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Journal Articles

Introduction of application examples of ultrasonic simulation in the development of nuclear reactor measurement technology

Abe, Yuta; Otaka, Masahiko; Sekiya, Naoki*; Makuuchi, Etsuyo*

Hihakai Kensa, 71(2), p.69 - 74, 2022/02

no abstracts in English

JAEA Reports

Handling of HTTR second driver fuel elements in assembling and storage working

Tomimoto, Hiroshi; Kato, Yasushi; Owada, Hiroyuki; Sato, Nao; Shimazaki, Yosuke; Kozawa, Takayuki; Shinohara, Masanori; Hamamoto, Shimpei; Tochio, Daisuke; Nojiri, Naoki; et al.

JAEA-Technology 2009-025, 29 Pages, 2009/06

JAEA-Technology-2009-025.pdf:21.78MB

The first driver fuel of the HTTR (High Temperature Engineering test Reactor) was loaded in 1998 and the HTTR reached first criticality state in the same year. The HTTR has been operated using the first driver fuel for a decade. In Fuel elements assembling, 4770 of fuel rods which consist of 12 kinds of enrichment uranium are loaded into 150 fuel graphite blocks for HTTR second driver fuel elements. Measures of prevention of fuel rod miss loading, are employed in fuel design. Additionally, precaution of fuel handling on assembling are considered. Reception of fuel rods, assembling of fuel elements and storage of second driver fuels in the fresh fuel storage rack in the HTTR were started since June, 2008. Assembling, storage and pre-service inspection were divided into three parts. The second driver fuel assembling was completed in September, 2008. This report describes concerns of fuel handling on assembling and storage work for the HTTR fuel elements.

JAEA Reports

Progress report for large leak sodium-water reaction study No.1; Large leak sodium-water reaction test report (No.7)

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

PNC TN941 78-32, 84 Pages, 1978/01

PNC-TN941-78-32.pdf:2.35MB

The study to establish the safe design of LMFBR MONJU's steam generator system against the large leak sodium-water reaction is conducing by Power Reactor & Nuclear Fuel Development Corporation. This report includes seven topics which have been presented to the meetings of Atomic Energy Society of Japan, etc. during 1977. Summaries of these topics are as follow; (1)A computer program SWAC-11 was developed to predict the water leak rate from the ruptured heat transfer tube in the steam generator. The basic equations used and method of numerical calculation were explained. The sensitivity survey calculations for various parameters used in the code and the demonstration calculation of the case of MONJU's evaporator were reported. (2)A computer program SWAC-13 was developed to ptedict the pressure and flow behavior in the secondary cooling system, i.e. the quasistatic pressure build up, sodium/hydrogen gas flow in the SG vessels and secondary sodium circuit, and flow characteristics of pressure relief line and the hydrogen gas reliese into the atmosphere. The modeling, basic equations and the method of numerical calculation of the program were reported, and the applicability to SWAT-3 tests was demonstrated. (3)The reaction vessel drain line of SWAT-3 facility was chocked with sodium-water reaction products after water injection test Run-3. In order to understand the cause of chocking, the investigations of the reaction product and temperature distributions in the drain line, and the chemical analysis and the freezing temperature measurement of the reaction products were performed. Those results were summarized quantitatively in this chapter. (4)In order to make clear the flow characteristic of the pressure relief line, we tried to rearrange the data of two test runs of SWAT-1 rig which were obtained by the various kind of the sensors attached on the relief line. It became clear from those investigations that the sodium/hydrogen-gas two phase flow in the ...

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