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Kato, Tomoko; Fujishima, Atsushi; Ueno, Kenichi; ; Notoya, Shin; Sonobe, Hitoshi
JNC TN8450 2001-003, 205 Pages, 2001/01
JNC-TN8450-2001-003.pdf:77.1MB
We have compiled and refined the information materials to explain ENTRY (Engineering Scale Test and Research Facility) and QUALITY (Quantitative Assessment Radionuclide Migration Experimental Facility). These include information materials to show activities for research and development of radioactive waste disposal in Tokai Works such as panels of experimental equipments. This work was carried out by a working group in Waste Isolation Research Division, Waste Management and Fuel Cycle Research Center; Tokai Works in 1998
2000. We have developed database for above information materials including typical experimental equipments of ENTRY and QUALITY. In the future, it can be easily refind in case of reconstruction of the experimental equipments. This report presents the database including the experimental equipments and several pamphlet.
Kaminaga, Masanori; Sasaki, Shinobu; Haga, Katsuhiro; Aso, Tomokazu; Kinoshita, Hidetaka; Kogawa, Hiroyuki; Akimoto, Atsushi*; Adachi, Junichi*; Hino, Ryutaro
JAERI-Tech 2000-060, 37 Pages, 2000/11
JAERI-Tech-2000-060.pdf:5.54MB
no abstracts in English
; Sato, Wakaei*;
JNC TN9400 2000-037, 87 Pages, 2000/03
JNC-TN9400-2000-037.pdf:3.48MB
ln order to compare the nuclear characteristics of water-cooled bleeder cores with that of LMFBR, MOX fuel cell models are established for boiling and non-boiling LWR, non-boiling HWR and sodium-cooled reactor. Frst, the comarison is made between the heterogeneous cell calculation results by SRAC and those by SLAROM. The results show some differences as for neutron energy spectrum, one-grouped cross section and conversion ratio due to the different grouped cross section library (both are based on JENDL-3.2, though) used for each code, however, the difference is acceptably small for grasping the basic characteristics of the above-mentioned cores. Second, using the SLAROM code, main core parameters such as mean neutron energy, ratio of fast neutron and 
-value, are analyzed. The comparison between the cores show that softened neutron spectrum by the scattering effect of hydrogen or heavy hydrogen increase the contribution of nuclear reaction (especially for neutron capture reaction rather than fission reaction) in lower energy region comparing with LMFBR. ln order to overcome the effect, tighter lattice than LMFBR is necessary for water-cooled cores to realize the breeding of fissile nuclides. Third, effects of Pu isotopic composition on the breeding ratio are evaluated using SRAC burnup calculation. From the results, it is confirmed that degraded Pu (larger ratio of Pu-240) show the larger breeding ratio. At last, sensitivity analyses are made for k-effective and main reaction ratios. As for k-effective, using a temporary covariance data of JENDL-3.2, uncertainty resulting from the cross sections' error is analyzed for a boiling LWR and a sodium-cooled reactor. The boiling LWR core shows larger sensitivity in lower energy region than the sodium-cooled reactor (especially for the energy region lower than 1kev), And, 18-group analysis that is considered acceptably good for LMFBR analysis, should not be enough for accurate sensitivity estimation of ...
Sugino, Hiroyuki; Fujita, Tomoo; Taniguchi, Wataru; Iwasa, Kengo; Hasegawa, Hiroshi
JNC TN8400 99-096, 23 Pages, 1999/12
The Japan Nuclear Cycle Development Institute (JNC) has prepared a second progress report (entitled H12) on research and development for geological disposal of high-level waste (HLW) in Japan. H12 report consist of a Project Overview Report and three Supporting Reports which cover the three major fields described in the AEC Guidelines: (1)evaluation of the geological environment, (2)repository design and engineering technology, (3)performance assessment. This report is prepared to explain background information of buffer design which is descried in Supporting Report 2 (Repository Design and Engineering Technology). In buffer design of H12 report, the design requirements of the buffer are assumed and the relationship between buffer thickness and density was shown corresponding design requirement as an area map. This report describes the background information such as the numerical formulations, assumptions, engineering judgement and so on.
