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Journal Articles

Fission gas release from irradiated mixed-oxide fuel pellet during simulated reactivity-initiated accident conditions; Results of BZ-3 and BZ-4 tests

Kakiuchi, Kazuo; Udagawa, Yutaka; Amaya, Masaki

Annals of Nuclear Energy, 155, p.108171_1 - 108171_11, 2021/06

 Times Cited Count:2 Percentile:23.22(Nuclear Science & Technology)

Journal Articles

Criticality characteristics of MCCI products possibly produced in reactors of Fukushima Daiichi Nuclear Power Station

Tonoike, Kotaro; Okubo, Kiyoshi; Takada, Tomoyuki*

Proceedings of International Conference on Nuclear Criticality Safety (ICNC 2015) (DVD-ROM), p.292 - 300, 2015/09

The damaged Unit 1-3 reactors of the Fukushima Daiichi Nuclear Power Station may contain fuel debris of a significant amount that is in a form of molten-core-concrete-interaction (MCCI) product with porous structure. Such low density MCCI product including fissile material is a great concern for its criticality control, especially under submerged condition, due to its fairly good neutron moderation. This report shows computation results of basic criticality characteristics of the MCCI product, which will facilitate criticality risk assessments during decommissioning of the reactors. The results imply that water bound in concrete may raise the risk from the viewpoints of possibility of criticality events and of effectiveness of mitigation measures such as neutron poison injection into coolant water.

JAEA Reports

Report on the fuel treatment facility operation

Kokusen, Junya; Seki, Masakazu; Abe, Masayuki; Nakazaki, Masato; Kida, Takashi; Umeda, Miki; Kihara, Takehiro; Sugikawa, Susumu

JAERI-Tech 2005-004, 53 Pages, 2005/03

JAERI-Tech-2005-004.pdf:5.92MB

This report presents operating records of dissolution of uranium dioxide and concentration of uranyl nitrate solution and acid removal, which have been performed from 1994 through 2003, for the purpose of feeding 10% and 6% enriched uranyl nitrate solution fuel to Static Experimental Critical Facility(STACY) and Transient Experimental Critical Facility(TRACY) in Nuclear Fuel Safety Engineering Facility(NUCEF).

JAEA Reports

Mechanical properties changes of high burnup PWR fuel cladding by temperature transient

Nagase, Fumihisa; Uetsuka, Hiroshi

JAERI-Research 2002-023, 23 Pages, 2002/11

JAERI-Research-2002-023.pdf:1.94MB

To obtain basic data to evaluate fuel rod integrity during abnormal transient and accident of LWRs, high burnup PWR fuel claddings were heated for 0 to 600s at temperatures of 673 through 1173K, and the mechanical property changes were examined by using ring tensile test at room temperature. As a result of the test, it was shown that strength and ductility of the cladding are changed depending on heating temperature and time. The mechanical property changes by temperature transients are considered to be correspondent mainly to recovery of irradiation defect, recovery and recrystallization of the Zircaloy, phase transformations, and associated change of the hydride distribution and morphology. Comparison with unirradiated claddings suggested that irradiation effects are not completely annealed out by the short-term annealing at high temepratures. Radial change of hydrogen concentration was measured for the high burnup PWR fuel cladding and very high hydrogen concentration of about 2400wtppm was detected at the cladding periphery.

JAEA Reports

OECD/NEA burnup credit criticality benchmarks phase IIIB; Burnup calculations of BWR fuel assemblies for storage and transport

Okuno, Hiroshi; Naito, Yoshitaka*; Suyama, Kenya

JAERI-Research 2002-001, 181 Pages, 2002/02

JAERI-Research-2002-001.pdf:10.89MB

The report describes the final results of the Phase IIIB Benchmark conducted by the Expert Group on Burnup Credit Criticality Safety under the auspices of the OECD/NEA. The Benchmark was intended to compare the predictability of current computer code and data library combinations for the atomic number densities of an irradiated BWR fuel assembly model, which was irradiated under specific power of 25.6 MW/tHM up to 40 GWd/tHM and cooled for five years. The void fraction was assumed to be uniform and constant, at 0, 40 and 70%, during burnup. In total, 16 results were submitted from 13 institutes of 7 countries. The calculated densities of 12 actinides and 20 fission product nuclides were found mostly within a range of +- 10% relative to the average, although some results, esp. 155Eu and gadolinium isotopes, exceeded the band. Pin-wise burnup results agreed well among the participants. The results in the multiplication factor also accorded well with each other for void fractions of 0 and 40%; however some results deviated from the average noticeably for the void fraction of 70%.

Journal Articles

Criticality safety studies related to advisory material for the IAEA regulations

;

Proc. of PATRAM'98, 1, p.217 - 223, 1998/00

no abstracts in English

Journal Articles

Computer code system DSOCEAN for assessing the collective dose of Japanese due to radionuclides released to the ocean from a reprocessing plant

Togawa, Orihiko

Journal of Nuclear Science and Technology, 33(10), p.792 - 803, 1996/10

 Times Cited Count:3 Percentile:32.75(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Irradiation test of medium enriched uranium fuel elements in Japan Materials Testing Reactor; Confirmation of integrity of fuel elements by measurement of fission products in water

Yamamoto, Katsumune; ; ; Yokouchi, Iichiro; ;

Nihon Genshiryoku Gakkai-Shi, 28(5), p.425 - 427, 1986/00

 Times Cited Count:0 Percentile:0.00(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Diffusion of fission products in the matrix graphite for VHTR fuel compacts

; ; Ikawa, Katsuichi

Journal of Nuclear Science and Technology, 21(2), p.126 - 132, 1984/00

 Times Cited Count:9 Percentile:67.73(Nuclear Science & Technology)

no abstracts in English

Oral presentation

Correlation with cobalt-60 for radioactive nuclides in solid waste generated at Fukushima Daiichi NPS

Takahatake, Yoko; Koma, Yoshikazu

no journal, , 

The radioactive nuclides dispersed without control to on- and off-site of Fukushima Daiichi Nuclear Power Station (F1NPS) at its accident in 2011. A large amount of radioactive solid waste has been generated during its decommissioning. Radioactivity inventory of the solid waste should be estimated for R&Ds on waste management. Cobalt-60 is often selected as a key-nuclide for determining activity in solid waste generated at nuclear power plant, and is also detected in various F1NPS waste. In this study, radiochemical analysis data opened to public was investigated for correlation between Co-60 and nuclides of activation and fission products as well as actinides (H-3, C-14, Ni-63, Sr-90, Tc-99, I-129, Cs-137, Eu-154, Am-241, Pu-239+240). Cobalt-60 well correlated with Ni-63, Eu-154, Am-241 and Pu-239+240 in spite of differences in physical/chemical properties and process of contamination. This finding suggests a possibility to apply Co-60 as a key-nuclide of scaling factor method.

Oral presentation

Modification of STACY for study of criticality characteristics of fuel debris, 10; Criticality analysis of experimental core using debris structural material rods and investigation of acceptable core configuration

Yoshikawa, Tomoki; Araki, Shohei; Arakaki, Yu; Izawa, Kazuhiko; Gunji, Satoshi; Suyama, Kenya

no journal, , 

In order to clarify the criticality characteristics of fuel debris, we plan experiments in the Static Experiment Critical Facility (STACY) by using concrete and steel simulating the structural materials of the core of the Fukushima Daiichi Nuclear Power Plant. In order to carry out the experiments, it is necessary to obtain a construction permit for the core with the debris structural material simulant mentioned above. In this presentation, we present the analysis results and the feasible core configuration for obtaining the certification from the regulatory body.

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