Sugita, Yutaka; Taniguchi, Naoki; Makino, Hitoshi; Kanamaru, Shinichiro*; Okumura, Taisei*
Nippon Genshiryoku Gakkai Wabun Rombunshi, 19(3), p.121 - 135, 2020/09
A series of structural analysis of disposal containers for direct disposal of spent fuel was carried out to provide preliminary estimates of the required pressure resistance thickness of the disposal container. Disposal containers were designed to contain either 2, 3 or 4 spent fuel assemblies in linear, triangular or square arrangements, respectively. The required pressure resistance thickness was evaluated using separation distance of the housing space for each spent fuel assembly as a key model parameter to obtain the required thickness of the body and then the lid of the disposal container. This work also provides additional analytical technical knowledge, such as the validity of the setting of the stress evaluation line and the effect of the model length on the analysis. These can then be referred to and used again in the future as a basis for conducting similar evaluations under different conditions or proceeding with more detailed evaluations.
Okita, Shoichiro; Tasaki, Seiji*; Abe, Yutaka*
Nippon Genshiryoku Gakkai Wabun Rombunshi, 19(3), p.178 - 184, 2020/09
The Kyoto University Accelerator-based Neutron Source (KUANS) is a compact neutron source that is mainly used for spectrometer and detector development. In addition, it is also suited for experiments to study the neutronic design of moderators owing to the relatively low neutron generation yield by Be(p,n). We present a neutronic design of the neutron moderator on a reentrant-hole configuration for KUANS to enhance the neutron emission, and some experiments are conducted at KUANS for verification. A polyethylene moderator on a reentrant-hole configuration is designed by PHITS calculation and is introduced to KUANS to obtain intense oblong neutron beams. The intensity of the pulsed neutron beam is experimentally measured. The results reveal that the intensity becomes approximately 1.9 times stronger than that of the conventional rectangular design. In addition, the ratio of its intensity to the conventional intensity increases to approximately threefold as the neutron wavelength increases. It is interesting to note that the longer the neutron wavelength, the more efficiently they are extracted from the inside of the moderator owing to the existence of the reentrant-hole configuration.
Ono, Masato; Fujiwara, Yusuke; Matsumoto, Tetsuro*; Iigaki, Kazuhiko
Nippon Genshiryoku Gakkai Wabun Rombunshi, 19(2), p.110 - 120, 2020/06
Integrity confirmation for buildings against collisions of projectiles has been conducted to evaluate collisions between a projectile with simple shape and a wall using empirical formulas. It is a matter of fact, there is a possibility that structures with complex shape such as stack may collide with a reactor building. However, there were not so many studies of collisions between structures with complex shape and buildings in the literature. Impact evaluation was carried out using reactor building and stack with real shape and adequate physical property. It was found that ceiling of reactor building was not damaged by the collision, confirming that there was no effect inside of reactor building.
Kakiuchi, Kazuo; Amaya, Masaki
Nippon Genshiryoku Gakkai Wabun Rombunshi, 19(1), p.24 - 33, 2020/03
The irradiation growth behavior of the improved Zr alloys for light-water reactor fuel cladding was investigated. The coupon specimens, which were prepared from fuel cladding tubes with improved Zr alloys, had been irradiated in the Halden reactor in Norway at temperatures of 300 and 320C under a typical water chemistry condition of PWR and 240C under the coolant condition of the Halden reactor up to a fast neutron fluence of 810 (n/cm, E 1 MeV). During and after the irradiation test, the amount of irradiation growth of each specimen was evaluated. The effect of the difference in alloy composition on irradiation growth behavior seemed insignificant if the other conditions e.g. the final heat treatment condition at fabrication, irradiation temperature and the amount of hydrogen pre-charged in the specimen were the same.
Akiba, Miyuki*; Hotta, Akitoshi*; Abe, Yutaka*; Sun, Haomin
Nippon Genshiryoku Gakkai Wabun Rombunshi, 19(1), p.1 - 15, 2020/02
Tests at three different scales were conducted in order to understand the mechanisms of pool scrubbing. In the small-scale separate effect test, high resolution two-phase flow measurement techniques such as a high-speed camera, wire mesh sensor and PIV were applied to capture the behaviors of a single bubble and two-phase flow structures. In the large-scale integral effect test, the dependence of the aerosol removal efficiency on submergence and pool temperature was measured under constant pressure and depressurized conditions. To clarify relationships between individual phenomena and combined phenomena observed in two tests, the mid-scale integral effect test was undertaken.
