Yamane, Yuichi; Amano, Yuki; Tashiro, Shinsuke; Abe, Hitoshi; Uchiyama, Gunzo; Yoshida, Kazuo; Ishikawa, Jun
Journal of Nuclear Science and Technology, 53(6), p.783 - 789, 2016/06
The release behavior of radioactive materials from high active liquid waste (HALW) has been experimentally investigated under boiling accident conditions. In the experiments using HALW obtained through laboratory scale reprocessing, release ratio was measured for the FP nuclides such as Ru, Tc, Cs, Sr, Nd, Y, Mo, Rh and actinides such as Cm, Am. As a result, the release ratio was 0.20 for Ru and 1 for the FP and Ac nuclides. Ru was released into the gas phase in the form of both mist and gas. For its released amount, weak dependency was found to the initial concentration in the test solution. The release ratio decreased with the initial concentration. For other FP nuclides and actinides as non-volatile, released into the gas phase in the form of mist, the released amount increased with the initial concentration. The release ratio of Ru and NOx concentration increased with temperature of the test solutions. They were released almost at the same temperature between 200 and 300C. Size distribution of the mist and other particle was measured.
Uchiyama, Gunzo; Tashiro, Shinsuke; Amano, Yuki; Abe, Hitoshi; Yamane, Yuichi; Yoshida, Kazuo; Ishikawa, Jun
Proceedings of 21st International Conference & Exhibition; Nuclear Fuel Cycle for a Low-Carbon Future (GLOBAL 2015) (USB Flash Drive), p.1056 - 1063, 2015/09
The experimental study for source term data of radioactive materials has been conducted at a boiling accident of high active liquid waste (HALW) in reprocessing plants. In the study, three kinds of tests have been conducted including a cold small scale test, a cold engineering scale test and a hot small scale test. The following results were obtained: Ruthenium and Technetium were released into the gas phase in the form of both mist and gas under the boiling accident conditions of a simulated HALW. Non-volatile fission products (FPs) such as Nd and Cs were released into the gas phase in the form of mist. The release ratios of non-volatile FPs from a vessel of the simulated HALW were about 10. The release ratios of actinide nuclides such as Am were almost the same as those of non-volatile FPs.
Amano, Yuki; Watanabe, Koji; Tashiro, Shinsuke; Yamane, Yuichi; Ishikawa, Jun; Yoshida, Kazuo; Uchiyama, Gunzo; Abe, Hitoshi
Nihon Genshiryoku Gakkai Wabun Rombunshi, 14(2), p.86 - 94, 2015/06
Radioactive materials could be released into air due to the accidental boiling of high active liquid waste (HALW) in reprocessing plants. Volatile radioactive nuclides, such as ruthenium, are released from the tanks into the atmosphere. Nitrogen oxides (NOx) are also released due to the thermal decomposition of metal nitrates in HALW. The released NOx transport volatile ruthenium and cause redox reactions associated with the composition or decomposition of volatile ruthenium. In this study, NOx release data were obtained by heating simulated HALW up to 600C. As a result, the release of NOx from the simulated HALW was observed from 200C to 600C, and the main release of NOx was observed at about 340C. All the lanthanide nitrates were found to decompose in the simulated HALW, and the thermal decomposition temperature of the lanthanide nitrates decreased after the addition of ruthenium dioxide to the mixed lanthanide nitrates solution.
Ishijima, Yasuhiro; Ueno, Fumiyoshi
Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 4 Pages, 2015/05
In this study, to evaluate the effect of thermal aging on creep properties of Alloy 625, we carried out creep tests on aged and solution-treated Alloy 625 at 1073 K. According to the creep test results, time-to-rupture decreased by thermal aging when test stress was more than 100 MPa, but did not change when test stress was less than 100 MPa for any specimens. In the solution-treated alloy, creep deformation behaviors changed over 100 MPa. These results show that time-to-rupture was constant because intermetallic compounds precipitated when the test stress was less than 100 MPa in solution-treated alloy. The observed relationship between creep strain rate and test time showed that the precipitation started after 100 hr for solution treated alloys. These results suggest that intermetallic compounds precipitate immediately after furnace operation. And it is appropriate to use creep data of thermal-aged Alloy 625 for the reducing roasting furnace lifetime prediction.
