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Onoe, Hironori; Kosaka, Hiroshi*; Matsuoka, Toshiyuki; Komatsu, Tetsuya; Takeuchi, Ryuji; Iwatsuki, Teruki; Yasue, Kenichi
Genshiryoku Bakkuendo Kenkyu (CD-ROM), 26(1), p.3 - 14, 2019/06
In this study, it is focused on topographic changes due to uplift and denudation, also climate perturbations, a method which is able to assess the long-term variability of groundwater flow conditions using the coefficient variation based on some steady-state groundwater flow simulation results was developed. Spatial distribution of long residence time area which is not much influenced due to long-term topographic change and recharge rate change during the past one million years was able to estimate through the case study of the Tono area, Central Japan. By applying this evaluation method, it is possible to identify the local area that has low variability of groundwater flow conditions due to topographic changes and climate perturbations from the regional area quantitatively and spatially.
Kato, Tomoko; Fukaya, Yukiko*; Sugiyama, Takeshi*; Nakai, Kunihiro*; Oda, Chie; Oi, Takao
JAEA-Data/Code 2019-002, 162 Pages, 2019/03
JAEA-Data-Code-2019-002.pdf:2.78MB
The radioactive waste generated from Fukushima Daiichi nuclear power station (FDNPS) accident have features such as wide range of radioactivity level (from low to high) and huge amount etc. It would be necessary for the waste from the FDNPS accident to develop suitable disposal concept and to be disposed safely and reasonably. When considering such appropriate disposal concepts in site-generic phase, it is necessary to appropriately develop models and parameters depending on the disposal concepts, such as disposal depth and specification of engineered barrier. In addition, it is desirable to evaluate the safety of repository with common models and parameters independent on the disposal concepts. In the safety assessment of disposal, it is useful to show the difference in performance of repository with "dose" as an indicator of safety assessment. Biosphere model and parameter set and flux-to-dose conversion factors calculated using them are originally dependent on the disposal concepts. However, the biosphere models and the parameter set in safety assessment of near-surface disposal, sub-surface disposal and geological disposal are prepared in each case, and are different according to the age and purpose of the discussion. In this study, an example of biosphere model and parameter-set of groundwater sceinario commonly applicable to various disposal concepts were shown, to calculate flux-to-dose conversion factors, as common indicators independent to disposal concept. And, a set of flux-to-dose conversion factors was also calculated by using the commonly available biosphere model and parameter set. By applying the flux-to-dose conversion factors, it is possible to compare the performance of disposal concepts to the waste generated from FDNPS accident, focusing on the parts depending on the disposal concepts.
Ogawa, Masuro
Nuclear Engineering and Design, 308, p.133 - 141, 2016/11
Times Cited Count：2 Percentile：19.57(Nuclear Science & Technology)A new basic concept on safety; Not causing any serious catastrophe by any means and a new basic concept on radioactive waste; Not returning any waste that possibly affects the environment are proposed in the present study, aiming at nuclear power plants which everybody can accept, in consideration of the serious catastrophe that happened at Fukushima in 2011. In the present study, physical phenomena are used to continue confining, rather than confine. To continue confining is meant to apply natural correction to fulfill inherent safety function. Fission products must be detoxified to realize the new basic concept on radioactive waste, aiming at the final processing and disposal of radioactive wastes as same as that in the other wastes such as PCB. The New HTGR is proposed based on the new basic concepts. It is indicated that the New HTGR can response to social requirements for safety and environmental conservability against radioactive wastes, industrial requirements for economy, uranium resource sustainability and so on, and national requirements for non-proliferation and environmental protection against carbon dioxide.
Hatanaka, Koichiro; Shibata, Masahiro
Tekisuto "Kakunenryo Saikuru" (Internet), 6 Pages, 2014/06
no abstracts in English
Oigawa, Hiroyuki; Nishihara, Kenji; Minato, Kazuo; Kimura, Takaumi; Arai, Yasuo; Morita, Yasuji; Nakayama, Shinichi; Katakura, Junichi
JAERI-Review 2005-043, 193 Pages, 2005/09
JAERI-Review-2005-043.pdf:16.13MB
JAERI has been conducting research and development on partitioning and transmutation (P&T) technology for long-lived nuclides to develop the double-strata fuel cycle concept, in accordance with the Atomic Energy Commission's "Research and Development of Technologies for Partitioning and Transmutation of Long-lived Nuclides - Status and Evaluation Report" issued in 2000. The double-strata fuel cycle concept consists of four major processes: partitioning, fuel fabrication, transmutation, and fuel processing. The five-year achievement and future perspectives for the technology on these processes are presented in this report. It also provides an analytical study on impacts of introducing P&T technology on waste management, and on deployment of P&T for the future nuclear energy system.
