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Futemma, Akira; Sanada, Yukihisa; Nagakubo, Azusa; Kawasaki, Yoshiharu*; Iwai, Takeyuki*; Hiraga, Shogo*; Haginoya, Masashi*; Matsunaga, Yuki*; Akutsu, Yuichiro*; Arai, Yoshinori*; et al.
JAEA-Technology 2023-027, 146 Pages, 2024/03
JAEA-Technology-2023-027.pdf:18.12MB
By the accident at Tokyo Electric Power Company's (TEPCO's) Fukushima Daiichi Nuclear Power Station (FDNPS), caused by tsunami triggered by the 2011 off the Pacific coast of Tohoku Earthquake, a large amount of radioactive material was released into the surrounding environment. After the accident, Airborne Radiation Monitoring (ARM) via manned helicopter has been applied as a method to quickly and extensively measure the distribution of radiation. Japan Atomic Energy Agency (JAEA) has continuously conducted ARM via manned helicopter around FDNPS. In this report, we summarize the results of the ARM around FDNPS in the fiscal year 2022, evaluate the changes of ambient dose rates and other parameters based on the comparison to the past ARM results, and discuss the causes of such changes. In order to contribute to improve the accuracy of ambient dose rate conversion, we analyzed the ARM data taking into account undulating topography, and evaluated the effect of this method. Furthermore, the effect of radon progenies in the air on the ARM was evaluated by applying the discrimination method to the measurement results.
Ogiyanagi, Jin; Hanawa, Satoshi; Suzuki, Motoe; Nagase, Fumihisa
Nuclear Engineering and Design, 253, p.77 - 85, 2012/12
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)Irradiation behavior of medium and high burnup BWR fuels during base-irradiation and power ramp test is analyzed by a fuel performance code FEMAXI-7. The calculated values such as fission gas release after the base-irradiation and a cladding diameter profile before and after the ramp test show a reasonable agreement with measured PIE data. It was also found that the code can reasonably predict the FGR at high temperature condition up to 1800
C of pellet center temperature by using the FGR model indigenous to the code and the enhanced value of the original Turnbull model of fission gas atoms diffusion constant. For the ridging deformation of the cladding before and after the ramp test, the local PCMI analysis with 2-D geometry in FEMAXI-7 gave a reasonable agreement with the PIE data. Thus, it is demonstrated that the FEMAXI-7 code can give an appropriate insight into the complicated thermal and mechanical interactions in medium and high burnup BWR fuel rods.
Hanawa, Satoshi; Ogiyanagi, Jin; Suzuki, Motoe
Journal of Nuclear Science and Technology, 49(5), p.516 - 525, 2012/05
Times Cited Count：2 Percentile：17.8(Nuclear Science & Technology)He pressurization effect on fission gas release (FGR) of BWR fuel rods under power transient conditions was analyzed by the fuel performance code FEMAXI-7. The experimental data provided to this study was obtained in the Halden reactor. Two rods were irradiated in the Halden reactor for 12 years in the IFA-409 as base-irradiation, then provided to the IFA-535 for power ramp tests to understand He-pressurization effect on fission gas release under power transient conditions, by adjusting internal pressure of the rods before power ramp test. FEMAXI-7 reasonably reproduced the experimental data of cladding elongation change and FGR behavior during the power ramp test. Based on the calculation results, the cause that apparent He-pressurization effect was not observed in the experiment was considered to be caused by insufficient gas communication during strong PCMI and gap thermal conductance by the solid-solid contact due to gap closure.
Ise, Hideo; Ogiyanagi, Jin; Nakamura, Jinichi; Sasajima, Hideo; Takasa, Akira; Hanawa, Satoshi; Kawaguchi, Yoshihiko; Chimi, Yasuhiro; Nishiyama, Yutaka; Nakamura, Takehiko
FAPIG, (180), p.22 - 25, 2010/02
Refurbishment of Japan Materials Testing Reactor (JMTR) is conducted in Japan Atomic Energy Agency (JAEA) in order to solve irradiation related issues for safe long-term operation of current light water reactors (LWRs) and development of advanced LWRs. JMTR will restart its operation in FY 2011. Manufacturing and installation of the irradiation test facilities on safety research of fuels and materials are also in progress. The outline of the fuels and materials irradiation test plan is described in this report.
