Refine your search:     
Report No.
 - 
Search Results: Records 1-20 displayed on this page of 27

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

Quasi-Monte Carlo sampling method for simulation-based dynamic probabilistic risk assessment of nuclear power plants

Kubo, Kotaro; Jang, S.*; Takata, Takashi*; Yamaguchi, Akira*

Journal of Nuclear Science and Technology, 59(3), p.357 - 367, 2022/03

 Times Cited Count:0

Dynamic probabilistic risk assessment (PRA), which handles epistemic and aleatory uncertainties by coupling the thermal-hydraulics simulation and probabilistic sampling, enables a more realistic and detailed analysis than conventional PRA. However, enormous calculation costs are incurred by these improvements. One solution is to select an appropriate sampling method. In this paper, we applied the Monte Carlo, Latin hypercube, grid-point, and quasi-Monte Carlo sampling methods to the dynamic PRA of a station blackout sequence in a boiling water reactor and compared each method. The result indicated that quasi-Monte Carlo sampling method handles the uncertainties most effectively in the assumed scenario.

Journal Articles

Toward mechanistic evaluation of critical heat flux in nuclear reactors, 2; Recent studies and future challenges toward mechanistic and reliable CHF evaluation

Okawa, Tomio*; Mori, Shoji*; Liu, W.*; Ose, Yasuo*; Yoshida, Hiroyuki; Ono, Ayako

Nihon Genshiryoku Gakkai-Shi ATOMO$$Sigma$$, 63(12), p.820 - 824, 2021/12

The evaluation method of the critical heat flux based on the mechanism is needed for the efficient design and development of fuel in reactors and the appropriate safety evaluation. In this paper, the current researches relating to the mechanism of the critical heat flux are reviewed, and the issue to be considered in the future are discussed.

Journal Articles

Extension of PASCAL4 code for probabilistic fracture mechanics analysis of reactor pressure vessel in boiling water reactor

Lu, K.; Katsuyama, Jinya; Li, Y.

Proceedings of ASME 2020 Pressure Vessels and Piping Conference (PVP 2020) (Internet), 10 Pages, 2020/08

Journal Articles

Boron chemistry during transportation in the high temperature region of a boiling water reactor under severe accident conditions

Miwa, Shuhei; Takase, Gaku; Imoto, Jumpei; Nishioka, Shunichiro; Miyahara, Naoya; Osaka, Masahiko

Journal of Nuclear Science and Technology, 57(3), p.291 - 300, 2020/03

 Times Cited Count:2 Percentile:42.44(Nuclear Science & Technology)

For the evaluation of transport behavior of control material boron in a severe accident of BWR from the viewpoint of chemical effects on cesium and iodine behavior, boron chemistry during transportation in the high temperature region above 400 K was experimentally investigated. The heating tests of boron oxide specimen were conducted using the dedicated experimental apparatus reproducing fission product release and transport in steam atmosphere. Released boron oxide vapor was deposited above 1,000 K by the condensation onto stainless steel. The boron deposits and/or vapors significantly reacted with stainless steel above 1,000 K and formed the stable iron-boron mixed oxide (FeO)$$_{2}$$BO$$_{3}$$. These results indicate that released boron from degraded BWR control blade in a severe accident could remain in the high temperature region such as a Reactor Pressure Vessel. Based on these results, it can be said that the existence of boron deposits in the high temperature region would decrease the amount of transported cesium vapors from a Reactor Pressure Vessel due to possible formation of low volatile cesium borate compounds by the reaction of boron deposits with cesium vapors.

Journal Articles

Report on the stay in Norway as a secondee of OECD Halden Reactor Project

Hata, Kuniki

Hoshasen Kagaku (Internet), (103), P. 65, 2017/04

no abstracts in English

Journal Articles

Critical power prediction for tight lattice rod bundles

Liu, W.; Onuki, Akira; Tamai, Hidesada; Akimoto, Hajime

Proceedings of International Conference on Nuclear Energy System for Future Generation and Global Sustainability (GLOBAL 2005) (CD-ROM), 5 Pages, 2005/10

In this research, the newest version of critical power correlation for tight-lattice rod bundles is proposed by using 7-rod and 37-rod bundle data derived in Japan Atomic Energy Research Institute (JAERI). For comparatively high mass velocity region, the correlation is written in local critical heat flux - critical quality type. For low mass velocity region, it is written in critical quality - annular flow length type. The correlation is verified by JAERI data and Bettis Atomic Power Laboratory data. It is confirmed the correlation is able to give good prediction for the effects of mass velocity, inlet temperature, pressure and heated equivalent diameter on critical power. The correlation is further implemented into TRAC code to analyze flow decrease and power increase transients. It is confirmed transient BT can be predicted within the accuracy of the implemented critical power correlation.

Journal Articles

Critical power correlation for tight-lattice rod bundles

Liu, W.; Kureta, Masatoshi; Onuki, Akira; Akimoto, Hajime

Journal of Nuclear Science and Technology, 42(1), p.40 - 49, 2005/01

 Times Cited Count:7 Percentile:47.08(Nuclear Science & Technology)

In this research, critical power correlation for tight-lattice rod bundles is newly proposed using 7-rod axially uniform-heated data, 7-rod and 37-rod axially double-humped-heated data at Japan Atomic Energy Research Institute (JAERI). For low mass velocity region ($$<$$ 300 kg/m$$^{2}$$s), the correlation is written in critical quality - annular flow length type. For high mass velocity region ($$>$$ 300 kg/m$$^{2}$$s), it is written in local critical heat flux - critical quality type. The standard deviation of ECPR (Experimental Critical Power Ratio) to the whole JAERI data (694 data points) is 6%. The correlation is verified by Bettis Atomic Power Laboratory data (177 points, standard deviation: 7.7%). The correlation is confirmed being able to give good prediction for the effects of mass velocity, inlet temperature, pressure and heated equivalent diameter on critical power. The applicable range of the correlation is: gap between rods from 1.0 to 2.29 mm, heated length from 1.26 to 1.8 m, mass velocity from 150 to 2000 kg/m$$^{2}$$s and pressure from 2 to 11 MPa.

