Refine your search:     
Report No.
 - 
Search Results: Records 1-20 displayed on this page of 356

Presentation/Publication Type

Initialising ...

Refine

Journal/Book Title

Initialising ...

Meeting title

Initialising ...

First Author

Initialising ...

Keyword

Initialising ...

Language

Initialising ...

Publication Year

Initialising ...

Held year of conference

Initialising ...

Save select records

Journal Articles

FEMAXI-7 analysis for modeling benchmark for FeCrAl

Yamaji, Akifumi*; Susuki, Naomichi*; Kaji, Yoshiyuki

IAEA-TECDOC-1921, p.199 - 209, 2020/07

The thermo-physical models and irradiation behavior of FeCrAl as defined by the benchmark organizer have been implemented to FEMAXI-7. Analyses were carried out firstly for the specified normal operation condition. Then, some sensitivity analyses were carried out with different assumptions and model parameters. Under the normal operating condition, the predicted FeCrAl cladded fuel performance was similar to that of Zry cladded fuel with notable, but not major difference regarding late gap closure. Under the simulated LOCA conditions, the burst pressure could be evaluated. The predicted cladding creep strain at burst was mainly attributed to creep strain with negligible plastic strain. Overall, FEMAXI-7 analyses have demonstrated excellent robustness and flexibility in modeling FeCrAl-UO$$_{2}$$ system under normal and LOCA conditions.

Journal Articles

Study on loss-of-cooling and loss-of-coolant accidents in spent fuel pool; Confirmation of fuel temperature calculation function with oxidation reaction in the SAMPSON code

Suzuki, Hiroaki*; Morita, Yoshihiro*; Naito, Masanori*; Nemoto, Yoshiyuki; Kaji, Yoshiyuki

Mechanical Engineering Journal (Internet), 7(3), p.19-00450_1 - 19-00450_17, 2020/06

In this study, the SAMPSON code was modified to evaluate severe accidents in a spent fuel pool (SFP). Air oxidation models based on oxidation data obtained on the Zircaroy-4 cladding (ANL model) and the Zircaroy-2 cladding (JAEA model) were included in the modified SAMPSON code. Experiments done by Sandia National Laboratory using simulated fuel assemblies equivalent to those of an actual BWR plant were analyzed by the modified SAMPSON code to confirm the functions for analysis of the severe SFP accidents. The rapid fuel rod temperature rise due to the Zr air oxidation reaction could be reasonably evaluated by the SAMPSON analysis. The SFP accident analyses were conducted with different initial water levels which were no water, water level at bottom of active fuel, and water level at half of active fuel. The present analysis showed that the earliest temperature rise of the fuel rod surface occurred when there was no water in the SFP and natural circulation of air became possible.

Journal Articles

Expansion of high temperature creep test data for failure evaluation of BWR lower head in severe accident

Yamaguchi, Yoshihito; Katsuyama, Jinya; Kaji, Yoshiyuki; Osaka, Masahiko; Li, Y.

Mechanical Engineering Journal (Internet), 7(3), p.19-00560_1 - 19-00560_12, 2020/06

Since the Fukushima Daiichi nuclear power plant accident, we have been developing a failure evaluation method that considers creep damage mechanisms using detailed three-dimensional finite element analysis model of lower head including penetration, stub tubes, and weld parts, etc., for the early completion of the decommissioning of the nuclear power plants in Fukushima Daiichi. For the finite element analysis, we have been obtaining material properties for which no data are provided in existing databases or in the literature. In particular, creep data corresponding to the high temperature region near the melting point of materials is important in evaluating creep deformation under severe accident conditions. In this study, we obtained the uniaxial tensile and creep properties for low-alloy steel, stainless steel, and Ni-based alloy. In particular, creep test data with long rupture times at high temperatures are expanded using a tensile test machine that can measure the elongation of test specimens in a noncontact measurement system. The parameters related to the failure evaluation were improved on the basis of the expanded creep database.

Journal Articles

Corrosion monitoring in humidity-controlled environment simulating gamma ray irradiation

Omori, Atsushi*; Akiyama, Eiji*; Abe, Hiroshi*; Hata, Kuniki; Sato, Tomonori; Kaji, Yoshiyuki; Inoue, Hiroyuki*; Taguchi, Mitsumasa*; Seito, Hajime*; Tada, Eiji*; et al.