; ; Tanai, Kenji
JNC TN8400 99-047, 54 Pages, 1999/11
This paper reports on the design process for a carbon-steel overpack as a key component in the engineered barrier system of a deep geological repository described in the 2nd progress report. The results of the research and development regarding design requirements, configuration, manufacturing and inspection of overpack are also described. The concept of a composite overpack composed of two different materials is also considered. First, the design requirements for an overpack and presume environmental and design conditions for a repository are provided. For a candidate material of carbon steel overpack, forging material is selected considering enough experience of using this material in nuclear power boilers and other components. Second, loading conditions after emplacement in a repository are set and the pressure-resistant thickness of overpack is calculated. The corrosion thickness to achieve an assigned 1000 year life time and the required thickness to prevent radiolysis of ground water which might enhance corrosion rate are also determined. As aresult, the total required thickness of a carbon-steel overpack is conservatively estimated to 190 mm. This is a reduction of about 30% from the previous estimate provided in the 1st Progress Report. Additional items that must be considered in manufacturring and operating overpacks (i.e. sealing of vitrified waste, examination of main body and sealing welding, mechanism of handling) are evaluated on the basis of current technology, specific future data needs are identified. With respect to the concept of composite overpack (i.e., an outer vessel to provide corrosion-allowance or corrosion-resistant performance and an inner vessel to provide pressure-resistance), the differences in design concepts between the carbon-steel overpack and such composite overpacks are analyzed. Future data needs and analytical capabilities with respect to overpacks are also summarized.
Hino, Ryutaro; Haga, Katsuhiro; Kaminaga, Masanori; Aso, Tomokazu; Kogawa, Hiroyuki; Ishikura, Shuichi*; ; Nakamura, Fumihito*; ;
JAERI-Tech 97-035, 194 Pages, 1997/07
no abstracts in English
; ; ; ;
PNC TN9420 92-014, 125 Pages, 1992/11
This report decribes the development of multi-array type probe for FBR steamgenerator tube. We studied integration between three kinds of probes, which were for axial flaw, for circumferential flaw and for wall thickness flaw, detecting method of accuracy locating probe and basic composition of multi-channel detector. It was comfirmed that each devices had object performance in performance test. We shall use this results to study design and manufacture of ultrasonic testing equipment for steamgenerator of FBR Monju.
; ; ; ;
PNC TN9410 92-254, 76 Pages, 1992/07
A verification test of the inspection system of Monju steam generator(SG) tubes will be performed in near future. Mockup Test Apparatus for the inspection system of SG tubes was manufactured and installed at Mechatronics Application Reserch Facility (MARF) in OEC. The test apparatus has the same specification, which is prepared for verification test, as Monju plant; for instance, which are dimension and material of tubes, and workability for the inspection equipment. About one hundred and forty SG tubes are radially arranged in tube sheets in Monju SG, however, three tubes, inner, center and outer one, are sellected in this test apparatus for testing of inspection system, It was verified that the test apparatus was manufactured with the same accuracy and dimension as Monju. System verification test is planned using this test apparatus.
; ; ; Sakurai, Satoshi;
JAERI-M 90-161, 41 Pages, 1990/09
no abstracts in English
Sakurai, Satoshi; Hirata, Masaru; ; Usuda, Shigekazu; Abe, Jiro; ; Tachimori, Shoichi; ; Kurihara, Masayoshi; Kobayashi, Iwao
JAERI-M 90-059, 35 Pages, 1990/03
no abstracts in English
Cr-1Mo Steel As a Pressure Vessel for VHTR; ; ;
JAERI-M 85-170, 36 Pages, 1985/10
This report describes the present status and items on the application of the 2Cr-1Mo steel, normalized and tempered condition (NT) with low silicon and high purity, to the pressure vessel of VHTR, which is the most important structural component within the pressure boundaries. The SCMV 4-2 steel in JIS, which corresponds to ASTM A387 Grade 22 Class 2, is selected as a candidate based on the design and operation conditions of the pressure vessel. In this report are also shown the specification of material fabrication, various results of mechanical and metallurgical tests and the evaluation of structural integrity of the vessel considering the material aging effects in this SCMV 4-2 steel.