Machida, Masahiko; Yamada, Susumu; Iwata, Ayako; Otosaka, Shigeyoshi; Kobayashi, Takuya; Watanabe, Masahisa; Funasaka, Hideyuki; Morita, Takami*
Nippon Genshiryoku Gakkai Wabun Rombunshi, 18(4), p.226 - 236, 2019/12
After direct discharges of highly-contaminated water from Unit 2 and 3 in Fukushima Daiichi Nuclear Power Plant (1F) in April to May 2011, Kanda suggested that relatively small run-off of radionuclides from 1F port into Fukushima coastal region has subsequently continued by using his estimation scheme. However, the estimation period was limited until September 2012, and there has been no report on the issue since the work. Therefore, this paper focuses on discharge inventory from 1F port until June 2018. In the missing period, the central government and Tokyo Electric Power Company Holdings have done continuous efforts to stop the discharge, and consequently sea water concentration inside 1F port has diminished gradually. We show monthly discharge inventory of Cs-137 until June 2018 by two schemes, i.e., Kanda's scheme partially improved by authors and more sophisticated one using Voronoi tessellation reflecting the increment of the number of monitoring points inside 1F port. The results show that the former always presents overestimated results compared to the latter but the ratio of former to latter is less than one order. Based on these results, we evaluate impact of discharge inventory from 1F port into the coastal area and radiation does via fish digestion.
Aihara, Jun; Yasuda, Atsushi*; Ueta, Shohei; Ogawa, Hiroaki; Honda, Masaki*; Ohira, Koichi*; Tachibana, Yukio
Nippon Genshiryoku Gakkai Wabun Rombunshi, 18(4), p.237 - 245, 2019/12
Development of fabrication and inspection technologies of oxidation resistant fuel element for improvement of safety of high temperature gas-cooled reactors (HTGRs) in severe oxidation accident was carried out. Simulated coated fuel particles (CFPs), alumina particles, were over-coated with mixed powder of Si, C and small amount of resin to form over-coated particles, and over-coated particles were molded and hot-pressed to sinter simulated oxidation resistant fuel elements with SiC/C mixed matrix. Simulated oxidation resistant fuel elements with matrix whose Si/C mole ratio is 1.00 were fabricated. Failure fraction of CFPs in fuel elements is one of very important inspection subjects of HTGR fuel. It is essential that CFPs are extracted from fuel elements without additional failure. Development of method for extraction of CFPs was carried out. Desolation of SiC by KOH method or pressurized acidolysis method should be applied to extraction of CFPs.
Sudo, Ayako; Mizusako, Fumiki*; Hoshino, Kuniyoshi*; Sato, Takumi; Nagae, Yuji; Kurata, Masaki
Nippon Genshiryoku Gakkai Wabun Rombunshi, 18(3), p.111 - 118, 2019/08
Cooling rate of molten core materials during solidification significantly affects the segregation of major constituents of fuel debris. To understand general tendency of the segregation, liquefaction/solidification tests of simulated corium (UO, ZrO, FeO, BC and sim-FP oxides) were performed. Simulated corium was heated up to 2600C under Ar atmosphere and then cooled down with two different cooling processes; furnace cooling (average cooling rate is approximately 744C/min) and slow cooling (cooling rate in 2600C2300C is 5C/min and in 2300C1120C is approximately 788C/min). Element analysis detected three oxide phases with different composition and one metal phase in both solidified samples. Solubility of FeO in these oxide phases was mostly fixed to be 125at% in both samples, which is in reasonable accordance with the value estimated from UO-ZrO-FeO phase diagrams. However, a significant grain-growth of one oxide phase, rich in Zr-oxide, was detected only in the slowly cooled sample. The composition of this particular oxide phase is comparable to the initial average composition. The condensation is considered to be caused by the connection of remaining liquid agglomerates during slow solidification.
Yoshida, Kazuo; Tamaki, Hitoshi; Yoshida, Naoki; Yoshida, Ryoichiro; Amano, Yuki; Abe, Hitoshi
Nippon Genshiryoku Gakkai Wabun Rombunshi, 18(2), p.69 - 80, 2019/06
An accident of evaporation to dryness by boiling of high level liquid waste (HLLW) is postulated as one of the severe accidents at a fuel reprocessing plant. In this case, volatile radioactive materials, such as ruthenium (Ru) are released from the tanks with water and nitric-acid mixed vapor into atmosphere. In addition to this, nitrogen oxides are also released formed by the thermal decomposition of metal nitrates of fission products (FP) in HLLW. It has been observed experimentally that nitrogen oxide affects strongly to the transport behavior of Ru. Chemical reactions of nitrogen oxide with water and nitric acid are also recognized as the complex phenomena to undergo simultaneously in the vapor and liquid phases. An analysis method has been developed with coupling two types of computer codes to simulate not only thermo-hydraulic behavior but also chemical reactions in the flow paths of carrier gases. A simulation study has been also carried out with a typical facility building.