Ando, Ryosuke; Abe, Teruyoshi; Nakamura, Takahisa
E-Journal of Advanced Maintenance (Internet), 6(4), p.153 - 164, 2015/02
Wall thinning of serviced carbon steel piping of the secondary cooling system after long term operation of Advanced Thermal prototype Reactor (ATR) Fugen power station has been investigated as a series of evaluation of validity and availability of utilization of serviced materials on research projects focused on aging management. Reliability of wall thinning rates of the steel piping has been examined referring the previous inspection data. Examinations also have been made on prediction of wall thinning rates, rationalization of management of pipe wall thinning and verification of countermeasures against wall thinning.
Abe, Teruyoshi; Nogiwa, Kimihiro*; Onitsuka, Takashi*; Nakamura, Takahisa; Sakakibara, Yasuhide
E-Journal of Advanced Maintenance (Internet), 6(4), p.146 - 152, 2015/02
Thermal embrittlement of cast austenitic stainless steel components from the decommissioned advanced thermal prototype reactor Fugen has been characterized. Cast stainless steel materials were obtained from recirculation pump casing. The actual time at temperature for the materials was 138,000 h at 275C. The "Fugen" material show modest decrease in Charpy-impact properties and a small increase in micro-Vickers hardness in ferrite phase because of thermal aging at relatively low service temperatures. The fracture toughness prediction method (H3T model) predicts slightly lower values for Charpy-impact energy obtained from the Fugen material. The results from microstructural analysis suggest that the prediction method have the potential to provide higher accuracy by considerations of the activation energy for embrittlement at low service temperatures.
Shimada, Taro; Tanaka, Tadao
Journal of Radioanalytical and Nuclear Chemistry, 303(2), p.1345 - 1349, 2015/02
In order to understand the production and dispersion behavior of radioactive aerosols during dismantling of nuclear facilities, plasma arc cutting experiments were conducted. Particle size distribution of the aerosols was obtained by sampling air into ELPI which could classify particles into 12 stages of 50% cutoff aerodynamic diameter ranging from 0.007 to 9.9 m. Co specific radioactivity of diameter 0.05 m during cutting of surface contaminated piping indicated the maximum value of approximately 2.7E+4 Bq/g which was fifty times as much as the average value of all of aerosols. That of 9.9m was approximately 100 Bq/g which was the eighth part of the average value. Compared with those for the activated piping, the difference of specific radioactivity between maximum and minimum values were larger in contaminated piping. It is considered that contaminants on the piping were directly melted and vaporized by plasma arc and then condensed into smaller particles.
Katsuyama, Jinya; Yamaguchi, Yoshihito; Li, Y.*; Onizawa, Kunio
Nuclear Engineering and Design, 278, p.222 - 228, 2014/10
Weld residual stress is one of the most important factors in stress corrosion cracking (SCC) of nuclear reactor piping. To assess the integrity of piping, it is necessary to understand the effects of excessive cyclic loading, caused by earthquake, on residual stress. In this study, finite element analyses were performed using a model of a 250A pipe butt weld of Type 316L stainless steel. The accuracy of the welding simulation was verified by comparing the calculated results with the experimental measurements. After conducting the welding simulation and residual stress analysis, several loading patterns for the axial cyclic loadings were applied to the model by varying the amount of maximum load, in order to study the effect of excessive cyclic loading. The analysis indicated that higher loading caused a greater relaxation of the weld residual stress near the piping welds. It was thus concluded that excessive cyclic loading on piping butt welds affects the suppression of SCC growth.
Yamaguchi, Yoshihito; Katsuyama, Jinya; Udagawa, Makoto; Onizawa, Kunio; Nishiyama, Yutaka; Li, Y.*
Proceedings of 22nd International Conference on Nuclear Engineering (ICONE-22) (DVD-ROM), 8 Pages, 2014/07
Uchiyama, Gunzo; Tashiro, Shinsuke; Amano, Yuki; Abe, Hitoshi; Yamane, Yuichi; Yoshida, Kazuo; Ishikawa, Jun
Proceedings of International Waste Management Symposia 2014 (WM 2014) (Internet), 9 Pages, 2014/05
The experimental study for source term data of radioactive materials has been conducted at a boiling accident of high active liquid waste (HALW) in a reprocessing plant. In the small scale cold test using a non-radioactive simulated HALW, the release behavior of FP elements from the simulated HALW were investigated under various boiling accident conditions. In the engineering scale cold test, the release behavior of FP elements at boiling accident conditions was investigated mainly as a spatial function. In the small scale hot test using a radioactive simulated HALW, the release behavior of radioactive materials (FP, alpha nuclides) were obtained under typical boiling accident conditions. In the small scale hot test, the release fractions of Ru and non-volatile FPs obtained were almost the same as those of the small scale cold test.
