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Saito, Shigeru; Wan, T.*; Okubo, Nariaki; Obayashi, Hironari; Watanabe, Nao; Ohdaira, Naoya*; Kinoshita, Hidetaka; Yamaki, Kenichi*; Kita, Satoshi*; Yoshimoto, Hidemitsu*; et al.
JPS Conference Proceedings (Internet), 33, p.011041_1 - 011041_6, 2021/03
An Accelerator Driven System (ADS) for waste transmutation investigated in JAEA employs lead-bismuth eutectic (LBE) as a neutron production target material and coolant. The neutrons are to be produced via the spallation with 1.5 GeV proton beam injection. As materials irradiation data are important for ADS development, JAEA plans to construct an irradiation facility with LBE neutron production target in J-PARC. There are many technical issues on LBE for practical use. In JAEA, various R&Ds are being carried out. Concerning corrosion study, conditioning operation and functional tests of OLLOCHI started. Oxygen concentration control technology has also developing. In the large scale LBE loop experiment, the operation for steady state and transient experiments was performed by using IMMORTAL. In the area of instrument, development of ultrasonic flow meter and freeze seal valve are progressing as a key technology for the LBE loop system. Investigation of behavior of impurities in LBE, which is important for design of the irradiation facility, started. In this paper, the status of the LBE studies and experimental plan will be presented.
Information Systems Operating Division
JAERI-Review 2005-032, 151 Pages, 2005/08
JAERI-Review-2005-032.pdf:11.51MB
Center for Promotion of Computational Science and Engineering (CCSE) of Japan Atomic Energy Research Institute (JAERI) installed large computer systems including super-computers in order to support research and development activities in JAERI. CCSE operates and manages the computer system and network system. This report presents usage records of the JAERI computer system and the big users' research and development activities by using the computer system in FY2004 (April 1, 2004 - March 31, 2005).
Center for Promotion of Computational Science and Engineering
JAERI-Review 2005-008, 199 Pages, 2005/03
JAERI-Review-2005-008.pdf:30.15MB
The Center for Promotion of Computational Science and Engineering (CCSE) in Japan Atomic Energy Research Institute (JAERI) installed large computer system including super-computers in order to support research and development activities in JAERI. CCSE is also operate anad manage the computer system and network system. This report presents the survey results of research and development activities by using JAERI computer system in FY2003.
Sakurai, Kiyoshi; Yamamoto, Toshihiro
Nihon Genshiryoku Gakkai Wabun Rombunshi, 2(2), p.202 - 214, 2003/06
Monte Carlo Working Group of Special Committee on Nuclear Code Evaluation at JAERI investigated present status of application of Monte Carlo Calculation for large nuclear facilities in Japan. Application of Monte Carlo method has already been popular even to large facilities, but the calculated results are not compared with measured datain most of those applications. The authors believe that accuracy and precision of the method should be examined through comparisons with measured data, and that databases of measured data at experimental or commercial facilities should be developed for further comparison with the analysis results.
Kawamata, Yoichi; Yonekawa, Izuru; Kurihara, Kenichi; Sakakibara, Satoru*; Nishimura, Kiyohiko*
Annual Report of National Institute for Fusion Science; April 2002 - March 2003, 59 Pages, 2003/00
no abstracts in English
Myodo, Masato; Okane, Shogo; Miyajima, Kazutoshi
JAERI-Tech 2001-025, 59 Pages, 2001/03
JAERI-Tech-2001-025.pdf:4.73MB
no abstracts in English
Myodo, Masato; Okane, Shogo; Miyajima, Kazutoshi
Dekomisshoningu Giho, (23), p.2 - 16, 2001/03
no abstracts in English
*
JNC TN9440 2000-005, 164 Pages, 2000/06
JNC-TN9440-2000-005.pdf:4.51MB
This report summarizes the operating and irradiation data of the experimental reactor "JOYO" 34th cycle, and estimates the 35th cycle irradiation condition. Irradiation tests in the 34th cycle are as follows: (1)C-type irradiation rig (C4F) (a)High burnup perfomance test of advanced austenitic stainless steel cladding fuel pins (in collaboration with France) (2)C-type irradiation rig (C6D) (a)Large diameter fuel pins irradiation tests (3)Absorber Materials Irradiation Rig (AMIR-6) (a)Run to absorber pin's cladding breach (4)Core Materials Irradiation Rig (CMIR-5) (a)Cladding tube materials irradiation tests for "MONJU" (5)Structure Materials Irradiation Rigs (SMIR) (a)Decision of material design base standard of structure materials for prototype reactor and large reactor (6)Upper core structure irradiation Plug Rig (UPR-1-5) (a)Upper core neutron spectrum effect and accelerated irradiation effect (7)SurVeillance un-instrument Irradiation Rig (SVIR) (a)Confirmation of surveillance irradiation condition for "JOYO" (b)Material irradiation tests (in collaboration with universities) The maximum burnup driver assembly "PFD537" reached 68,500MWd/t(pin average).
