Refine your search:     
Report No.
Search Results: Records 1-20 displayed on this page of 395

Presentation/Publication Type

Initialising ...


Journal/Book Title

Initialising ...

Meeting title

Initialising ...

First Author

Initialising ...


Initialising ...


Initialising ...

Publication Year

Initialising ...

Held year of conference

Initialising ...

Save select records

Journal Articles

Evaluation of core material energy change during the in-vessel phase of Fukushima Daiichi Unit 3 based on observed pressure data utilizing GOTHIC code analysis

Sato, Ikken; Arai, Yuta*; Yoshikawa, Shinji

Journal of Nuclear Science and Technology, 58(4), p.434 - 460, 2021/04

 Times Cited Count:6 Percentile:74.69(Nuclear Science & Technology)

Journal Articles

The Effect of base irradiation on failure behaviors of UO$$_{2}$$ and chromia-alumina additive fuels under simulated reactivity-initiated accidents; A Comparative analysis with FEMAXI-8

Udagawa, Yutaka; Mihara, Takeshi; Taniguchi, Yoshinori; Kakiuchi, Kazuo; Amaya, Masaki

Annals of Nuclear Energy, 139, p.107268_1 - 107268_9, 2020/05


 Times Cited Count:3 Percentile:37.46(Nuclear Science & Technology)

Journal Articles

Thresholds for failure of high-burnup LWR fuels by pellet cladding mechanical interaction under reactivity-initiated accident conditions

Udagawa, Yutaka; Sugiyama, Tomoyuki; Amaya, Masaki

Journal of Nuclear Science and Technology, 56(12), p.1063 - 1072, 2019/12

 Times Cited Count:5 Percentile:63.23(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Empirical equations of crack growth rates based on data fitting of neutron irradiated stainless steel under high temperature water simulating boiling water reactor core conditions

Kasahara, Shigeki; Chimi, Yasuhiro; Hata, Kuniki; Fukuya, Koji*; Fujii, Katsuhiko*

Proceedings of 19th International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors (Internet), p.1345 - 1355, 2019/08

This paper describes empirical equation development of crack growth rates (CGR) in consideration of IASCC of neutron irradiated stainless steel to contribute to structural integrity assessment of BWR reactor internals. Empirical equations of CGR (da/dt) were developed based on a formula of da/dt = M$$times$$K$$^{n}$$, assuming that "M" and "n" tend to be saturated with increasing neutron fluence. To obtain the empirical equations for normal water chemistry (NWC) and hydrogen water chemistry (HWC) of BWR, a data fitting with least square method was applied to the datasets consisting of F, K and CGR from post irradiation examinations of neutron irradiated stainless steel under simulated NWC and HWC conditions from open literature. As a result, calculated results by the equation for NWC showed good agreement with measured CGR data, meanwhile those for HWC did not. The above difference was seemed to be attributed that CGR data obtained under HWC conditions were scattered extensively.

Journal Articles

Empirical equations for tensile properties and stress-strain curves of neutron irradiated stainless steels in LWR conditions

Fukuya, Koji*; Fujii, Katsuhiko*; Chimi, Yasuhiro; Hata, Kuniki

Proceedings of 19th International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors (Internet), p.523 - 531, 2019/08

For structural integrity assessment on reactor internals of light water reactors, empirical equations of tensile properties as a function of neutron dose, and trend curves of stress-strain relations of neutron-irradiated austenitic stainless steels was proposed by fitting to recently developed database. The data in the database were obtained from reports of national projects in Japan and open literature, which was summarized in the form of data sheets. The empirical equations for tensile properties were formulated by using a saturation-type formulae. The equations were for CW 316 and SA 304/316 stainless steels in the temperature range of 280-350$$^{circ}$$C and the dose range up to 80 dpa. Stress-strain relation curves were reproduced based on the Swift model. Obtained calculated results by the empirical equations and stress-strain relations were reasonably well fitted to experimental data. The effects of composition and cold-working, etc. on tensile properties were discussed.

