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Ito, Takashi; Higemoto, Wataru; Koda, Akihiro*; Nakamura, Jumpei*; Shimomura, Koichiro*
Interactions (Internet), 245(1), p.25_1 - 25_7, 2024/12
Tsai, Y. H.*; Kobata, Masaaki; Fukuda, Tatsuo; Tanida, Hajime; Kobayashi, Toru; Yamashita, Yoshiyuki*
Applied Physics Letters, 124(11), p.112105_1 - 112105_5, 2024/03
Times Cited Count:0Horii, Yuta; Hirooka, Shun; Uno, Hiroki*; Ogasawara, Masahiro*; Tamura, Tetsuya*; Yamada, Tadahisa*; Furusawa, Naoya*; Murakami, Tatsutoshi; Kato, Masato
Journal of Nuclear Materials, 588, p.154799_1 - 154799_20, 2024/01
Times Cited Count:1 Percentile:68.31(Materials Science, Multidisciplinary)The thermal conductivities of near-stoichiometric (U,Pu,Am)O doped with NdO/SmO, which is major fission product (FP) generated by a uranium-plutonium mixed oxides (MOX) fuel irradiation, as simulated fission products are evaluated at 1073-1673 K. The thermal conductivities are calculated from the thermal diffusivities that are measured using the laser flash method. To evaluate the thermal conductivity from a homogeneity viewpoint of Nd/Sm cations in MOX, the specimens with different homogeneity of Nd/Sm are prepared using two kinds of powder made by ball-mill and fusion methods. A homogeneous Nd/Sm distribution decreases the thermal conductivity of MOX with increasing Nd/Sm content, whereas heterogeneous Nd/Sm has no influence. The effect of Nd/Sm on the thermal conductivity is studied using the classical phonon transport model (A+BT). The dependences of the coefficients A and B on the Nd/Sm content (C and C, respectively) are evaluated as: A(mK/W)=1.70 10 + 0.93C + 1.20C, B(m/W)=2.39 10.
Hirota, Noriaki; Nakano, Hiroko; Fujita, Yoshitaka; Takeuchi, Tomoaki; Tsuchiya, Kunihiko; Demura, Masahiko*; Kobayashi, Yoshinao*
The IV International Scientific Forum "Nuclear Science and Technologies"; AIP Conference Proceedings 3020, p.030007_1 - 030007_6, 2024/01
Dynamic strain aging (DSA) and intergranular stress corrosion cracking (intragranular SCC) occur in high temperature pressurized water simulating a boiling water reactor environment due to changes in dissolved oxygen (DO) content, respectively. In order to clearly understand the difference between these phenomena, the mechanism of their occurrence was summarized. As a result, it was found that DSA due to intragranular cracking occurred in SUS304 stainless steel at low DO 1 ppb, while DSA was suppressed at DO 100 to 8500 ppb due to the formation of oxide films on the surface. On the other hand, when DO was increased to 20000 ppb, the film was peeled from the matrix, O element diffused to the grain boundary of the matrix, resulting in intergranular SCC. These results are indicated that the optimum DO concentration must be adjusted to suppress crack initiation due to DSA and intergranular SCC.
McGrady, J.; Kumagai, Yuta; Watanabe, Masayuki; Kirishima, Akira*; Akiyama, Daisuke*; Kimuro, Shingo; Ishidera, Takamitsu
Journal of Nuclear Science and Technology, 60(12), p.1586 - 1594, 2023/12
Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)Ito, Takashi; Higemoto, Wataru; Shimomura, Koichiro*
Physical Review B, 108(22), p.224301_1 - 224301_11, 2023/12
Times Cited Count:1 Percentile:0(Materials Science, Multidisciplinary)Yoshida, Kazuo; Tamaki, Hitoshi; Hiyama, Mina*
JAEA-Research 2023-001, 26 Pages, 2023/05
An accident of evaporation to dryness by boiling of high-level radioactive liquid waste (HLLW) is postulated as one of the severe accidents caused by the loss of cooling function 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 the atmosphere. Accurate quantitative estimation of released Ru is one of the important issues for risk assessment of those facilities. To resolve this issue, an analytical approach has been developed using computer simulation programs to assess the radioactive source term from those facilities. The proposed approach consists analyses with three computer programs. At first, the simulation of boiling behavior in the HLLW tank is conducted with SHAWED code. Next step, the thermal-hydraulic behavior in the facility building is simulated with MELCOR code based on the results at the first step simulation such as flowed out mixed steam flow rate, temperature and volatilized Ru from the tank. The final analysis step is carried out for estimating amount of released radioactive materials with SCHERN computer code which simulates chemical behaviors of nitric acid, nitrogen oxide and Ru based on the condition also simulated MELCOR. Series of sample simulations of the accident at a hypothetical typical facility are presented with the data transfer between those codes in this report.