Zairyo To Kankyo, 69(4), p.107 - 111, 2020/04

To evaluate the effect of oxidants, which are formed by radiolysis of water under gamma ray irradiation, on the corrosion of a carbon steel in humid environment, ozone was introduced as a model oxidant in to humidity-controlled air at 50$$^{circ}$$C in a thermo-hygrostat chamber. Corrosion monitoring was performed by using an Atmospheric Corrosion Monitor-type (ACM) sensor consisting of a carbon steel anode and an Ag cathode. The output current of the ACM sensor was increased with the increase in relative humidity and it was obviously increased with the increase in the introduced ozone concentration at each relative humidity. The results indicate that ozone accelerates the corrosion of the carbon steel. The effect of ozone on the corrosion acceleration is attributed to the fast reduction reaction and fast dissolution reaction in to water compared to that of oxygen.

Journal Articles

Overview of accident-tolerant fuel R&D program in Japan

Yamashita, Shinichiro; Ioka, Ikuo; Nemoto, Yoshiyuki; Kawanishi, Tomohiro; Kurata, Masaki; Kaji, Yoshiyuki; Fukahori, Tokio; Nozawa, Takashi*; Sato, Daiki*; Murakami, Nozomu*; et al.

Proceedings of International Nuclear Fuel Cycle Conference / Light Water Reactor Fuel Performance Conference (Global/Top Fuel 2019) (USB Flash Drive), p.206 - 216, 2019/09

After the nuclear accident at Fukushima Daiichi Power Plant, research and development (R&D) program for establishing technical basis of accident-tolerant fuel (ATF) started from 2015 in Japan. Since then, both experimental and analytical studies necessary for designing a new light water reactor (LWR) core with ATF candidate materials are being conducted within the Japanese ATF R&D Consortium for implementing ATF to the existing LWRs, accompanying with various technological developments required. Until now, we have accumulated experimental data of the candidate materials by out-of-pile tests, developed fuel evaluation codes to apply to the ATF candidate materials, and evaluated fuel behavior simulating operational and accidental conditions by the developed codes. In this paper, the R&D progresses of the ATF candidate materials considered in Japan are reviewed based on the information available such as proceedings of international conference and academic papers, providing an overview of ATF program in Japan.

Journal Articles

Benchmark of fuel performance codes for FeCrAl cladding behavior analysis

Pastore, G.*; Gamble, K. A.*; Cherubini, M.*; Giovedi, C.*; Marino, A.*; Yamaji, Akifumi*; Kaji, Yoshiyuki; Van Uffelen, P.*; Veshchunov, M.*

Proceedings of International Nuclear Fuel Cycle Conference / Light Water Reactor Fuel Performance Conference (Global/Top Fuel 2019) (USB Flash Drive), p.1038 - 1047, 2019/09

Oxidation-resistant iron-chromium-aluminum (FeCrAl) steels have been proposed for application as cladding materials in light water reactor fuel rods with improved accident tolerance. Within the Coordinated Research Project ACTOF of the International Atomic Energy Agency (IAEA), a fuel performance modeling benchmark for FeCrAl cladding behavior was conducted. During this effort, calculations were performed with various fuel performance codes for a set of fuel rod problems with FeCrAl steel as cladding material, and results were compared to each other.

JAEA Reports

Report of the erosion-corrosion of metallic materials under solid-liquid two phase flow

Otani, Kyohei; Sato, Tomonori; Kaji, Yoshiyuki; Yamamoto, Masahiro

JAEA-Review 2019-007, 15 Pages, 2019/06

JAEA-Review-2019-007.pdf:2.35MB

Metallic pipes under solid-liquid two phase flow is damaged by collision of solid particle to the pipe walls, and this phenomenon is named "erosion". In the case of the liquid is corrosive solution, further damage is occurred on the pipe walls chemically, and this named "erosion-corrosion". In the Fukushima Daiichi decommissioning project, the fuel debris will be crushed during removal operation of the debris and micro debris particles would be generated. It is estimated that the pipes of the circulating cooling system would be damaged under the solid-liquid two phase flow containing fuel debris particles. For the reason, the previous study about erosion and erosion-corrosion of metallic materials under solid-liquid two phase flow was surveyed. The survey showed that the damage rate by erosion and erosion-corrosion is influence by a lot of parameter in comparison to the corrosion rate which occurred in no-flow solution. Therefore, it is necessary to pay attention to selecting the experimental method and condition before the investigation about erosion-corrosion of metallic materials under solid-liquid two phase flow is carried out.

