Miyahara, Naoya; Miwa, Shuhei; Goullo, M.*; Imoto, Jumpei; Horiguchi, Naoki; Sato, Isamu*; Osaka, Masahiko
Journal of Nuclear Science and Technology, 57(12), p.1287 - 1296, 2020/12
In order to clarify the cesium iodide (CsI) transport behavior with a focus on the mechanisms of gaseous iodine formation in the reactor coolant system of LWR under a severe accident condition, a reproductive experiment of CsI transport behavior was conducted using a facility equipped with a thermal gradient tube. Various analyses on deposits and airborne materials during transportation could elucidate two mechanisms for the gaseous iodine formation. One was the gaseous phase chemical reaction in Cs-I-O-H system at relatively high-temperature region, which led to gaseous iodine transport to the lower temperature region without any further changes in gas species due to the kinetics limitation effects. The other one was the chemical reactions related to condensed phase of CsI, namely those of CsI deposits on walls with surface of stainless steel to form CsCrO compound and CsI aerosol particles with steam, which were newly found in this study.
Tanaka, Kosuke; Sato, Isamu*; Onishi, Takashi; Ishikawa, Takashi; Hirosawa, Takashi; Katsuyama, Kozo; Seino, Hiroshi; Ohno, Shuji; Hamada, Hirotsugu; Tokoro, Daishiro*; et al.
Journal of Nuclear Materials, 536, p.152119_1 - 152119_8, 2020/08
In order to obtain the release rate coefficients from fuels for fast reactors (FRs), heating tests and the subsequent analyses of the fission products (FPs) and actinides that are released were carried out using samples of uranium-plutonium mixed oxide (MOX) fuel pellets irradiated at the experimental fast reactor Joyo. Three heating tests targeting temperatures of 2773, 2973 and 3173 K were conducted using an FP release behavior test apparatus equipped with a high-frequency induction furnace and solid FP sampling systems consisting of a thermal gradient tube (TGT) and filters. Irradiated fuel pellets were placed into a tungsten crucible, then loaded into the induction furnace. The temperature was raised continuously at a heating rate of 10 K/s to the targeted temperature and maintained for 500 s in a flowing argon gas atmosphere. The FPs and actinides released from the MOX fuels and deposited in the TGT and filters were quantified by gamma-ray spectrometry and inductively coupled plasma mass spectrometry (ICP-MS) analysis. Based on the analysis, the release rates of radionuclides from MOX fuels for FR were obtained and compared with literature data for light water reactor (LWR) fuels. The release rate coefficients of FPs obtained in this study were found to be similar to or lower than the literature values for LWR fuels. It was also found that the release rate coefficient data for actinides were within the range of variation of literature values for LWR fuels.
Miyahara, Naoya; Miwa, Shuhei; Horiguchi, Naoki; Sato, Isamu*; Osaka, Masahiko
Journal of Nuclear Science and Technology, 56(2), p.228 - 240, 2019/02
In order to improve LWR source term under severe accident conditions, the first version of a fission product (FP) chemistry database named "ECUME" was developed. The ECUME is intended to include major chemical reactions and their effective kinetic constants for representative SA sequences. It is expected that the ECUME can serve as a fundamental basis from which FP chemical models in the SA analysis codes can be elaborated. The implemented chemical reactions in the first version were those for representative gas species in Cs-I-B-Mo-O-H system. The chemical reaction kinetic constants were evaluated from either literature data or calculated values using ab-initio calculations. The sample chemical reaction calculation using the presently constructed dataset showed meaningful kinetics effects at 1000 K. Comparison of the chemical equilibrium compositions by using the dataset with those by chemical equilibrium calculations has shown rather good consistency for the representative Cs-I-B-Mo-O-H species. From these results, it was concluded that the present dataset should be useful to evaluate FP chemistry in Cs-I-B-Mo-O-H system under LWA SA conditions.