Maamoun, I.; Rushdi, M.*; Falyouna, O.*; Eljamal, R.*; Eljamal, O.*
Separation and Purification Technology, 308, p.122863_1 - 122863_16, 2023/03
Times Cited Count:3 Percentile:32.55(Engineering, Chemical)Collaborative Laboratories for Advanced Decommissioning Science; University of Fukui*
JAEA-Review 2022-046, 108 Pages, 2023/01
The Collaborative Laboratories for Advanced Decommissioning Science (CLADS), Japan Atomic Energy Agency (JAEA), had been conducting the Nuclear Energy Science & Technology and Human Resource Development Project (hereafter referred to "the Project") in FY2021. The Project aims to contribute to solving problems in the nuclear energy field represented by the decommissioning of the Fukushima Daiichi Nuclear Power Station, Tokyo Electric Power Company Holdings, Inc. (TEPCO). For this purpose, intelligence was collected from all over the world, and basic research and human resource development were promoted by closely integrating/collaborating knowledge and experiences in various fields beyond the barrier of conventional organizations and research fields. The sponsor of the Project was moved from the Ministry of Education, Culture, Sports, Science and Technology to JAEA since the newly adopted proposals in FY2018. On this occasion, JAEA constructed a new research system where JAEA-academia collaboration is reinforced and medium-to-long term research/development and human resource development contributing to the decommissioning are stably and consecutively implemented. Among the adopted proposals in FY2021, this report summarizes the research results of the "Clarification of debris formation conditions on the basis of the sampling data and experimental study using simulated fuel debris and reinforcement of the analytical results of severe accident scenario" conducted in FY2021. The research on fuel debris so far is based on TMI-2 accident that is typical PWR accident but resent scenario analysis of sever accident progression and sampling data of the in reactor materials predict that fuel debris is diversity and piled up complicatedly depending on the unit and in reactor position. We are necessary to presume the thermodynamic condition of fuel debris during the accident in order to estimate accumulation state of debris.
Watanabe, Masashi; Kato, Masato
Frontiers in Nuclear Engineering (Internet), 1, p.1082324_1 - 1082324_9, 2023/01
Since the oxygen potential and the oxygen coefficient of UO have a significant impact on fuel performance, many experimental data have been obtained. However, experimental data of the oxygen potential and the oxygen diffusion coefficient in the high temperature region above 1673 K are very limited. In the present study, we aimed to obtain these data and analyze them by defect chemistry. The oxygen potentials and the oxygen chemical diffusion coefficient of UO were measured by the gas equilibrium method in the near stoichiometric region at temperatures ranging from 1673 to 1873 K. A data set of oxygen potentials was made together with literature data and analyzed by defect chemistry. The oxygen potential of UO was determined as a function of O/U ratio and temperature, and an equation representing the relationship was derived. The oxygen chemical diffusion coefficient values obtained in this study were reasonably close to the literature values. The oxygen partial pressure dependence of the oxygen chemical diffusion coefficients was predicted from the evaluated results of the oxygen potential data, but no clear dependence was observed.
Fukushima, Masahiro; Okajima, Shigeaki*; Mukaiyama, Takehiko*
Journal of Nuclear Science and Technology, 20 Pages, 2023/00
Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)A series of integral experiments was conducted to evaluate the fission and the capture cross- sections of transuranic (TRU) nuclides at the fast critical facility FCA of the Japan Atomic Energy Agency (JAEA). The experiments were carried out using seven uranium-fueled assemblies of the FCA. The neutron energy spectra of the core regions were adjusted so as to change from an intermediate neutron spectrum to a fast neutron spectrum on an assembly-by-assembly basis. The integral data measured with these experimental configurations provide some neutron energy characteristics: 1) fission rate ratios (FRRs) of Np, Pu, Pu, Am, Am, and Cm relative to Pu by using absolutely calibrated fission chambers, 2) small sample reactivity worths (SRWs) of Np, Pu, Pu, Am, and Am where oxide powders of around 15 to 20 grams were used, 3) criticalities, and 4) spectral indices such as fission rate ratios of U relative to U. In this paper, details of the SRW measurements are reported, and the latest Japanese Evaluated Nuclear Data Library JENDL-5 is tested by using the integral data obtained in systematically varied neutron energy spectra.
