigation and evaluation of the cause about the loss of split pin which was set to pipe support system of the primary cooling system of Monju

Ichikawa, Shoichi; Kawanago, Sho; Nishio, Ryuichi; Wakimoto, Fumitsugu; Fujimura, Tomofumi; Kobayashi, Takanori; Sakamoto, Tsutomu

JAEA-Review 2015-009, 210 Pages, 2015/07

JAEA-Review-2015-009-01.pdf:60.82MB
JAEA-Review-2015-009-02.pdf:63.6MB
JAEA-Review-2015-009-03.pdf:66.2MB
JAEA-Review-2015-009-04.pdf:62.99MB

The loss of the retaining split pins (four pieces) for clevis pin were confirmed at the inspection of the pipe supports in the Monju prototype fast-breeder reactor in May, 2014. The split pins (two pieces) of ROD RESTRAINT and CONSTANT HANGER were fallen off. The split pins (two pieces) of MECHANICAL SNUBBER were broken at both ends of them. As a result of investigation, a dimple pattern was observed in a fracture surface of broken split pin. This observation result showed that fracture morphology is ductile fracture. A reproduction test, whether split pin was broken by loading the external force to the clevis pin, also gave the same fracture morphology. As the result of all cause investigation, the reason of the broken split pins is that the split pins were loaded shearing stress by the external force loaded to the clevis pin axial direction. The result of the cause investigation and a recurrence prevention measure of this trouble was be reported by this report.

Maintenance of the emergency power generators and overhaul of the gas turbine engines in HTTR

Inoi, Hiroyuki; Shimizu, Atsushi; Kameyama, Yasuhiko; Kobayashi, Shoichi; Shinozaki, Masayuki; Ota, Yukimaru; Kubo, Tsukasa*; Emori, Koichi

JAEA-Technology 2009-048, 48 Pages, 2009/10

JAEA-Technology-2009-048.pdf:5.0MB

The emergency power feeders of the High Temperature Engineering Test Reactor (HTTR) have gas turbine generators which are composed of gas turbin engines, generators and current breakers. The gas turbine generators have been overhauled and maintained to keep the performance. The maintenance technology was upgraded by improving their parts and surveillance method on the basis of the operational and maintenance experience. It can be clarified that the deterioration levels and the sudden deterioration timing are judged at an early stage by measuring the max exhaust temperature at the time of start in addition to check the starting time of the Gas Turbine Engines.

Report of investigation on malfunction of reserved shutdown system in HTTR

Hamamoto, Shimpei; Iigaki, Kazuhiko; Shimizu, Atsushi; Sawahata, Hiroaki; Kondo, Makoto; Oyama, Sunao; Kawano, Shuichi; Kobayashi, Shoichi; Kawamoto, Taiki; Suzuki, Hisashi; et al.

JAEA-Technology 2006-030, 58 Pages, 2006/03

JAEA-Technology-2006-030.pdf:10.69MB

During normal operation of High Temperature engineering Test Reactor (HTTR) in Japan Atomic Energy Agency (JAEA), the reactivity is controlled by the Control Rods (CRs) system which consists of 32 CRs (16 pairs) and 16 Control Rod Drive Mechanisms (CRDMs). The CR system is located in stand-pipes accompanied by the Reserved Shutdown System (RSS). In the unlikely event that the CRs fail to be inserted, the RSS is provided to insert BC/C pellets into the core. The RSS shall be designed so that the reactor should be held subcriticality from any operation condition by dropping in the pellets. The RSS consists of BC/C pellets, hoppers which contain the pellets, electric plug, driving mechanisms, guide tubes and so on. In accidents when the CRs cannot be inserted, an electric plug is pulled out by a motor and the absorber pellets fall into the core by gravity. A trouble, malfunction of one RSS out of sixteen, occurred during a series of the pre-start up checks of HTTR on February 21, 2005. We investigated the cause of the RSS trouble and took countermeasures to prevent the issue. As the result of investigation, the cause of the trouble was attributed to the following reason: In the motor inside, The Oil of grease of the multiplying gear flowed down from a gap of the oil seal which has been deformed and was mixed with abrasion powder of brake disk. Therefore the adhesive mixture prevented a motor from rotating.

Control rod position and temperature coefficients in HTTR power-rise tests; Interim report

Fujimoto, Nozomu; Nojiri, Naoki; Takada, Eiji*; Saito, Kenji; Kobayashi, Shoichi; Sawahata, Hiroaki; Kokusen, Shigeru

JAERI-Tech 2000-091, 49 Pages, 2001/03

JAERI-Tech-2000-091.pdf:2.0MB

no abstracts in English

Evaluation of Carburization and Decarburization Behavior of Fe-9Cr-Mo Ferritic Steels in Sodium Environment

Ito, Takushi; Kato, Shoichi; Aoki, Masanori; Yoshida, Hidekazu; Kobayashi, Toshimi; Wada, Yusaku

Journal of Nuclear Science and Technology, 29(4), p.367 - 377, 1992/04

https://doi.org/10.3327/jnst.29.367

None

Mock-up apparatus for preparation of plutonium fuel solution for criticality experiments at NUCEF

Sakurai, Satoshi; Hirata, Masaru; ; Usuda, Shigekazu; Abe, Jiro; ; Tachimori, Shoichi; ; Kurihara, Masayoshi; Kobayashi, Iwao

JAERI-M 90-059, 35 Pages, 1990/03

JAERI-M-90-059.pdf:1.38MB

no abstracts in English