Yoshizawa, Atsufumi*; Oba, Kyoko; Kitamura, Masaharu*
Nippon Genshiryoku Gakkai Wabun Rombunshi, 18(2), p.55 - 68, 2019/06
This study aims to improve the potential of an emergency response by analyzing the workload management during the accident at the Emergency Response Center (ERC) of TEPCO's Fukushima Daiichi Nuclear Power Plant. Specifically, the research focused on the response of the ERC during the time between the discontinuation of Unit 3 core water injection and its recovery. It identified the different types of workload at the ERC had and how they had been managed based on the record of a TV conference. It also deduced the casual factors of the responses, supplementing the interview record of the director of ERC at the time by applying workload management analysis. On the basis of these findings, lessons to enhance the potential of the on-site emergency response have been obtained for ERC and outside organizations.
Aihara, Jun; Honda, Masaki*; Ueta, Shohei; Ogawa, Hiroaki; Ohira, Koichi*; Tachibana, Yukio
Nippon Genshiryoku Gakkai Wabun Rombunshi, 18(1), p.29 - 36, 2019/03
Japan Atomic Energy Agency carried out development of fabrication technology of oxidation resistant fuel element for improvement of safety of high temperature gas-cooled reactors in serious oxidation accident, based on precursor research in former JAEA. Dummy coated fuel particles (alumina particles) were over-coated with mixed powder of Si, C and small amount of resin to form over-coated particles, and over-coated particles were molded and hot-pressed to sinter dummy oxidation resistant fuel elements with SiC/C mixed matrix. We fabricated dummy oxidation resistant fuel elements with matrix whose Si/C mole ratio (about 0.551) is three times as large as that in precursor research. Si peak was not detected by X-ray diffraction of matrix. Better oxidation resistant was confirmed with oxidation test in 20% O at 1673 K than that of ordinal fuel compact with ordinal graphite/carbon matrix. All dummy coated fuel particles were held in specimen after 10 h oxidation.
Sakatani, Keiichi*; Nakatani, Takayoshi; Funabashi, Hideyuki
Nippon Genshiryoku Gakkai Wabun Rombunshi, 18(1), P. 43, 2019/03
The article authored by Keiichi Sakatani et al. and titled "Estimation of corrosion rate of Zr-2.5wt%Nb alloy (Fugen pressure tube) under low-temperature, low-oxygen, and high-pH conditions", Journal of the Atomic Energy Society of Japan, Vol.14, No.4, p.261-267 (2015), has been found to have a problem about reliability of the corrosion data acquisition, and thus retracted by the authors.
Takamatsu, Yuki*; Ishii, Hiroto*; Oishi, Yuji*; Muta, Hiroaki*; Yamanaka, Shinsuke*; Suzuki, Eriko; Nakajima, Kunihisa; Miwa, Shuhei; Osaka, Masahiko; Kurosaki, Ken*
Nippon Genshiryoku Gakkai Wabun Rombunshi, 17(3/4), p.106 - 110, 2018/12
In order to establish the synthesis method of simulated fuel contacting Cesium (Cs) which is required for the evaluation of physical/chemical characteristics in fuel and release behavior of Cs, sintering tests of the cerium dioxide (CeO) based simulated fuels containing Cesium iodide (CsI) are performed by using spark plasma sintering (SPS) method. The sintered CeO pellets with homogeneous distribution of several micro meter of CsI spherical precipitates were successfully obtained by optimizing SPS conditions.
Yamashita, Susumu; Tada, Kenichi; Yoshida, Hiroyuki; Suyama, Kenya
Nippon Genshiryoku Gakkai Wabun Rombunshi, 17(3/4), p.99 - 105, 2018/12
In order to reveal melt relocation behaviors of core internals phenomenologically and to reduce the uncertainties of the melt relocation analysis in existing SA analysis codes, in JAEA, the numerical simulation code for melt relocation and accumulation behaviors based on computational fluid dynamics named JUPITER has been developed. In this paper, to consider the estimation method for fuel debris composition and its re-criticality, we performed the melt accumulating and spreading simulation to the pedestal region by JUPITER and also performed re-criticality analysis by Monte Carlo Codes for Neutron Transport Calculations based on Continuous Energy and Multi-group Methods (MVP) using detailed fuel debris composition data obtained by JUPITER. From the coupled analysis on fuel debris distribution by JUPITER and MVP, we had prospects for a detailed possibility of re-criticality of fuel debris with detailed fuel debris distribution.
Sakamoto, Hiroyuki*; Akagi, Yosuke*; Yamada, Kazuo*; Tachi, Yukio; Fukuda, Daisuke*; Ishimatsu, Koichi*; Matsuda, Mikiya*; Saito, Nozomi*; Uemura, Jitsuya*; Namihira, Takao*; et al.