Idham, M. F.*; Falyouna, O.*; Eljamal, R.*; Maamoun, I.; Eljamal, O.*
Journal of Water Process Engineering (Internet), 50, p.103289_1 - 103289_16, 2022/12
Times Cited Count:13 Percentile:91.35(Engineering, Environmental)Akiyama, Daisuke*; Kusaka, Ryoji; Kumagai, Yuta; Nakada, Masami; Watanabe, Masayuki; Okamoto, Yoshihiro; Nagai, Takayuki; Sato, Nobuaki*; Kirishima, Akira*
Journal of Nuclear Materials, 568, p.153847_1 - 153847_10, 2022/09
Times Cited Count:3 Percentile:66.21(Materials Science, Multidisciplinary)FeUO, CrUO, and FeCrUO are monouranates containing pentavalent U. Even though these compounds have similar crystal structures, their formation conditions and thermal stability are significantly different. To determine the factors causing the difference in thermal stability between FeUO and CrUO, their crystal structures were evaluated in detail. A Raman band was observed at 700 cm in all the samples. This Raman band was derived from the stretching vibration of the O-U-O axis band, indicating that FeCrUO was composed of a uranyl-like structure in its lattice regardless of its "x"' value. Mssbauer measurements indicated that the Fe in FeUO and FeCrUO were trivalent. Furthermore, FeCrUO lost its symmetry around Fe with increasing electron densities around Fe, as the abundance of Cr increased. These results suggested no significant structural differences between FeUO and CrUO. Thermogravimetric measurements for UO, FeUO, and CrUO showed that the temperature at which FeUO decomposed under an oxidizing condition (approximately 800 C) was significantly lower than the temperature at which the decomposition of CrUO started (approximately 1250 C). Based on these results, we concluded that the decomposition of FeUO was triggered by an "in-crystal" redox reaction, i.e., Fe U Fe U, which would not occur in the CrUO lattice because Cr could never be reduced under the investigated condition. Finally, the existence of Cr in FexCrUO effectively suppressed the decomposition of the FeCrUO crystal, even at a very low Cr content.
Nakashio, Nobuyuki*; Osugi, Takeshi; Kurosawa, Shigenobu; Ishikawa, Joji; Hemmi, Ko; Iketani, Shotaro; Yokobori, Tomohiko
JAEA-Technology 2022-016, 47 Pages, 2022/08
The Nuclear Science Research Institute (NSRI) of the Japan Atomic Energy Agency (JAEA) started operation of the Advanced Volume Reduction Facilities (AVWF) for production of waste packages for disposal of low-level radioactive solid wastes (LLW). To clarify the operating conditions for homogenization of non-metallic LLW, preliminary tests were carried out using the plasma melting furnace of the non-metal melting unit. The fluidity of molten waste influences homogenization conditions of solidified products. It was clarified that the viscosity, which is determined by the chemical composition and the melting temperature, influence the fluidity of molten waste greatly through previous literature review and the small-scale melting tests. In the preliminary tests, the simulated waste with a cold tracer loaded in 200 L drums were melted. Using the waste chemical components (basicity, iron oxide concentration) as an experimental parameter, the homogeneity of the chemical components of the solidified product was investigated and the homogenization conditions of melting tests were examined. The retention ratio of the tracer in the molten bath was also confirmed. The viscosity of the molten wastes was measured and the correlation with homogeneity was examined. In addition, the technical requirements that should be concerned in advance for future actual operation were discussed.
Rizaal, M.; Nakajima, Kunihisa; Saito, Takumi*; Osaka, Masahiko; Okamoto, Koji*
ACS Omega (Internet), 7(33), p.29326 - 29336, 2022/08
Times Cited Count:3 Percentile:27.64(Chemistry, Multidisciplinary)Kirishima, Akira*; Akiyama, Daisuke*; Kumagai, Yuta; Kusaka, Ryoji; Nakada, Masami; Watanabe, Masayuki; Sasaki, Takayuki*; Sato, Nobuaki*
Journal of Nuclear Materials, 567, p.153842_1 - 153842_15, 2022/08
Times Cited Count:5 Percentile:76.47(Materials Science, Multidisciplinary)To understand the chemical structure and stability of nuclear fuel debris consisting of UO, Zr, and Stainless Steel (SUS) generated by the Fukushima Daiichi Nuclear Power Plant accident in Japan in 2011, simulated debris of the UO-SUS-Zr system and other fundamental component systems were synthesized and characterized. The simulated debris were synthesized by heat treatment for 1 to 12 h at 1600C, in inert (Ar) or oxidative (Ar + 2% O) atmospheres. Np and Am tracers were doped for the leaching tests of these elements and U from the simulated debris. The characterization of the simulated debris was conducted by XRD, SEM-EDX, Raman spectroscopy, and Mssbauer spectroscopy, which provided the major uranium phase of the UO -SUS-Zr debris was the solid solution of UO (s.s.) with Zr(IV) and Fe(II) regardless of the treatment atmosphere. The long-term immersion test of the simulated debris in pure water and that in seawater revealed the macro scale crystal structure of the simulated debris was chemically very stable in the wet condition for a year or more. Furthermore, the leaching test results showed that the actinide leaching ratios of U, Np, Am from the UO-SUS-Zr debris were very limited and less than 0.08 % for all the experiments in this study.