Journal Articles

An Interpretation of Fukushima-Daiichi Unit 3 plant data covering the two-week accident-progression phase based on correction for pressure data

Sato, Ikken

Journal of Nuclear Science and Technology, 56(5), p.394 - 411, 2019/05

 Times Cited Count:10 Percentile:75.66(Nuclear Science & Technology)

Water columns were adopted in the pressure measurement system of Fukushima-Daiichi Unit-3. Part of these water columns evaporated during the accident condition jeopardizing correct understanding on actual pressure. Through comparison of RPV (Reactor Pressure Vessel) and S/C pressures with D/W pressure, such water-column effect was evaluated. Correction for this effect was developed enabling clarification of slight pressure difference among RPV, S/C and D/W. This information was then integrated with other available data such as, water level, CAMS and environmental dose rate, into an interpretation of accident focusing on RPV and PCV pressurization/depressurization and radioactive material release to environment. It is suggested that dryout of in-vessel and ex-vessel debris was likely causing pressure decrease. S/C water poured into pedestal heated by relocated debris was the likely cause of pressurization. Cyclic reflooding of pedestal debris and dryout was likely.

Journal Articles

IV. Electrochemical measurements in various environments; Nuclear power plant I (Electrochemical measurement in nuclear power plant)

Ueno, Fumiyoshi

Zairyo To Kankyo, 68(1), p.2 - 8, 2019/01

It is important to control the cooling water of light water reactors (boiling water reactor and pressurized water reactor) to suitable quality in order to reduce corrosion of structural materials and generation of radioactive corrosion products. For that purpose, monitoring of water quality using electrochemical measurement method is necessary. In this article, the application of ECP measurement to BWR is mainly focused, I describe the water quality of light water reactors and the necessity of electrochemical measurement.

Journal Articles

Mass transfer inside narrow crevice of SUS316L in high temperature water

Yamamoto, Masahiro; Soma, Yasutaka; Igarashi, Takahiro; Ueno, Fumiyoshi

Proceedings of Annual Congress of the European Federation of Corrosion (EUROCORR 2018) (USB Flash Drive), 7 Pages, 2018/09

In order to clarify the SCC behavior of SUS316L under BWR environment, mass transfer inside crevice of SUS316L in high temperature water using various crevice gap samples was investigated. The samples were prepared by put together two SUS316L sheets. Crevice gap differs from 0.005 mm to 0.1 mm. Corrosion tests were conducted in 8 ppm dissolved oxygen (DO) conditions. Surface oxide film was analysed by laser Raman spectroscopy (LRS) after immersion. Numerical simulations were also conducted by using COMSOL Maltiphysics. Diffusion process of DO and the other chemical species were calculated with connected to electrochemical process. Electrical conductivities inside the crevice were 100 times larger than these of outer water. The reason of high conductivity is existence of Fe$$^{2+}$$ ions at the DO depletion crevice.

Journal Articles

In-situ measurement of electrical conductivity of solution within crevice of stainless steel in high temperature and high purity water

Soma, Yasutaka; Komatsu, Atsushi; Ueno, Fumiyoshi

Zairyo To Kankyo, 67(9), p.381 - 385, 2018/09

In-situ measurement of electrical conductivity of solution within crevice of SUS316L stainless steel in 288$$^{circ}$$C water has been conducted with newly developed electrochemical sensor system. The sensor measures local electrical conductivity of crevice solution beneath the electrode ($$kappa$$$$_{crev}$$) with electrochemical impedance method. The sensors were installed at different positions within tapered crevice of SUS316L stainless steel. The crevice specimen with the sensors were immerged into 288$$^{circ}$$C, 8 MPa, pure oxygen saturated high purity water for 100 h. $$kappa$$$$_{crev}$$ at a position with crevice gap of $$approx$$59.3$$mu$$m was 8-11$$mu$$S/cm, least deviate from conductivity of 288$$^{circ}$$C pure water (4.4$$mu$$S/cm) and no localized corrosion occurred. On the contrary, $$kappa$$$$_{crev}$$ at a position with crevice gap of $$approx$$4.4$$mu$$m increased with time and showed maximum value of $$approx$$1600$$mu$$S/cm at 70 h. Localized corrosion occurred in the vicinity of this position. Thermodynamic equilibrium calculation showed $$kappa$$$$_{crev}$$ of 1600$$mu$$S/cm being equivalent to pH of 3 to 3.7. It can be concluded that acidification occurred in tight crevice even under high purity bulk water and resulted in localized corrosion.