Nishihara, Tetsuo; Sakaki, Akihiro*; Inagaki, Yoshiyuki; Takami, Kazuo*
Nihon Genshiryoku Gakkai Wabun Rombunshi, 3(4), p.381 - 387, 2004/12
JAERI has been carried out research and development on the HTTR hydrogen production system. One of the key components in this system is a high temperature isolation valve (HTIV) installed on the hot helium gas piping penetrating the reactor containment vessel. Angle valve with inner thermal insulator was selected for HTIV and conceptual design was performed. The structural integrity of HTIV was clarified by the stress analyses. Allowable helium leak rate of HTIV was discussed. Helium leak tests using small-scaled valve seat models were performed to decide the seat surface shape and valve closing force. The test results show that the leak rate of wedge shape seat increased in proportion to the number of simulated temperature and stress cycles loaded on the seat models before helium leak test, however that of flat seat did not depend on the number of cycles. So flat seat is adopted for HTIV. It was found that the seat closing force of 30 MPa is reasonable to meet the allowable helium leak rate.
Sakaki, Hironao; Nakamura, Naoki*; Takahashi, Hiroki; Yoshikawa, Hiroshi
JAERI-Tech 2004-022, 28 Pages, 2004/03
In High Intensity Proton Accelerator Project (J-PARC), the peak current 50mA(max.) proton beam is accelerated to 50GeV. Therefore, when the magnet trouble etc. occurs, and the beam collides toward the accelerating structure, the thermal shock destruction is caused on the surface of the material of the structure. This report shows the design policy of "Prototype-unit for the machine protection system" that is necessary to evade the thermal shock destruction.
Kobayashi, Kaoru*; Kaminaga, Masanori; Haga, Katsuhiro; Kinoshita, Hidetaka; Aso, Tomokazu; Hino, Ryutaro
JAERI-Review 2002-010, 52 Pages, 2002/05
JAERI-Review-2002-010.pdf:3.38MB
no abstracts in English
Sasaki, Noriaki
JNC TN8420 2001-006, 56 Pages, 2001/12
Geologic disposal cannot be implemented based only on the consensus of the engaged technical community, and needs the wide social agreement and confidence for it. This is now a common understanding in many countries. Under this kind of recognition, the concept of retrievability in geologic disposal of radioadive wastes has been rapidly interested inrecent years and has being discussed in several European countries. For example, EC has cooperated the concerted action on the retrievability of long-lived radioactive waste with the joining of nine countries, and the expert group on disposal concepts for radioactive waste (EKRA) set up by the Swiss government has presented its findings on the new concept of the long-lived radioactive waste management considering the retrievability. The OECD/NEA has also discussed on this issue and published the documents. There are some countries where the legislation requires the retrievability. This paper briefly summarizes the important findings and recommendations on the concept of retrievability, as the results of review of some interesting documents from European countries, for the purpose of reflecting to the research and development of geologic disposal in Japan.
Sakai, Akihiro; Yoshimori, Michiro; Okoshi, Minoru; Yamamoto, Tadatoshi; Abe, Masayoshi
JAERI-Tech 2001-018, 88 Pages, 2001/03
JAERI-Tech-2001-018.pdf:5.66MB
no abstracts in English
Yabuuchi, Noriaki; Takahashi, Masao*; Nakazawa, Toshio; Sato, Kazuo*; Shimazaki, Junya; Ochiai, Masaaki
JAERI-Research 2000-063, 69 Pages, 2001/02
JAERI-Research-2000-063.pdf:4.41MB
no abstracts in English
Sato, Haruo
JNC TN8410 2001-003, 40 Pages, 2001/01
JNC-TN8410-2001-003.pdf:1.13MB
A program (TDROCK1.FOR) for simulation and analysis of through-diffusion experiments for a single layer of diffusion media was developed. This program was made by Pro-Fortran language, which was suitable for scientific and technical calculations, and relatively easy explicit difference method was adopted for an analysis. In the analysis, solute concentration in the tracer cell as a function of time that we could not treat to date can be input and the decrease in the solute concentration as a function of time by diffusion from the tracer cell to the measurement cell, the solute concentration distribution in the porewater of diffusion media and the solute concentration in the measurement cell as a function of time can be calculated. In addition, solution volume in both cells and diameter and thickness of the diffusion media are also variable as an input condition. This simulation program could well explain measured result by simulating solute concentration in the measurement cell as a function of time for case which apparent and effective diffusion coefficients were already known. Based on this, the availability and applicability of this program to actual analysis and simulation were confirmed. This report describes the theoretical treatment for the through-diffusion experiments for a single layer of diffusion media, analytical model, an example of source program and the manual.