Journal Articles

Critical power in 7-rod tight lattice bundle

Liu, W.; Kureta, Masatoshi; Akimoto, Hajime

JSME International Journal, Series B, 47(2), p.299 - 305, 2004/05

Experimental research on critical power in tight lattice bundle that simulates the Reduced-Moderation Water Reactor (RMWR) has been carried out in Japan Atomic Energy Research Institute (JAERI). The bundle consists one center rod and six peripheral rods. The 7 rods are arranged on a 14.3 mm equilateral triangular pitch. Each rod is 13 mm in outside diameter. An axial 12-step power distribution is employed to simulate the complicate heating condition in RMWR. Experiments are carried out under ${it G}$ = 100-1400 kg/m$$^{2}$$s, ${it P}$$$_{ex}$$ = 2-8.5 MPa. Effects of mass velocity, inlet temperature, pressure, radial peaking factor and axial peaking factor on critical power and critical quality are discussed. Compared with axial uniform heating condition, the axial non-uniform heating condition causes an obvious decrease in critical quality. Arai correlation, which is the only correlation that has been optimized for tight lattice condition, is verified with the present experimental data. The correlation is found to be able to give reasonable prediction only around RMWR nominal operating condition.

JAEA Reports

Behavior of irradiated BWR fuel under reactivity-initiated-accident conditions; Results of tests FK-1, -2 and -3

Sugiyama, Tomoyuki; Nakamura, Takehiko; Kusagaya, Kazuyuki*; Sasajima, Hideo; Nagase, Fumihisa; Fuketa, Toyoshi

JAERI-Research 2003-033, 76 Pages, 2004/01

JAERI-Research-2003-033.pdf:17.46MB

Boiling water reactor (BWR) fuels with burnups of 41 to 45 GWd/tU were pulse-irradiated in the Nuclear Safety Research Reactor (NSRR) to investigate fuel behavior under cold startup reactivity-initiated-accident (RIA) conditions. BWR fuel segment rods of 8$$times$$8BJ (STEP I) type from Fukushima-Daiichi Unit 3 nuclear power plant were refabricated into short test rods, and they were subjected to prompt enthalpy insertion from 293 to 607 J/g (70 to 145 cal/g) within about 20 ms. The fuel cladding had enough ductility against the prompt deformation due to pellet cladding mechanical interaction. The plastic hoop strain reached 1.5% at the peak location. The cladding surface temperature locally reached about 600 deg C. Recovery of irradiation defects in the cladding due to high temperature during the pulse irradiation was indicated via X-ray diffractometry. Fission gas release during the pulse irradiation was from 3.1% to 8.2%, depending on the peak fuel enthalpy and the normal operation conditions.

JAEA Reports

Study on high-performance fuel cladding materials; Joint research report in FY 1999-2000 (Phase 1) (Joint research)

Kiuchi, Kiyoshi; Ioka, Ikuo; Tachibana, Katsumi; Suzuki, Tomio; Fukaya, Kiyoshi*; Inohara, Yasuto*; Kambara, Shozo; Kuroda, Yuji*; Miyamoto, Satoshi*; Ogura, Kazutomo*

JAERI-Research 2002-008, 63 Pages, 2002/03

JAERI-Research-2002-008.pdf:7.85MB

no abstracts in English

JAEA Reports

Development of a BWR core burn-up calculation code COREBN-BWR

*; Okumura, Keisuke

JAERI-M 92-068, 107 Pages, 1992/05

JAERI-M-92-068.pdf:2.79MB

no abstracts in English

JAEA Reports

JAEA Reports

JOINT RESEARCH ON FUEL IRRADIATION IN HALDEN BOILING WATER REACTOR

*; ; *; *; *; *; ; *; *; *

JAERI-M 84-031, 285 Pages, 1984/02

JAERI-M-84-031.pdf:7.12MB

no abstracts in English

JAEA Reports

A Study on Quench Phenomena During Reflood Phase, 1; Quench Models for Bottom Flooding

Murao, Yoshio;

JAERI-M 6984, 39 Pages, 1977/03

JAERI-M-6984.pdf:1.09MB

no abstracts in English

JAEA Reports

Data Report on Series,3; Reflood Experiment

; Murao, Yoshio; ; Sudo, Yukio; ; ; ;

JAERI-M 6983, 298 Pages, 1977/03

JAERI-M-6983.pdf:5.99MB

no abstracts in English

JAEA Reports

Report on Series 4 Reflood Experiment

Murao, Yoshio; ; ; Sudo, Yukio; ; ; ;

JAERI-M 6982, 44 Pages, 1977/03

JAERI-M-6982.pdf:1.1MB

no abstracts in English

JAEA Reports

Report on Series 3 Reflood Experiment

Murao, Yoshio; ; ; Sudo, Yukio; ; ; ;

JAERI-M 6981, 71 Pages, 1977/03

JAERI-M-6981.pdf:1.98MB

no abstracts in English

Oral presentation

Development of plasma heating technology for simulation of LWR severe accident behavior, 2; Outcomes of JFY2015

Abe, Yuta; Sato, Ikken; Nakagiri, Toshio; Nagae, Yuji; Ishimi, Akihiro

no journal, , 

no abstracts in English

27 (Records 1-20 displayed on this page)