Journal Articles

Phenomenology of BWR fuel assembly degradation

Kurata, Masaki; Barrachin, M.*; Haste, T.*; Steinbrueck, M.*

Journal of Nuclear Materials, 500, p.119 - 140, 2018/03

 Times Cited Count:27 Percentile:66.83(Materials Science, Multidisciplinary)

Severe accidents occurred at the Fukushima-Daiichi Nuclear Power Station (FDNPS) required an immediate re-examination of fuel degradation phenomenology. The present paper reviews the updated knowledge on the phenomenology of the fuel degradation, focusing mainly on the BWR fuel assembly degradation at the macroscopic scale and that of the individual interactions at the meso-scale. Oxidation of boron carbide (B$$_{4}$$C) control rods potentially generate far larger amounts of heat and hydrogen under BWR accident conditions. All integral tests with B$$_{4}$$C control rods or control blades have shown early failure, liquefaction, relocation and oxidation of B$$_{4}$$C starting at temperatures around 1250$$^{circ}$$C, well below the significant interaction temperatures of UO$$_{2}$$-Zry. These interactions or reactions potentially influence the progress of fuel degradation in the early phase. The steam-starved conditions, which are being discussed as a likely scenario at the FDNPS accident, highly influence the individual interactions and potentially lead the fuel degradation in non-prototypical directions. The detailed phenomenology of individual interactions and their influence on the transient and on the late phase of the severe accidents are also discussed.

Journal Articles

Analysis of used BWR fuel assay data with the integrated burnup code system SWAT4.0

Tada, Kenichi; Kikuchi, Takeo*; Sakino, Takao; Suyama, Kenya

Journal of Nuclear Science and Technology, 55(2), p.138 - 150, 2018/02

 Times Cited Count:3 Percentile:30.9(Nuclear Science & Technology)

The criticality safety of the fuel debris in Fukushima Daiichi Nuclear Power Plant is one of the most important issues and the adoption of the burnup credit is desired for the criticality analysis. The assay data of used nuclear fuel irradiated in 2F2 is evaluated to validate SWAT4.0 for BWR fuel burnup problem. The calculation results revealed that number density of many heavy nuclides and FPs showed good agreement with the experimental data except for $$^{235}$$U, $$^{237}$$Np, $$^{238}$$Pu and Sm isotopes. The cause of the difference is assumption of the initial number density and void ratio and overestimation of the capture cross section of $$^{237}$$Np. The C/E-1 values do not depend on the types of fuel rods (UO$$_{2}$$ or UO$$_{2}$$-Gd$$_{2}$$O$$_{3}$$) and it is similar to that for the PWR fuel. These results indicate that SWAT4.0 appropriately analyzes the isotopic composition of the BWR fuel and it has sufficient accuracy to be adopted in the burnup credit evaluation of the fuel debris.

Journal Articles

Evaluation of crack growth rates and microstructures near the crack tip of neutron-irradiated austenitic stainless steels in simulated BWR environment

Chimi, Yasuhiro; Kasahara, Shigeki; Seto, Hitoshi*; Kitsunai, Yuji*; Koshiishi, Masato*; Nishiyama, Yutaka

Proceedings of the 18th International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors, Vol.2, p.1039 - 1054, 2018/00