Ogiyanagi, Jin; Chimi, Yasuhiro; Shimada, Sachio*; Nakamura, Takehiko; Abe, Katsuhiro*
Journal of Nuclear Science and Technology, 47(2), p.197 - 201, 2010/02
Times Cited Count：1 Percentile：9.99(Nuclear Science & Technology)The terminal solid solubility (TSS) of hydrogen during hydride dissolution/precipitation is determined by the differential scanning calorimetry technique for non-irradiated hafnium (Hf) that is used as control rods of light water reactors. The hydrogen concentration in hydrogenated Hf samples ranged from 27 to 300 wt ppm. The reliability of the TSS data obtained for Hf was confirmed by those for Zircaloy-2 (Zry-2) obtained in this study with the literature data, and best-fit equations for the obtained TSS curves for Hf are derived. The TSS for Hf at 573 K, which corresponds to an operational temperature of control rods in boiling water reactors, is in the range of 10-15 wt ppm, and is found to be 1/5-1/7 of the TSS for Zry-2.
Hanawa, Satoshi; Ogiyanagi, Jin; Inaba, Yoshitomo; Sasajima, Hideo; Nakamura, Jinichi; Nakamura, Takehiko
Proceedings of Top Fuel 2009 (DVD-ROM), p.350 - 356, 2009/09
In order to perform power transient tests of new design LWR's fuels, new power transient test capsules, the natural convection capsule and the forced convection capsule, are being developed. The natural convection capsule has relatively simple structure, and the test fuel rod is cooled by the natural convection of the coolant. The basic technologies for the natural convection capsule have already been established and the power transient tests will be started by using this capsule. The forced convection capsule has relatively complicated structure for circulating the cooling water and controlling the cooling water temperature. By performing several mock-up test, we confirmed that the target linear heat rate is achievable by the capsules, and elemental technologies to realize the forced convection capsule is feasible.
Nakamura, Takehiko; Nishiyama, Yutaka; Chimi, Yasuhiro; Sasajima, Hideo; Ogiyanagi, Jin; Nakamura, Jinichi; Suzuki, Masahide; Kawamura, Hiroshi
Proceedings of 16th Pacific Basin Nuclear Conference (PBNC-16) (CD-ROM), 6 Pages, 2008/10
In order to maintain and enhance safety of light water reactors in long-term and up-graded operations, proper understanding of irradiation behavior of fuels and materials is essentially important. Japanese government and the Japan Atomic Energy Agency have decided to refurbish the JMTR and to install new tests rigs, in order to play an active role for solving irradiation related issues on plant aging and high-duty uses of the current LWRs and on development of next-generation reactors. New tests on fuel integrity under simulated abnormal transients and high-duty irradiation conditions are planned in the JMTR. Power ramp tests of new design fuel rods will also be performed in the first stage of the program, which is expected to start in year 2011 after refurbishment of the JMTR. Combination of the JMTR tests with simulated reactivity initiated accident tests in the NSRR and loss of coolant accident tests in hot laboratories would serve as the integrated fuel safety research on the high performance fuels at extended burnups, covering from the normal to the accident conditions, including abnormal transients. For the materials irradiation, fracture toughness of reactor vessel steels and stress corrosion cracking behavior of stainless steels are being studied in addition to basic irradiation behavior of nuclear materials such as hafnium. The irradiation studies would contribute not only to solve the current problems but also to identify possible seeds of troubles and to make proactive responses.
Hanawa, Satoshi; Sato, Tomonori; Mori, Yuichiro; Ogiyanagi, Jin; Kaji, Yoshiyuki; Uchida, Shunsuke
Journal of Power and Energy Systems (Internet), 1(2), p.123 - 133, 2007/00
In order to evaluate the water chemistry in the irradiation field during IASCC irradiation test, a water radiolysis code for IASCC irradiation loop system was developed. In the water radiolysis code, a multiple node model was introduced since the irradiation loop system has a wide rage temperature distribution as well as the dose distribution. To investigate the applicability of developed water radiolysis code, water chemistry at the water sampling point of the irradiation loop system was measured and compared with analytical results under several water chemistry conditions. Further, water chemistry distribution in the in-pile region as well as in the out-pile region was calculated by the developed water radiolysis code.
Hanawa, Satoshi; Sato, Tomonori; Mori, Yuichiro; Ogiyanagi, Jin; Kaji, Yoshiyuki; Uchida, Shunsuke*
Proceedings of 14th International Conference on Nuclear Engineering (ICONE-14) (CD-ROM), 9 Pages, 2006/07
no abstracts in English
Hanawa, Satoshi; Ogiyanagi, Jin; Mori, Yuichiro*; Saito, Junichi; Tsukada, Takashi
JAEA-Conf 2006-003, p.350 - 357, 2006/05
no abstracts in English
Meguro, Yoshihiro; Ogiyanagi, Jin*; Tomioka, Osamu; Imura, Hisanori*; Ohashi, Kozaburo*; Yoshida, Zenko; Nakashima, Mikio
Proceedings of 2nd International Symposium on Supercritical Fluid Technology for Energy and Environment Applications (Super Green 2003), p.175 - 179, 2004/00
One of the most attractive properties of SFE is that changing solvent properties by tuning pressure can control distribution behavior of a metal ion. Distribution ratio (D) of uranium(VI) and plutonium(IV) with tributyl phosphate (TBP) from a nitric acid solution and palladium(II) with 2-methyl-8-qunolinol (HMQ) from a hydrochloric solution were determined in SFE at various pressures. In the extraction system using TBP, a linear relationship between the logarithmic distribution ratio (log D) and the solubility parameter of CO
was observed. The solubility parameter is difined based on the regular solution theory and is one of the parameters depending on the pressure. On the other hand, a linear relationship with a positive slope between log D and the solubility parameter was observed in the extraction system using HMQ. Most of the extractant was dissolved in the aqueous phase as HMQ
under the extraction condition examined.