Journal Articles

Study on loss-of-cooling and loss-of-coolant accidents in spent fuel pool, 2; Fuel cladding oxidation

Nemoto, Yoshiyuki; Kaji, Yoshiyuki; Kanazawa, Toru*; Nakashima, Kazuo*; Tojo, Masayuki*

Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 8 Pages, 2019/05

Oxidation behaviour of Zr cladding in SFP accident condition was evaluated by using a thermobalance in this work, and the obtained data were applied to construct oxidation model for SFP accident condition. For the validation of the constructed oxidation model, oxidation tests using a long cladding tube 500mm in length were conducted in conditions simulating SFP accidents, such as flow rate of the atmosphere in spent fuel rack, temperature gradient along the axis of cladding, and heating-up history. Thickness of oxide layer formed on the surface of cladding samples was evaluated by cross sectional observation, and compared with calculation results obtained by using the oxidation model. The detail of experimental results and validation of the oxidation model will be discussed.

Journal Articles

Study on loss-of-cooling and loss-of-coolant accidents in spent fuel pool, 4; Investigation of fuel loading effects in BWR spent fuel rack

Tojo, Masayuki*; Kanazawa, Toru*; Nakashima, Kazuo*; Iwamoto, Tatsuya*; Kobayashi, Kensuke*; Goto, Daisuke*; Nemoto, Yoshiyuki; Kaji, Yoshiyuki

Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 13 Pages, 2019/05

In this study, fuel loading effects in BWR spent fuel rack accidents are widely investigated using three-dimensional analysis methods from both nuclear and thermal hydraulics viewpoints, including: (a) Decay heat of spent fuel after discharge, (b) The maximum temperature of spent fuel cladding in the spent fuel rack depending on heat transfer phenomena, and (c) Criticality of the spent fuel rack after collapsing of the fuel due to a severe accidents in the BWR spent fuel pool (SFP).

Journal Articles

Expansion of high temperature creep test data for failure evaluation of BWR lower head in a severe accident

Yamaguchi, Yoshihito; Katsuyama, Jinya; Kaji, Yoshiyuki; Osaka, Masahiko; Li, Y.

Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 7 Pages, 2019/05

After the Fukushima Daiichi (1F) Nuclear Power Plant accident, we have been developing a prediction method for rupture time and location considering creep damage mechanisms using finite element analysis for early completion of the decommissioning of nuclear power plants in 1F. We have also been obtaining material properties at high temperature near the melting point which are not provided in existing database or literature for the finite element analysis. In this study, we performed uni-axial tensile and creep tests for low alloy steel, Ni-based alloy steel and stainless steels and expanded existing database of material properties. Especially, creep data with longer rupture time at high temperature were obtained by a creep test equipment with a noncontact measurement system. To improve the accuracy of failure evaluation under severe accident conditions, we determined parameters of creep constitutive law based on the expanded database.

Journal Articles

Study on loss-of-cooling and loss-of-coolant accidents in spent fuel pool, 1; Overview

Kaji, Yoshiyuki; Nemoto, Yoshiyuki; Nagatake, Taku; Yoshida, Hiroyuki; Tojo, Masayuki*; Goto, Daisuke*; Nishimura, Satoshi*; Suzuki, Hiroaki*; Yamato, Masaaki*; Watanabe, Satoshi*

Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 8 Pages, 2019/05

In this research program, cladding oxidation model in SFP accident condition, and numerical simulation method to evaluate capability of spray cooling system which was deployed for spent fuel cooling during SFP accident, have been developed. These were introduced into the severe accident codes such as MAAP and SAMPSON, and SFP accident analyses were conducted. Analyses using Computational Fluid Dynamics (CFD) code were conducted as well for the comparison with SA code analyses and investigation of detail in the SFP accident. In addition, three-dimensional criticality analysis method was developed as well, and safer loading pattern of spent fuels in pool was investigated.

Journal Articles

Study on loss-of-cooling and loss-of-coolant accidents in spent fuel pool, 5; Investigation of cooling effects of SFP spray and alternate water injection with MAAP code

Nishimura, Satoshi*; Satake, Masaaki*; Nishi, Yoshihisa*; Nemoto, Yoshiyuki; Kaji, Yoshiyuki

Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 8 Pages, 2019/05

In this study, accident progression analyses in the SFP were performed to investigate cooling effects of the SFP spray and an alternate water injection in the loss-of-pool water accident with MAAP ver. 5.05 beta. Fuel cladding oxidation model which was created by JAEA based on their experimental data was selected and applied in the present calculations. In case of an assessment of SFP spray effects, decay heat, spray fraction going into the fuel assembly, spray droplet diameter, spray start time were selected as analytical parameters. When the SFP spray of 12.5 kg/s (200 GPM) starts 4 hours after the onset of the accident against the spent fuels with 4 months cooling and if the spray fraction going into the fuel assembly is greater than 30%, the maximum cladding temperature can be maintained under 727$$^{circ}$$C (1000 K), resulting in avoiding the cladding failure.