Anraku, Sohtaro; Matsubara, Isamu*; Morimoto, Kazuya*; Sato, Tsutomu*
Nendo Kagaku, 55(2), p.17 - 30, 2017/00
Anionic radionuclides are important for the long-term safety assessment of Japanese transuranic (TRU) waste disposal facilities. Degradation of cementitious materials used to construct the TRU waste disposal facilities, however, can produce a hyperalkaline leachate and so it is necessary to understand the reaction mechanisms that will control the behavior and fate of anionic radionuclides under these hyperalkaline conditions. An excellent natural analogue site to study relevant reaction mechanisms is provided in Oman where hyperalkaline spring waters (pH 11) from serpentinized peridotites discharge into moderately alkaline rivers. Aragonite was found in all secondary mineral samples, with accessory minerals of calcite, layered double hydroxide (LDH) and brucite. LDH was observed at the high Al concentration springs and brucite at the low Al concentration springs. Calcite was only found close to the springs. Distal calcite formation was inhibited due to high Mg concentrations in the river water. The spatial distribution of minerals therefore implicates the importance of the mixing ratio of spring to river water and the relative chemical compositions of the spring and river waters. Supporting mixing model calculations could successfully reproduce the precipitation of aragonite and LDH. The observed decrease in Ca concentration could be explained by aragonite precipitation. pH exerted a strong control on the precipitation of LDH and so too, therefore, on Al concentration. In the mixing water experiments containing up to 40% river water, LDH and brucite were both oversaturated, but brucite was not always identified by XRD. The possible inhibition of brucite by LDH precipitation was an unexpected result.
Shizukawa, Yuta; Sekio, Yoshihiro; Sato, Isamu*; Maeda, Koji
Proceedings of 54th Annual Meeting of Hot Laboratories and Remote Handling (HOTLAB 2017) (Internet), 5 Pages, 2017/00
Electrochemical corrosion behavior under salt water in a type 304L stainless steel used to a part of BWR core materials was investigated to evaluate the possibility of crevice corrosion occurrence for the fuel assemblies which experienced seawater exposure in Fukushima Daiichi Nuclear Power Plant (1F) accident. Especially, focusing on the upper end plug part having the 304L SS crevice structure, measurement of repassivation potential for crevice corrosion () were carried out using the crevice test pieces fabricated by 304L SS plates. From the results, was lower than the spontaneous potential () when the conditions of 2500 ppm chloride ion concentration at over 50 C or that of 2500 ppm at over 80 C, which are included in the SFP water quality conditions. Therefore, in the 304L SS parts of the 1F fuel assemblies that experienced seawater exposure, there is a possibility of crevice corrosion occurrence.
Shimomura, Yusuke; Hanari, Akira*; Sato, Isamu*; Kitamura, Ryoichi
JAEA-Technology 2015-062, 47 Pages, 2016/03
In response to new standards for regulating waste management facilities, it was carried out impact assessment of forest fires on the waste management facilities existed in Oarai Research and Development Center of Japan Atomic Energy Agency. At first, a fire spread scenario of forest fires was assumed. The intensity of forest fires was evaluated from field surveys, forest fire evaluation models and so on. As models of forest fire intensity evaluation, Rothermel Model and Canadian Forest Fire Behavior Prediction (FBP) System were used. Impact assessment of radiant heat to the facilities was carried out, and temperature change of outer walls for the assumed forest fires was estimated. The outer wall temperature of facilities was estimated around 160C at the maximum, it was revealed that it doesn't reach allowable temperature limit. Consequently, it doesn't influence the strength of concrete. In addition, a probability of fire breach was estimated to be about 20%. This report illustrates an example of evaluation of forest fires for the new regulatory standards through impact assessment of the forest fires on the waste management facilities.
Sakauchi, Hitoshi; Sato, Isamu*; Donomae, Yasushi; Kitamura, Ryoichi
JAEA-Technology 2015-059, 352 Pages, 2016/03
OWTF (Oarai Waste Reduction Treatment Facility) is constructed for volume reduction processing and stabilization treatment of solid waste, which was generated from hot facilities in Oarai Research and Develop Center of Japan Atomic Energy Agency, using in-can type high frequency induction heating by remote control. This report describes corroborative tests, in which incinerating and melting performance for OWTF is confirmed with a full-scale testing furnace. We have been carrying out the tests of incinerating and melting treatment with some kinds of simulated wastes, such as enclosure form of radioactive wastes, material and articles.