Tanaka, Tadao; Shimada, Taro; Sukegawa, Takenori
Progress in Nuclear Science and Technology (Internet), 4, p.832 - 835, 2014/04
According to a basic policy of Japan, nuclear power plant sites are allowed to be released from nuclear safety regulations after the plants are decommissioned. It is necessary to confirm that there is no significant radioactivity remaining on the sites, for the site release beforehand. Cobalt 60 is one of the typical radionuclide for nuclear power plants. In the evaluation concept, all of cobalt 60, which is in reality distributed across the area of interest, are assumed to be the single point source located at the furthest position on the surface of the area from a Ge detector. In such a configuration, minimum detectable time was supplied by Monte Carlo calculations, and the minimum detectable time was approximately equal to the actual measurement time of the point source by the Ge detector. These results mean that the proposed evaluation method was reasonable for the conservative evaluation of cobalt 60 remaining in the nuclear power plant sites.
Munakata, Masahiro; Amano, Kenji; Tanaka, Tadao
JNES-RE-2013-9032, p.36 - 54, 2014/02
no abstracts in English
Masaki, Koichi; Katsuyama, Jinya; Onizawa, Kunio
Journal of Pressure Vessel Technology, 136(1), p.011208_1 - 011208_7, 2014/02
In order to apply a probabilistic fracture mechanics (PFM) analysis to the structural integrity assessment of a reactor pressure vessel (RPV), PFM analysis code has been developed at JAEA. Using the PFM analysis code, PASCAL version 3, the conditional probabilities of crack initiation and fracture for an RPV during pressurized thermal shock events have been analyzed. Sensitivity analyses on some input parameters were performed to clarify the effect on the conditional fracture probability. Comparison between the conditional probabilities and temperature margin from current deterministic analysis method were made for some model plant conditions of domestic typical old-type RPVs. From the analyses, a good correlation between temperature margin and the conditional probability of crack initiation was obtained.
Nogiwa, Kimihiro; Onitsuka, Takashi; Abe, Teruyoshi; Sakakibara, Yasuhide; Horie, Kaoru; Nakamura, Takahisa
Journal of Nuclear Science and Technology, 50(9), p.883 - 890, 2013/09
The degree of influence of thermal aging on cast stainless steels over a protracted time at low temperatures was investigated by means of a toughness test and microstructural characterization performed on components dismantled from the advanced thermal proto-type reactor "Fugen". The thermal embrittlement mechanism was examined using data obtained dismantled from the materials that had aged on site. The results of a Charpy impact test and microstructural characterization performed using 3DAP analysis reveal early signs of a thermal aging effect over a protracted period at low temperatures corresponding boiling-water reactor (BWR).
Amano, Yuki; Tashiro, Shinsuke; Uchiyama, Gunzo; Abe, Hitoshi; Yamane, Yuichi; Yoshida, Kazuo
Proceedings of International Nuclear Fuel Cycle Conference; Nuclear Energy at a Crossroads (GLOBAL 2013) (CD-ROM), p.1411 - 1417, 2013/09
Yamaguchi, Yoshihito; Katsuyama, Jinya; Onizawa, Kunio; Li, Y.*; Yagawa, Genki*
Journal of Pressure Vessel Technology, 135(4), p.041406_1 - 041406_9, 2013/08
Since the seismic regulatory guide was revised and the Niigata-ken Chuetsu-oki earthquake occurred, attention is being drawn to the evaluation of the effects of large scale earthquakes for piping systems in which cracks may potentially occur. In this work, crack growth behaviors after excessive loading were experimentally and analytically evaluated for carbon steel and austenitic stainless steel. Increase and decrease pattern of load amplitude and maximum load of cyclic loading were applied to fatigue crack growth test. According to these results, retardation effect of crack growth was confirmed after excessive loading. In addition, applicability to the retardation effect of the modified Wheeler model was elucidated. We confirmed the retardation effect hardly influence the failure probability based on seismic loading by the probabilistic fracture mechanics (PFM) analyses with the Wheeler model.