Nibe, Nobuaki; Shimakawa, Yoshio; ; Hayafune, Hiroki; ; ;
JNC TN9400 2000-074, 388 Pages, 2000/06
JNC-TN9400-2000-074.pdf:13.32MB
Large sized sodium-cooled fast breeder reactors of large-size are being studied and have been operated in Japan and many countries. ln this feasibility study, evaluation was made on technical feasibinty for design concepts or 1 loop type and 3 pool types, specially from the viewpoint of improvement of economical competence. The design concepts include the ideas of cost reduction measures such as large-scaled components, reduction of loop number and integration of components on the basic of utilization of sodium characteristics. From the results of the evaluation, it may be possible for all the concepts to attain the economical target of 200 thousands yen per kilowatt, though further confirmation should be made for technical feasibility of those concepts. ln addition, the following items were listed up as further cost-reduction measures. (1)Higher temperature cooling system and steam cycle efficiency (2)Shortening of construction term (3)Reduction of safety systems by using measuring instruments with high performmce (4)Adoption of SG-ACS
Kato, Hiroshige*; Sato, Mitsuyoshi*; Owada, Hitoshi*; Mihara, Morihiro;
JNC TN8430 2000-008, 53 Pages, 2000/05
Cementitious materials will be used in TRU waste disposal repository. In such cases, it is considered that the migration of alkaline leachates from cementitious materials, so called high pH plume, will cause dissolution of rock and precipitation of secondary minerals. In addition, the high pH plume will move along the flow of groundwater, so it is predicted that rock formation and components of high pH groundwater vary with time and space. However, time and spatial dependence of the variations of secondary minerals and groundwater components has not been clarified. In order to acquire the data of variations of secondary minerals and groundwater components, we carried out the rock alteration experiments with column method. The crushed granodiorite was filled into 4 meters length column (
3.7 cm) and artificial cement leachate (pH=13.3; Na=0,1 mol/l, K=0.1 mol/l, Ca=0.002 mol/l) was streamed at flow rates of 0.1 ml/min for 7 months at 80
C. As the result, secondary minerals confirmed on the rock were calcite and C-S-H at upstream of column and C-S-H at mid-downstream. The pH value of the fluid dominated by Na and K did not be decreased by reaction with the rock. In this study, the data relating to the effect of high pH plume on rock over the long term was acquired.
JNC TN9400 2000-038, 98 Pages, 2000/04
JNC-TN9400-2000-038.pdf:7.49MB
As an effort in the feasibility study on commercialized Fast Breeder Reactor cycle systems, an evaluation of the measures to prevent the energetic re-criticality in sodium-cooled large MOX core, which is one of the candidates for the commercialized reactor, has been performed. The core disruptive accident analysis of Demonstration FBR showed that the fuel compaction of the molten fuel by radial motion in a large molten core pool had a potential to drive the severe super-prompt re-criticality phenomena in ULOF sequence. ln order to prevent occurrence of the energetic re-criticality, a subassembly with an inner duct and the removal of a part of LAB are suggested based on CMR (Controlled Material Relocation) concept. The objective of this study is the comparison of the effectiveness of CMR among these measures by the analysis using SIMMER-III. The molten fuel in the subassembly with inner duct flows out faster than that from other measures. The subassembly with inner duct will work effectively in preventing energetic re-criticality. Though the molten fuel in the subassembly without a part of LAB flows out a little slower, it is still one of the promising measures. However, the UAB should be also removed from the same pin to prevent the fuel re-entries into the core region due to the pressurization by FCl below the core, unless it disturbs the core performance. The effect of the axial fuel length of the center pin to CMR behavior is small, compared to the effect of the existence of UAB.
Sakai, Takaaki; *; Ohshima, Hiroyuki; Yamaguchi, Akira
JNC TN9400 2000-033, 94 Pages, 2000/04
JNC-TN9400-2000-033.pdf:4.36MB
The feasibility study on several concepts for the commercial fast breeder reactor(FBR) in future has been conducted in JNC for the kinds of possible coolants and fuel types to confirm the direction of the FBR developments in Japan. ln this report, Lead and Lead-Bismuth eutectic coolants were estimated for the decay heat removal characteristics by the comparison with sodium coolant that has excellent features for the heat transfer and heat transport performance. Heavy liquid metal coolants, such as Lead and Lead-Bismuth, have desirable chemical inertness for water and atmosphere. Therefore, there are many economical plant proposals without an intermediate heat transport system that prevents the direct effect on a reactor core by the chemical reaction between water and the liquid metal coolant at the hypocritical tube fairer accidents in a steam generator. ln this study, transient analyses on the thermal-hydraulics have been performed for the decay heat removal events in "Equivalent plant" with the Lead, Lead-Bismuth and Sodium coolant by using Super-COPD code. And a resulted optimized lead cooled plant in feasibility study was also analyzed for the comparison. ln conclusion, it is become clear that the natural circulation performance, that has an important roll in passive safety characteristic of the reactor, is more excellent in heavy liquid metals than sodium coolant during the decay heat removal transients. However, we need to conform the heat transfer reduction by the oxidize film or the corrosion products expected to appear on the heat transfer surface in the Lead and Lead-Bismuth circumstance.