Collaborative Laboratories for Advanced Decommissioning Science; Tohoku University*
JAEA-Review 2022-009, 73 Pages, 2022/06
The Collaborative Laboratories for Advanced Decommissioning Science (CLADS), Japan Atomic Energy Agency (JAEA), had been conducting the Nuclear Energy Science & Technology and Human Resource Development Project (hereafter referred to "the Project") in FY2020. The Project aims to contribute to solving problems in the nuclear energy field represented by the decommissioning of the Fukushima Daiichi Nuclear Power Station, Tokyo Electric Power Company Holdings, Inc. (TEPCO). For this purpose, intelligence was collected from all over the world, and basic research and human resource development were promoted by closely integrating/collaborating knowledge and experiences in various fields beyond the barrier of conventional organizations and research fields. The sponsor of the Project was moved from the Ministry of Education, Culture, Sports, Science and Technology to JAEA since the newly adopted proposals in FY2018. On this occasion, JAEA constructed a new research system where JAEA-academia collaboration is reinforced and medium-to-long term research/development and human resource development contributing to the decommissioning are stably and consecutively implemented. Among the adopted proposals in FY2018, this report summarizes the research results of the "Basic research on the stability of fuel debris including alloy phase" conducted from FY2018 to FY2021 (this contract was extended to FY2021). Since the final year of this proposal was FY2021, the results for four fiscal years were summarized. The present study focus on fuel debris consisting of oxide phase and alloy phase generated by the high temperature chemical reaction between structure materials (SUS pipes, pressure vessels, etc.) and fuels (melted fuels, claddings components, etc.). We synthesize the simulated debris of UO-SUS system and UO-Zr(ZrO)-SUS system by high-temperature heat treatment, and measure their chemical property and dissolution behavior in water.
Eljamal, R.*; Maamoun, I.; Bensaida, K.*; Yilmaz, G.*; Sugihara, Yuji*; Eljamal, O.*
Renewable and Sustainable Energy Reviews, 158, p.112192_1 - 112192_13, 2022/04
Times Cited Count:25 Percentile:95.43(Green & Sustainable Science & Technology)Yoshida, Kazuo; Tamaki, Hitoshi; Hiyama, Mina*
JAEA-Research 2021-013, 20 Pages, 2022/01
An accident of evaporation to dryness by boiling of high level liquid waste (HLLW) is postulated as one of the severe accidents caused by the loss of cooling function 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. An idea has been proposed to implement a steam condenser as an accident countermeasure. This measure is expected to prevent nitric acid steam diffusing in facility building and to increase gaseous Ru trapping ratio into condensed water. A simulation study has been carried out with a hypothetical typical facility building to analyze the efficiency of steam condenser. In this study, SCHERN computer code simulates chemical behaviors of Ru in nitrogen oxide, nitric acid and water mixed vapor based on the conditions obtained from simulation with thermal-hydraulic computer code MELCOR. The effectiveness of steam condenser has been analyzed quantitively in preventing mixed vapor diffusion and gaseous Ru trapping effect. Some issues to be solved in analytical model has been also clarified in this study.
Yoshida, Naoki; Ono, Takuya; Yoshida, Ryoichiro; Amano, Yuki; Abe, Hitoshi
JAEA-Research 2021-011, 12 Pages, 2022/01
In boiling and drying accidents involving high-level liquid waste in fuel reprocessing plants, emphasis is placed on the behavior of ruthenium (Ru). Ru would form volatile species, such as ruthenium tetroxide (RuO), and could be released to the environment with coexisting gases, including nitric acid, water, or nitrogen oxides. In this study, to contribute toward safety evaluations of these types of accidents, the migration behavior of gaseous Ru into the liquid phase has been experimentally measured by simulating the condensate during an accident. The gas absorption of RuO was enhanced by increasing the nitrous acid (HNO) concentration in the liquid phase, indicating the occurrence of chemical absorption. In control experiments without HNO, the lower the temperature, the greater was the Ru recovery ratio in the liquid phase. Conversely, in experiments with HNO, the higher the temperature, the higher the recovery ratio, suggesting that the reaction involved in chemical absorption was activated at higher temperatures.