Nippon Genshiryoku Gakkai Wabun Rombunshi, 17(2), p.57 - 66, 2018/05
Concrete debris contaminated with radioactive cesium and other nuclides have been generated from the accident in the Fukushima Dai-ichi Nuclear Power Plant and there will be generated due to the decommissioning of nuclear power plants in the future. Although conventional decontamination techniques are effective for flat concrete surfaces such as floors and walls, it is not clear what techniques to apply for decontaminating radioactive concrete debris. In this study, focusing on a pulsed power discharge technique, fundamental experimental works were carried out. Decontamination of concrete by applying the aggregate recycling technique using the pulsed power discharge technique was evaluated by measuring radioactivity of aggregate and sludge separated from the contaminated concrete. The results suggest that the separation into aggregate and sludge of the contaminated concrete debris could achieve decontamination and volume reduction of the radioactive concrete debris.
Tsuda, Shuichi; Tanigaki, Minoru*; Yoshida, Tadayoshi; Okumura, Ryo*; Saito, Kimiaki
Nippon Genshiryoku Gakkai Wabun Rombunshi, 17(1), p.11 - 17, 2018/03
Environmental dose rate monitoring has been performed using scintillation detectors since the Fukushima Dai-ichi Nuclear Power Plant accident happened. After the accident, various scintillation detectors with directional dependence are used in the measurement, though detectors with superior directional dependence are needed because photons come from various direction in the environment. To investigate the influence of crystal configurations on indicated values of dose rates, pulse height spectra were measured using scintillation-based detectors with different crystal configurations and ambient dose rates were obtained using a spectrum - dose conversion operator (G(E) function). It is found that the dose rate for a rectangular-parallelepiped crystal is 40% than that for a cylindrical one at the maximum. However, the values agreed within 10% among all the detectors irrespective of the crystal shapes, using G(E) functions determined in a rotational irradiation geometry.
Hamamoto, Shimpei; Tochio, Daisuke; Ishii, Toshiaki; Sawahata, Hiroaki
Nippon Genshiryoku Gakkai Wabun Rombunshi, 16(4), p.169 - 172, 2017/12
A melt wire was installed at the tip of the control rod in order to measure the temperature of High Temperature engineering Test Reactor (HTTR). After experience with reactor scrum from the state of reactor power 100%, the melt wire was taken out from the control rod and appearance has been observed visually. It was confirmed that the melt wires with a melting point of 505 C or less were melted, and the melt wires with a melting point of 651 C or more were not melted. Therefore, it was found that the highest arrival temperature of tip of the control rods where the melt wires are installed reaches within the range of 505 to 651 C. And it was found that the control rod temperature at the time of reactor scram does not exceed the using temperature criteria (900 C) of Alloy 800H of the control rod sleeve.
Okamoto, Yoshihiro; Nagai, Takayuki; Shiwaku, Hideaki; Inose, Takehiko*; Sato, Seiichi*
Nippon Genshiryoku Gakkai Wabun Rombunshi, 16(4), p.180 - 190, 2017/12
X-ray absorption fine structure (XAFS) and imaging XAFS analyses were performed to elucidate chemical state of rhodium in the simulated waste glass. The chemical forms of Rh in the glass were evaluated to be 84% RhO and 16% metal/alloy as the result of linear combination analysis of EXAFS data. According to the imaging XAFS analysis, the chemical form of Rh which was located together with Ru was mainly oxide (RhO). It suggests that stable (Ru,Rh)O solid solution exists in the simulated glass. On the other hand, that of Rh of which distribution did not accord with Ru in the glass was mainly metallic. In the case of metallic Rh in the glass, it tended to become an aggregation form. It can be concluded that the chemical state of Rh was much affected by the existence and distribution of Ru element.
Hotta, Akitoshi*; Morita, Akinobu*; Kajimoto, Mitsuhiro*; Maruyama, Yu
Nippon Genshiryoku Gakkai Wabun Rombunshi, 16(3), p.139 - 152, 2017/09
Ando, Masaki; Matsuda, Norihiro; Saito, Kimiaki
Nippon Genshiryoku Gakkai Wabun Rombunshi, 16(2), p.63 - 80, 2017/05
In order to discriminate the contribution of radioactive cesium due to the Fukushima Dai-ichi Nuclear Power Plant accident to the air dose rates measured by the car-borne surveys, natural background radiation was evaluated for eastern Japan area as the municipality averaged values. The window count method for distinction between natural and artificial radioactive nuclides was applied to the car-borne surveys using the KURAMA-II. Distribution of the evaluated natural background radiation showed geological feature, and it was found that the radiations measured along paved roads were reflecting the distribution of terrestrial -rays. The effect of the radioactive cesium as of 2014 for the municipalities designated as the Intensive Contamination Survey Area was beyond the uncertainty of the natural background radiation. That for the other municipalities, however, was found to be almost negligible.