 Times Cited Count:2 Percentile:58.64

In order to understand irradiation-assisted stress corrosion cracking (IASCC) growth behavior, crack growth rate (CGR) tests have been performed in simulated Boiling Water Reactor water conditions at $$sim$$288$$^{circ}$$C on neutron-irradiated 316L stainless steels (SSs) at $$sim$$12-14 dpa. After the tests, the microstructures near the crack tip of the specimens are examined with scanning transmission electron microscope (FE-STEM). In comparison with a previous study at $$<$$$$sim$$2 dpa, this result shows a less benefit of low electrochemical corrosion potential (ECP) conditions on CGR. A crack tip immersed over 1000 hours was filled with oxides, while almost no oxide film was observed near the crack front in the low-ECP conditions. In addition, a high density of deformation twins and dislocations were found near the fracture surface of the crack front. It is considered that both localized deformation and oxidation are possible dominant factors for the SCC growth in highly irradiated SSs.

Journal Articles

Analytical study of the applicability of FeCrAl-ODS cladding for BWR

Takano, Sho*; Kusagaya, Kazuyuki*; Goto, Daisuke*; Sakamoto, Kan*; Yamashita, Shinichiro

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

We focused on one of accident tolerant fuel (ATF) materials, Oxide Dispersion Strengthened Fe-Cr-Al Steel (FeCrAl-ODS). There is a reasonable prospect that FeCrAl-ODS is applied to BWRs, but relatively high neutron absorption should be compensated. To decrease adverse neutron economic impact, thin FeCrAl-ODS cladding was designed, and we evaluated characteristics of a core into which 9$$times$$9 Advanced BWR (ABWR) bundles with thin FeCrAl-ODS claddings were loaded. Thin FeCrAl-ODS water rods and channel boxes were also applied. We confirmed that FeCrAl-ODS core reactivity was sufficient by increasing enrichment of UO$$_{2}$$ fuel under the limit of 5 wt%. Moreover, some representative FeCrAl-ODS core characteristics were comparable to zircaloy core. We also confirmed that fuel thermal-mechanical behaviors of thin FeCrAl-ODS cladding at normal operation and transient conditions were acceptable. These results led to a conclusion that FeCrAl-ODS was applicable to BWR in the analysis range of this study.

Journal Articles

FEMAXI-7 prediction of the behavior of BWR-type accident tolerant fuel rod with FeCrAl-ODS steel cladding in normal condition

Yamaji, Akifumi*; Yamasaki, Daiki*; Okada, Tomoya*; Sakamoto, Kan*; Yamashita, Shinichiro

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

Features of the accident tolerant fuel performance were evaluated with FEMAXI-7 when the current Zircaloy(Zry) cladding is replaced with FeCrAl-ODS steel cladding (a type of oxide dispersion strengthened steel being developed under the Project on Development of Technical Basis for Safety Improvement at Nuclear Power Plants by Ministry of Economy, Trade and Industry (METI) of Japan) for BWR 9$$times$$9 type fuel rod. In particular, influences of the creep strain rate and thickness of the ODS cladding on the fuel temperature, fission gas release rate (FGR) and pellet-cladding mechanical interaction (PCMI) are investigated.

Journal Articles

Source term analysis considering B$$_{4}$$C/steel interaction and oxidation during severe accidents

Ishikawa, Jun; Shiotsu, Hiroyuki; Sugiyama, Tomoyuki; Maruyama, Yu

Proceedings of 25th International Conference on Nuclear Engineering (ICONE-25) (CD-ROM), 7 Pages, 2017/07

Journal Articles

Irradiation assisted stress corrosion cracking

Pokor, C.*; Herbelin, A.*; Couvant, T.*; Kaji, Yoshiyuki

NEA/NSC/R(2016)5 (Internet), p.317 - 360, 2017/05

In aged BWR plants, certain locations in the mid-plane of the core shroud experience fluence levels at which the materials become susceptible to irradiation assisted stress corrosion cracking (IASCC). BWRVIP (Boiling Water Reactor Vessel Internals Program) has developed crack growth disposition methodologies for evaluating intergranular stress corrosion cracking (IGSCC) in the internal components of BWRs and the Japan Nuclear Energy Safety organization (JNES) has been conducting a project related to IASCC crack growth rate data as a part of safety research and development study for the aging management and maintenance of the nuclear power plants. Although many investigators proposed prediction models for SCC and IASCC growth rates for austenitic stainless steels and Ni alloys, even more improvements of models are necessary as compared with the detailed experimental results, because these models are still preliminary models.