Takai, Toshihide; Sato, Isamu*; Yamashita, Shinichiro; Furukawa, Tomohiro
JAEA-Technology 2015-043, 56 Pages, 2016/02
Fundamental research on FP-chemistry for fission product release behaviors under severe accident was carried out for reinforcement of source term evaluation, and implementation of the 1F decommissioning R&D project. There were subjects to clarified (1) FP chemistry behavior between vapor species release and aerosol formation and (2) physical parameters which would be affect subsequent aerosol's chemical behavior, for improvement of FP transport model. Applicability of measuring/analyzing techniques presently used was studied for evaluating foregoing properties. And the validity was verified by trial measurements. In conclusion, Raman spectrometry and high temperature X-ray diffraction were hopeful to determine FP-chemical form against vapor/aerosol species and aerosol species, respectively. Combination use of cascade impactor and scanning type electron microscope with energy-dispersive X-ray spectrometry was hopeful to determine physical parameters of aerosol.
Okane, Tetsuo; Kobata, Masaaki; Sato, Isamu*; Kobayashi, Keisuke*; Osaka, Masahiko; Yamagami, Hiroshi
Nuclear Engineering and Design, 297, p.251 - 256, 2016/02
Tanaka, Kosuke; Sato, Isamu; Hirosawa, Takashi; Kurosaki, Ken*; Muta, Hiroaki*; Yamanaka, Shinsuke*
Journal of Nuclear Science and Technology, 52(10), p.1285 - 1289, 2015/10
Polycrystalline specimens of americium-containing barium plutonate have been prepared by mixing the appropriate amounts of (PuAm)O and BaCO powders followed by reacting and sintering at 1600 K under the flowing gas atmosphere of dry-air. The sintered specimens had a single phase of orthorhombic perovskite structure and were crack-free. The elastic moduli were determined from the longitudinal and shear sound velocities. The Debye temperature was also determined from the sound velocities and lattice parameter measurements. The thermal conductivity was calculated from the measured density at room temperature, literature values of heat capacity, and thermal diffusivity measured by laser flash method in vacuum. The thermal conductivity of americium-containing barium plutonate was roughly independent of the temperature and was almost the same magnitude as that of BaPuO and BaUO.
Sato, Isamu; Tanaka, Kosuke; Koyama, Shinichi; Matsushima, Kenichi*; Matsunaga, Junji*; Hirai, Mutsumi*; Endo, Hiroshi*; Haga, Kazuo*
Energy Procedia, 82, p.86 - 91, 2015/07
Experiments simulating overheating conditions of fast reactor severe accidents have been previously carried out with irradiated fuels. For the present study, the chemical forms of the fission products (FPs) included in the irradiated fuels were evaluated by thermochemical equilibrium calculations. At temperatures of 2773 K and 2973 K, the most stable forms of Cs, I, Te, Sb, Pd and Ag are gaseous compounds. Cs and Sb detected in the thermal gradient tube (TGT) in the experiments can take gaseous chemical forms of elemental Cs, CsI, CsMoO, CsO and elemental Sb, SbO, SbTe, respectively. By comparing experimental results and the estimations, it is seen CsI thermochemically behaves in a manner that traps it in the TGT, while elemental Cs trends to move as fine particles. The moving behavior of the gaseous FPs will obey not only thermochemical principles, but also those of particle dynamics.
Sato, Isamu; Onishi, Takashi; Tanaka, Kosuke; Iwasaki, Maho; Koyama, Shinichi
Journal of Nuclear Materials, 461, p.22 - 28, 2015/06
We observed one of the simplified processes by conducting primitive experiments. CsI was heated at 1323 K to be vaporized and deposited on sampling parts with a temperature range of 1023 - 423 K and then BO was vaporized at 1973 K to be reacted with Cs/I there. After heating tests, each sampling part was soaked into alkali water to dissolve the surface-deposits for ICP-MS analysis. The results showed that CsI deposited at the sampling parts kept above approx. 850 K was striped by BO vapour. This behaviour will be thermodynamically discussed to study the Cs/I/B chemistry in the severe accidents.