Journal Articles

Study on loss-of-cooling and loss-of-coolant accidents in spent fuel pool, 6; Analysis on oxidation behavior of fuel cladding tubes by the SAMPSON code

Morita, Yoshihiro*; Suzuki, Hiroaki*; Naito, Masanori*; Nemoto, Yoshiyuki; Kaji, Yoshiyuki

Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 9 Pages, 2019/05

In this study, the SAMPSON code was modified to evaluate severe accidents in a spent fuel pool (SFP). Not only the SFP but also upper spaces of the SFP, walls of the reactor building, and the blowout panel were included. Air oxidation models obtained by the Zircaroy-4 cladding (ANL model) and the Zircaroy-2 cladding (JAEA model) were included in the modified SAMPSON code. Experiments done by Sandia National Laboratory using simulated fuel assemblies equivalent to those of an actual BWR plant were analyzed by the modified SAMPSON code to confirm the functions for analysis of the severe SFP accidents.

Journal Articles

Study on loss-of-cooling and loss-of-coolant accidents in spent fuel pool, 7; Analysis on effectiveness of spray cooling by the SAMPSON code

Suzuki, Hiroaki*; Morita, Yoshihiro*; Naito, Masanori*; Nemoto, Yoshiyuki; Nagatake, Taku; Kaji, Yoshiyuki

Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 7 Pages, 2019/05

In this paper, modification of the SAMPSON code was carried out to enable the analysis of spray cooling. The SAMPSON analysis of a spray cooling experiment was performed to confirm reproducibility of spray cooling behavior of fuel claddings. The modified SAMPSON code was applied to a hypothetical loss-of-coolant accident analysis of the SFP. Effectiveness of spray cooling on cladding temperature behavior was investigated. The SAMPSON analysis showed that spraying from the top of the SFP was effective for cooling the fuel assemblies exposed to the gas phase.

Journal Articles

Study on loss-of-cooling and loss-of-coolant accidents in spent fuel pool, 8; Safety margin of spent fuel in large LOCA event by the simple assessment method

Someya, Takayuki*; Chitose, Hiromasa*; Watanabe, Satoshi*; Nemoto, Yoshiyuki; Kaji, Yoshiyuki

Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 9 Pages, 2019/05

In this study, CFD analysis has been conducted for the assessment of spent fuel integrity in large LOCA event and the maximum temperature of spent fuel assemblies has been evaluated. Then, it has been compared with the result of the simple assessment method. As a case study, additional CFD analysis has been conducted, where water level in SFP decreases to the Bottom of Active Fuel (BAF) due to boil-off. Since this scenario might be more severe than large LOCA scenario, the number of spent fuel assemblies, their decay heat and loading pattern to maintain spent fuel integrity are investigated.

Journal Articles

Evaluation of the effect of spent fuel layout on SFP cooling with MAAP5.04

Nishimura, Satoshi*; Satake, Masaaki*; Nishi, Yoshihisa*; Kaji, Yoshiyuki; Nemoto, Yoshiyuki

Proceedings of 11th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-11) (Internet), 3 Pages, 2018/11

After the accident of Fukushima-unit 1 Nuclear Power Plant, Japanese utilities are newly requested by regulatory body to take prompt measures to enhance the safety of spent fuel pool. The most important objective of this new Japanese standards of regulation is keeping a water level in a Spent Fuel Pool (SFP) under any situations in order to prevent fuel failures due to increase of fuel temperature and to avoid the occurrence of re-criticality accidents. The utilities are considered to install several kinds of safety measures for SFP. For example, a spray injection and an alternate water injection to keep pool water level, and a spent fuel layout, such as 1 by 4, 1 by 8, checkerboard to enhance cooling of the spent fuel in SFP. The objective of the present study is to investigate the effect of spent fuel layout on SFP cooling with MAAP5.04.