Journal Articles

Failure evaluation analysis of reactor pressure vessel lower head in BWR due to severe accident

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

Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 6 Pages, 2017/04

To investigate the inhomogeneous temperature and stress distribution by geometrical complex of BWR lower head, the detailed 3D model of RPV lower head with control rod guide tubes and shroud supports are constructed and the 3D thermal hydraulic analysis of simulated molten pool and creep deformation analysis of lower head are performed using ANSYS Fluent / Mechanical finite element code. It is found that failure for BWR lower head might be caused by combination between melting failure in inner surface of lower head and creep failure in outer surface of lower head.

Journal Articles

Effectiveness evaluation of filtered containment venting system using THALES-2

Kondo, Masahiro*; Yoshimoto, Tatsuya*; Ishikawa, Jun; Okamoto, Koji*

Hozengaku, 15(4), p.79 - 85, 2017/01

no abstracts in English

JAEA Reports

Results and progress of fundamental research on fission product chemistry; Progress report in 2015

Osaka, Masahiko; Miwa, Shuhei; Nakajima, Kunihisa; Di Lemma, F. G.*; Suzuki, Chikashi; Miyahara, Naoya; Kobata, Masaaki; Okane, Tetsuo; Suzuki, Eriko

JAEA-Review 2016-026, 32 Pages, 2016/12


A fundamental research program on fission product (FP) chemistry has started since 2012 for the purpose of establishment of a FP chemistry database in each region of LWR under severe accident and improvement of FP chemical models based on the database. Research outputs are reflected as fundamental knowledge to both the research and development of decommissioning of Fukushima Daiichi Nuclear Power Station (1F) and enhancement of LWR safety. Four research items have thus been established considering the specific issues of 1F and the priority in the source term research area, as follows: effects of boron (B) release kinetics and thermal-hydraulic conditions on FP behavior, cesium (Cs) chemisorption and reactions with structural materials, enlargement of a thermodynamic and thermophysical properties database for FP compounds and development of experimental and analytical techniques for the reproduction of FP behavior and for direct measurement methods of chemical form of FP compounds. In this report, the research results and progress for the year 2015 are described. The main accomplishment was the installation of a reproductive test facility for FP release and transport behavior. Moreover, basic knowledge about the Cs chemisorption behavior was also obtained. In addition to the four research items, a further research item is being considered for deeper interpretation of FP behavior by the analysis of samples outside of the 1F units.

Journal Articles

Influence of temperature histories during reactor startup periods on microstructural evolution and mechanical properties of austenitic stainless steel irradiated with neutrons

Kasahara, Shigeki; Kitsunai, Yuji*; Chimi, Yasuhiro; Chatani, Kazuhiro*; Koshiishi, Masato*; Nishiyama, Yutaka

Journal of Nuclear Materials, 480, p.386 - 392, 2016/11

 Times Cited Count:0 Percentile:0.01(Materials Science, Multidisciplinary)

This paper addresses influence of two different temperature profiles during startup periods in the Japan Materials Testing Reactor and a boiling water reactor upon microstructural evolution and mechanical properties of austenitic stainless steel irradiated with neutrons to about 1 dpa and 3 dpa. Tensile tests at 290$$^{circ}$$C and Vickers hardness tests at room temperature were carried out, and their microstructures were observed by FEG-TEM. Influence of difference in the temperature profiles was observed obviously in interstitial cluster formation, in particular, growth of Frank loops. The influence was also found certainly in loss of strain hardening capacity and ductility, although the influence on the yield strength and the Vickers hardness was not clearly observed. As a result, Frank loops, which were observed in austenitic stainless steel irradiated at doses of 1 dpa or more, were considered to contribute to deformation of the austenitic stainless steel.

395 (Records 1-20 displayed on this page)