; 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 ...
; Numata, Kazuyuki*
JNC TN9400 2000-036, 138 Pages, 2000/03
JNC-TN9400-2000-036.pdf:10.16MB
Japan Nuclear Cycle Development lnstitute (JNC) had developed the adjusted nuclear cross-section library in which the results of the JUPITER experiments were renected. Using this adjusted library, the distinct improvement of the accuracy in nuclear design of FBR cores had been achieved. As a recent research, JNC develops a database of other integral data in addition to the JUPITER experiments, aiming at further improvement for accuracy and reliability. ln this report, the authors describe the evaluation of the C/E values and the sensitivity analysis for the Experimental Fast Reactor "JOYO" MK-l core. The minimal criticality, sodium void reactivity worth, fuel assembly worth and burn-up coefficient were analyzed. The results of both the minimal criticality and the fuel assembly worth, which were calculated by the standard analytical method for JUPITER experiments, agreed well with the measured values. 0n the other hand, the results of the sodium void reactivity worth have a tendency to overestimate. As for the burn-up coefficient, it was seen that the C/E values had a dispersion among the operation cycles. The authors judged that further investigation for the estimation of the experimental error will increase the applicability of the integral data to the adjusted library. Furthermore, sensitivity analyses for the minimal criticality, sodium void reactivity worth and fuel assembly worth showed the characteristics of "JOYO" MK-l core in comparison with ZPPR-9 core of JUPITER experiments.
Higuchi, Hidekazu; Sato, Motoaki; Hirabayashi, Takakuni*; Tanaka, Mitsugu
Proceedings of 2nd International Conference on Safewaste 2000, Vol.1, p.314 - 322, 2000/00
no abstracts in English
JNC TN9410 2000-003, 52 Pages, 1999/12
JNC-TN9410-2000-003.pdf:3.51MB
In May, 1999, disassembly and cleansing of sodium residues contained in the large cold trap (50MWSG) were carried out. Two cold trap units, one from the primary sodium loop and the other from the for the secondary sodium loop were disassembled and cleaned. This report describes the procedures, methods, and tasks under taken in the clean-up effort, including countermeasures for safe handling of sodium. The disassembly of the cold trap was based an information regarding similar cleansing activities external to JNC. There was also same a priori knowledge of the type and amount of sodium-laden residues. As this result, we conducted disassembly and cleansing task as provisionally planned. In fact we learned that disassembly methods for the specific components could be conducted in an aerated atmosphere. We thus gained additional disassembly and sodium cleansing experience under manageable and safe conditions.
; ; Kamide, Hideki
JNC TN9400 2000-026, 70 Pages, 1999/11
JNC-TN9400-2000-026.pdf:10.93MB
The phenomena on a thermal-hydraulic field inside/outside the porous blockage has been investigated by the experiment employed the 4-subchamlel geometry water test facility. From existing experimental and numerical studies, it was shown that the now field around the blockage had significant effect on the temperature distribution inside the blockage. Therefore, We conducted the experiment of flow visualization with 4-subchannel geometry water test facility. The flow visualization showed that the recirculation flow and a lateral flow existed at the side of the porous blockage. The region of recirculation flow extended to the downstream side with increasing of the flow rate. The mixing due to the convection was dominant in the region of recirculation flow, and was getting effective as increasing of the flow rate. ln the wake region, the flow was unstable due to the interaction between the flows in the unplugged and plugged subchannels. The coolant was provided from the unplugged subchannels and also through the inside of porous blockage. These results obtained from the flow visualization gathering with the numerical simulation will be reflect to the evaluation of the local blockage issue of large scale Fast Breeder Reactor.
Mukaiyama, Takehiko
RIST News, (28), p.3 - 11, 1999/09
no abstracts in English
Research Evaluation Committee
JAERI-Review 99-020, p.77 - 0, 1999/08
JAERI-Review-99-020.pdf:3.99MB
no abstracts in English
Sato, Motoaki
RANDEC Nyusu, (42), p.8 - 9, 1999/07
no abstracts in English