Sato, Isamu; Onishi, Takashi; Tanaka, Kosuke; Iwasaki, Maho; Koyama, Shinichi
Journal of Nuclear Materials, 461, p.22 - 28, 2015/06
In order to evaluate B influence on the release and transport of Cs and I during severe accidents, basic experiments have been performed on the interaction between deposited Cs/I compounds and vapor/aerosol B compounds. CsI and BO were utilized as a Cs/I compound and a B compound, respectively. Deposited CsI on the thermal gradient tube (TGT), which is exposed to temperatures ranging from 423 K to 1023 K was reacted with vapor/aerosol BO, and then observed to determine how it changed Cs/I decomposition profiles. As a result, vapor/aerosol BO stripped a portion of deposited CsI within a temperature range from 830 K to 920 K to make gaseous CsBO and I. In addition, gaseous I was re-deposited at a temperature range from 530 K to 740 K, while CsBO travelled through the sampling tubes and filters without deposition. It is implied that B influences Cs carriers such as CsBO to transport Cs to the colder regions.
Sato, Isamu; Maeda, Koji; Suto, Mitsuo; Osaka, Masahiko; Usuki, Toshiyuki; Koyama, Shinichi
Journal of Nuclear Science and Technology, 52(4), p.580 - 587, 2015/04
Penetration behavior of radionuclides such as Cs into dried concrete material, dried mortar material and epoxy paint for a few dozen days was observed using a solution containing fission products extracted from irradiated fuels to obtain fundamental information on the radionuclide penetration rate and depth. Hardly any radionuclides could penetrate into the epoxy paint. The radionuclide solution penetrated into concrete and mortar materials to a depth of a few millimeters for a few dozen days. The penetration behavior observed near the surface of concrete and mortar materials was similar to the diffusion of nuclides in media such as water-saturated concrete, bentonite and cement materials.
Sato, Isamu; Miwa, Shuhei; Tanaka, Kosuke; Nakajima, Kunihisa; Hirosawa, Takashi; Iwasaki, Maho; Onishi, Takashi; Osaka, Masahiko; Takai, Toshihide; Amaya, Masaki; et al.
Proceedings of 2014 Water Reactor Fuel Performance Meeting/ Top Fuel / LWR Fuel Performance Meeting (WRFPM 2014) (USB Flash Drive), 6 Pages, 2014/09
A new research program on severe accidents is lunched for the evaluation of FP release and transport behavior in BWR system. The purpose of the program is to improve the FP release and transport model using experimental database about FP chemistry focusing on Cs and I chemistry. In this program, effects of B including in control rod materials, BC for the Cs and I chemistry are paid attention. The experimental database used for the improvement will consist of results to obtain with newly-prepared test device under atmosphere with broad-ranging oxygen and/or steam partial pressure simulated those in BWR. The state of preparation for these experimental studies and analyses is introduced. In addition, the preliminary test was moved into action to show B chemical effect on Cs and I transport under one of the processes, which is deposited Cs compounds and B vapor and aerosol interaction. In this experiment, a "B stripping effect" to deposited CsI was observed.
Maeda, Koji; Sasaki, Shinji; Kumai, Misaki; Sato, Isamu; Suto, Mitsuo; Osaka, Masahiko; Goto, Tetsuo*; Sakai, Hitoshi*; Chigira, Takayuki*; Murata, Hirotoshi*
Journal of Nuclear Science and Technology, 51(7-8), p.1006 - 1023, 2014/07
Since the start of the severe accident at the Fukushima Daiichi Nuclear Power Plant in March 2011, concrete surfaces within the reactor buildings have been exposed to radioactive contaminants. Released radiation sources still remain too high to permit entry into some areas of the RBs to allow the damage to be assessed and to allow carrying out the restoration of lost safety functions, decommissioning activities, etc. In order to clarify the situation of this contamination in the RBs, 18 samples were subjected to analyses to determine the surface radionuclide concentrations and to characterize the radionuclide distributions in the samples. Decontamination tests on the sample of Unit 2 were conducted to reduce the levels of radioactivity present near the sample surface. As a result of the tests, the level of radioactivity of the sample was reduced with the removal of 97% of the contamination present near the sample surface.