Journal Articles

Study on oxidation model for Zircalloy-2 cladding in SFP accident condition

Nemoto, Yoshiyuki; Kaji, Yoshiyuki; Onizawa, Takashi*; Kanazawa, Toru*; Nakashima, Kazuo*; Tojo, Masayuki*

Proceedings of Annual Congress of the European Federation of corrosion (EUROCORR 2018) (USB Flash Drive), 8 Pages, 2018/09

The authors proposed oxidation models based on oxidation data which previously obtained in high temperature oxidation tests on small sample of Zircalloy-2 (Zry2) cladding in dry air and in air/steam mixture environment. The oxidation models were implemented in computational fluid dynamics (CFD) code to analyse oxidation behavior of long cladding sample in hypothetical spent fuel pool (SFP) accident conditions. The oxidation tests were conducted using Zry2 cladding sample 500 mm in length. The oxide layer growth in dry air was well reproduced in the calculation using the oxidation model, meanwhile which in air/steam mixture was overestimated atmosphere composition change anticipated in the spent fuel rack during the accident, and its influence on the oxidation behaviour of the cladding were discussed in consideration of the oxidation model improvement.

Journal Articles

Effect of long-term thermal aging on SCC initiation susceptibility in low carbon austenitic stainless steels

Aoki, So; Kondo, Keietsu; Kaji, Yoshiyuki; Yamamoto, Masahiro

Proceedings of the 18th International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors, Vol.2, p.663 - 672, 2018/00

 Times Cited Count:1 Percentile:100

The objective of this study was to clarify effect of long-term thermal aging on SCC initiation susceptibility in low carbon austenitic stainless steels. Specimens used were Type 304L and 316L steels. Both steels were cold rolled to 20% thickness reduction (CW) and then followed by long-term thermal aging at 288$$^{circ}$$C for 14,000 h (LTA). Crevice Bent Beam (CBB) test was carried out to estimate the SCC initiation susceptibility under BWR simulated water condition at high temperature. The present results of the CBB tests showed that 304L CW + LTA exhibited no SCC susceptibility. In contrast, the SCC initiation susceptibility of 316L increased by the combination of cold rolling and long-term thermal aging. To understand these results, evaluation on the changes in microchemistry, microstructure and mechanical properties induced by the CW and LTA treatment has been developed, and their correlation with the SCC initiation susceptibility is discussed.

Journal Articles

Influence of the air/steam mixing ratio in atmosphere on zirconium cladding oxidation in spent fuel pool accident condition

Nemoto, Yoshiyuki; Kaji, Yoshiyuki; Ogawa, Chihiro*; Nakashima, Kazuo*; Tojo, Masayuki*

Proceedings of 2017 Water Reactor Fuel Performance Meeting (WRFPM 2017) (USB Flash Drive), 10 Pages, 2017/09

To cope with the hypothetical severe accident in spent fuel pools (SFPs), it is important to understand the high temperature oxidation behavior of the Zirconium claddings exposed in the air or in the atmosphere of air/steam mixture. In this study, oxidation tests on Zircaloy-2 (Zry2) and Zircaloy-4 (Zry4) short samples were conducted in the atmosphere of air - steam mixture, and mixing ratio was varied to evaluate its influence on the oxidation kinetics in the temperature range from 600 to 1100$$^{circ}$$C. From 900 to 1000$$^{circ}$$C for Zry2, and from 800 to 1000$$^{circ}$$C for Zry4, oxidation rates appeared higher in air - steam mixture than in dry air or in steam without air. This tendency was appeared more evident in post-breakaway transition phase after fracture of dense oxide layer on the surface of specimens. These results suggest importance of the oxidation model development in consideration of the air - steam mixture environment for the SFP accident analysis.

Journal Articles

Creep damage evaluations for BWR lower head in severe accident

Katsuyama, Jinya; Yamaguchi, Yoshihito; Nemoto, Yoshiyuki; Kaji, Yoshiyuki; Osaka, Masahiko

Transactions of 24th International Conference on Structural Mechanics in Reactor Technology (SMiRT-24) (USB Flash Drive), 11 Pages, 2017/08

It is difficult to assess rupture behavior of the lower head of reactor pressure vessel in boiling-water-type nuclear power plants due to severe accident like Fukushima Daiichi because Boiling Water Reactor (BWR) lower heads have geometrically complicated structure with a lot of penetrations. Therefore, we have been developing an analysis method to predict time and location of RPV lower head rupture of BWRs considering creep damage mechanisms based on coupled analysis of three-dimensional Thermal-Hydraulics (TH) and thermal-elastic-plastic-creep analyses. In this study, we performed creep damage evaluations to investigate the effects of the debris depth and heat generation locations on failure behavior of lower head. From the analysis results, we discussed the outflow paths of the relocated molten core to the containment, and it was concluded that failure regions of BWR lower head are only the control rod guide tubes or stub tubes under simulated conditions.

356 (Records 1-20 displayed on this page)