Yamaguchi, Yoshihito; Katsuyama, Jinya; Onizawa, Kunio; Li, Y.*
Proceedings of 2013 ASME Pressure Vessels and Piping Conference (PVP 2013) (DVD-ROM), 8 Pages, 2013/07
We performed fatigue crack growth tests under constant amplitude cyclic loading with a single excessive tensile/compressive load. The stress distribution in front of crack tip and crack blunting were estimated by FEM analyses. After the crack tip was blunted by the excessive tensile loading, the effect of the excessive loading on crack growth rate varied depending on the magnitude of the subsequent compressive loading. When a compressive load is enough to close the crack, the crack growth rate became higher than that before the excessive tensile loading while increasing the tensile stress in front of crack tip. A crack growth prediction method has been proposed considering the effects of the excessive loading based on the variation of the stress distribution in front of crack tip and the crack blunting. The predicted crack growth rate by the proposed method was correlated with the experimental ones.
Takeuchi, Masayuki; Sano, Yuichi; Nakajima, Yasuo; Uchiyama, Gunzo; Nojima, Yasuo*; Fujine, Sachio*
Journal of Energy and Power Engineering, 7(6), p.1090 - 1096, 2013/06
The corrosion behavior of a titanium-5% tantalum alloy (Ti-5Ta) in hot nitric acid condensate was investigated to understand aging behavior of reprocessing equipments. On the basis of long-term immersion tests, it was determined that the corrosion of Ti-5Ta in nitric acid condensate is accelerated with an increase in the concentration. The corrosion rate was nearly constant during the immersion test and the coupons suffered from uniform corrosion. In addition, it is important to note that the nitric acid concentration in the condensate increased on addition of metal salts to the heated nitric acid solution. The larger valence of metal ions was contributed to the increase in the concentration of nitric acid condensate. Consequently, the metal salt in the heated nitric acid solution accelerates the corrosion of Ti-5Ta in the condensate. Therefore, the nitric acid condensate condition should be carefully considered for the corrosion environment of titanium and its alloys.
Yamaguchi, Yoshihito; Li, Y.*; Katsuyama, Jinya; Onizawa, Kunio
Nihon Kikai Gakkai Rombunshu, A, 79(802), p.730 - 734, 2013/06
In this study, an effect of excessive tensile and compressive loading on the crack growth behavior of piping has been evaluated through cracked plate testing. It was observed that excessive loading had changed crack growth rate. Effect on the crack growth behavior of excessive load was evaluated focusing on a crack blunting and a stress distribution near the crack tip. The crack growth evaluation method under seismic loading was proposed. Four-point bending seismic loading tests were performed using piping specimen in order to confirm the applicability of the proposed method to the piping. It was indicated that the crack growth behavior on piping could be evaluated conservatively using the proposed method, while the crack growth behavior due to seismic loading is evaluated non-conservatively by the existing method.
Kato, Chiaki; Ueno, Fumiyoshi; Yamamoto, Masahiro; Ban, Yasutoshi; Uchiyama, Gunzo; Nojima, Yasuo*; Fujine, Sachio*
ECS Transactions, 53(21), p.45 - 55, 2013/05
Neptunium ion contained as one of the fission products in reprocessing solutions is known as a corrosion accelerator of the stainless steel. But it is not clear why remarkable acceleration of corrosion is caused by a slight amount of the Np ion in boiling nitric acid solution. Neptunium has several oxidation states in nitric acid solution. These changeable oxidation states of Np in nitric acid solution are regarded as the cause. Therefore an evaluation of the electrochemical behaviors on stainless steel in nitric acid solution related to the oxidation state of Np is required in order to understand the corrosion acceleration mechanism. A specially designed electrochemical test cell integrated with optical cell for spectroscopic analysis was used for this purpose. From results of electrochemical tests, cathodic reaction on stainless steel was activated by Np ions. Np(VI) ion made the corrosion potential shift nobler than Np(V) and nobler corrosion potential causes increasing corrosion current and accelerating corrosion of stainless steel in nitric acid solution. Np(V) was easily oxidized to Np(VI) in nitric acid solution and Np(VI) was the stable state in boiling 3M-HNO. It was considered that role of Np ions was that of mediator to accelerate corrosion due to activating cathodic reaction and re-oxidizing cycle in boiling 3M-HNO.