Tanaka, Kosuke; Miwa, Shuhei; Sato, Isamu; Hirosawa, Takashi; Sekine, Shinichi; Osaka, Masahiko; Obayashi, Hiroshi; Koyama, Shinichi
Journal of Nuclear Science and Technology, 51(7-8), p.876 - 885, 2014/07
As a first step for obtaining experimental data on the effects of high-temperature chemical interaction on fission product (FP) release behavior, we focused on the dissolution of irradiated uranium plutonium mixed oxide (MOX) fuel by molten zircaloy (Zry), and carried out a heating test under the reducing atmosphere. Pieces of an irradiated MOX fuel pellet and cladding were subjected to the heating test at 2373 K for 5 min. The fractional release rate of cesium (specifically Cs) was monitored during the test and its release behavior was evaluated. The observation of microstructures and measurements of elemental distribution in the heated specimen were also performed. We demonstrated experimentally that the fuel dissolution by molten Zry accelerated the release of Cs from the fuel pellets.
Usuki, Toshiyuki; Sato, Isamu; Suto, Mitsuo; Maeda, Koji; Osaka, Masahiko; Koyama, Shinichi; Tokoro, Daishiro*; Sekioka, Ken*; Ishigamori, Toshio*
JAEA-Testing 2014-001, 29 Pages, 2014/05
The penetration tests with solution containing radioactive nuclides were experimented to understand basic data for floor and wall materials of Fukushima Daiichi reactor buildings. The solution prepared from irradiated fuels was used as solution containing radioactive nuclides. The solution was applied to surface of epoxy paint, dried concrete and mortar used as specimens. Dose-rate profiles of direction of depth were given by radiation measurement and grinding of the specimens. The penetrations of radioactive nuclides for epoxy paint specimens were not clearly observed and the penetration depths would be within 0.4 mm. The penetrations of radioactive nuclides for dried concrete specimens proceeded. The penetration rates were substantially decreased when 16 days have elapsed from start. The dose rates of penetrated dried concrete specimens were reduced to background by grinding-2.0 mm. -ray spectrometry measurement showed that penetration behavior of near surface concrete are different among nuclides and the penetration behavior of radioactive nuclides into dried concrete and mortar materials through solution is similar to migration behavior of ions into those water-saturated materials.
Maeda, Koji; Sasaki, Shinji; Kumai, Misaki; Sato, Isamu; Suto, Mitsuo; Osaka, Masahiko
JAEA-Research 2013-025, 123 Pages, 2014/01
In order to clarify the situation of the contamination in the Fukushima Daiichi reactor buildings of Units 1, 2 and 3, selected samples were transported to the Oarai Engineering Center of JAEA where they were subjected to analyses to determine the surface radionuclide concentrations and to characterize the radionuclide distributions in the samples. The analysis results indicate that the situation of contamination in the building of Unit 2 was different from others, and the protective surface coatings on the concrete floors provided significant protection against radionuclide penetration. contaminants.
Tanaka, Kosuke; Miwa, Shuhei; Sato, Isamu; Hirosawa, Takashi; Sekine, Shinichi; Seki, Takayuki*; Tokoro, Daishiro*; Obayashi, Hiroshi; Koyama, Shinichi
JAEA-Research 2013-022, 62 Pages, 2014/01
In order to establish the method for heating tests focused on the fission product release resulting from the high temperature chemical interaction between fuel and cladding material and to obtain the novel data on fission product release behaviors, the heating test was carried out with irradiate MOX fuel pellet and cladding.