-diketone and tributylphosphateMeguro, Yoshihiro; Iso, Shuichi; Ogiyanagi, Jin; Yoshida, Zenko
Analytical Sciences (CD-ROM), 17(Suppl.), p.721 - 724, 2002/03
no abstracts in English
Ogiyanagi, Jin; Meguro, Yoshihiro; Yoshida, Zenko; Ohashi, Kozaburo*
Analytical Sciences (CD-ROM), 17(Suppl.), p.717 - 720, 2002/03
no abstracts in English
Mori, Yuichiro*; Hanawa, Satoshi; Sato, Tomonori; Ogiyanagi, Jin; Nabeya, Hideaki; Uchida, Shunsuke*
no journal, ,
no abstracts in English
Hanawa, Satoshi; Ogiyanagi, Jin; Sato, Tomonori; Mori, Yuichiro*; Miwa, Yukio; Uchida, Shunsuke*
no journal, ,
no abstracts in English
Sato, Tomonori; Hanawa, Satoshi; Uchida, Shunsuke; Ogiyanagi, Jin; Nabeya, Hideaki; Miwa, Yukio; Kaji, Yoshiyuki; Nemoto, Yoshiyuki; Tsukada, Takashi
no journal, ,
To evaluate IASCC, the crack initiation and propagation tests under irradiation conditions have been carried out by using the in-pile loop installed at the Japan Material Testing Reactor (JMTR) of JAEA. In order to evaluate corrosive conditions, a water radiolysis model has been developed. In order to evaluate concentrations of chemical species not only in the test region but also along the sampling line, the model has to cover wide temperature range from room temperature to 288 C. The dissolved O, H and HO in sampled water was measured. The conclusions are summarized as follows; (1) The calculated O, H and HO concentrations comparatively agreed with measured values. (2) Decomposition of HO generated OH radicals, which contributed to recombination of HO and H. (3) The applicability and accuracy of the WRAC-JM to estimation of the water chemistry in the in-pile loop had been confirmed.
Ogiyanagi, Jin
no journal, ,
The failure mode of BWR fuels under a power transient condition has been found to change at the burnup range of 50 - 60 GWd/t. In order to confirm the fuel integrity with higher burnup, it is important to clarify the fuel failure behaviors for high burnup fuels (
60 GWd/t) at the power transient. JAEA is planing the power transient tests to obtain information such as a failure threshold and other behaviors of the high burnup fuels by using the test facility with high linear heat rate (LHR) irradiation, flexible LHR control, on-power capsule handling and on-line measurement of the fuel in JMTR.
Chimi, Yasuhiro; Ogiyanagi, Jin; Shimada, Sachio*; Nishiyama, Yutaka; Nakamura, Takehiko; Abe, Katsuhiro*
no journal, ,
In order to understand basic irradiation behavior of hafnium (Hf) which is used as the control rod for the Boiling Water Reactor (BWR), we are planning irradiation tests of Hf at Japan Materials Testing Reactor (JMTR). Prior to the irradiation tests, some of the fundamental properties of unirradiated Hf were obtained. In the present report, we discuss the results of optical microscope and SEM observations, and crystal orientation measurement.
Shimada, Sachio*; Ogiyanagi, Jin; Chimi, Yasuhiro; Nishiyama, Yutaka; Nakamura, Takehiko; Abe, Katsuhiro*
no journal, ,
no abstracts in English
Nakamura, Takehiko; Nishiyama, Yutaka; Sasajima, Hideo; Chimi, Yasuhiro; Ogiyanagi, Jin; Nakamura, Jinichi
no journal, ,
In order to contribute to development and safety research on light water reactors, Japan Atomic Energy Agency is planning to re-start the Japan Materials Testing Reactor (JMTR) in 2011, after its refurbishment. Irradiation test facilities of the fuels and materials are being installed in the JMTR, in parallel to the refurbishment. Outlines of the project is presented in a series of three presentations.