Kusunoki, Tsuyoshi; Odano, Naoteru; Nakajima, Nobuya; Fukuhara, Yoshifumi*; Ochiai, Masaaki
Nihon Genshiryoku Gakkai-Shi, 42(11), p.1195 - 1203, 2000/11
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)no abstracts in English
Tachimori, Shoichi
Nihon Genshiryoku Gakkai-Shi, 42(11), p.1124 - 1129, 2000/11
Times Cited Count：1 Percentile：81.71(Nuclear Science & Technology)no abstracts in English
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
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JNC TN9400 2000-066, 52 Pages, 2000/06
JNC-TN9400-2000-066.pdf:1.82MB
Phase I of feasibility studies on commercialized fast reactor system is being peformed for two years from Japanese Fiscal Year 1999. In this report, results of the study on fluid fuel reactors (especialiy a molten salt fast breeder reactor concept) are described from the viewpoint of technical and economical concerns of the plant system design. ln JFY1999, we have started to investigate the fluid fuel reactors as alternative concepts of sodium cooled FBR systems with MOX fuel, and selected the unique concept of a molten chloride fast, breeder reactor, whose U-Pu fuel cycle can be related to both light water reactors and fast breeder reactors on the basis of present technical data and design experiences. We selected a preliminary composition of molten fuel and conceptual plant design through evaluation of technical and economical issues essential for the molten salt reactors and then compared them with reference design concepts of sodium cooled FBR systems under limited information on the molten chloride fast breeder reactors. The following results were obtained. (1)The molten chloride fast breeder reactors have inherent safety features in the core and plant performances, ad the fluid fuel is quite promising for cost reduction of the fuel fabrication and reprocessing. (2)On the other hand, the inventory of the molten chloride fuel becomes high and thermal conductivity of the coolant is inferior compared to those of sodium cooled FBR systems, then, the size of main components such as lHX's becomes larger and the amount of construction materials is seems to be increased. (3)Furthermore economical vessel and piping materials which contact with the molten chloride salts are required to be developed. From the results, it is concluded that further steps to investigate the molten chloride fast breeder reactor concepts are too early to be conducted.
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JNC TN9400 2000-060, 168 Pages, 2000/05
JNC-TN9400-2000-060.pdf:4.09MB
Optimal vertical isolation characteristics were studied for the structural concept of vertical seismic isolation system, which uses a common deck and a set of large coned dish springs. Four kinds of earthquake wave and three kinds of artificial seismic wave were used. The earthquake response analysis of a base isolated building was carried out considering some ground conditions and some vertical vibration characteristics of the building isolator. Floor response and acceleration time history at the vertical isolation level were arranged. Using the acceleration time history as a seismic input, the earthquake response analysis of the vertical isolation system according to single degree of freedom model was carried out. Linear analysis and non-linear analysis were made. ln the linear analysis, vertical isolation frequency was examined within 0.8 to 2.5 Hz, and damping ratio was examined within 2 to 60%. ln the non-linear analysis, it was examined within vertical isolation frequency 0.5 to 5Hz, which depended only on the rigidity of the coned disk spring, rigidity ratio of the damping devise 1 to 20 and yield seismic intensity of the damping devise 0.01 to 0.2. As the optimal vertical isolation characteristics of the system, the criterion of largest relative displacement, maximum acceleration and maximum value of the floor response acceleration between 5 to 12Hz was set, the combination region of the appropriate parameter were examined. ln case of largest relative displacement 50mm, acceleration response magnification of 0.75, floor response magnification of 0.33 were used as a criterion, from the result of the linear analysis, vertical frequency was set at 0.8 to l.2 Hz, and by combining the damping ratio over 20 %, it was proven that appropriate vertical isolation characteristics were obtained. The result of the non-linear analysis showed that the combination of the coned disk spring of vertical frequency 0.8 to 1.0 Hz and the damping element of rigidity ...
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JNC TN8410 2000-011, 185 Pages, 2000/05
JNC-TN8410-2000-011.pdf:4.67MB
This report describes the neutronic design calculational methods used in Fuel Design and Evaluation Group in order to inform other related sections of FBR core analysis technology and hand down the technology. Especially we show the neutronics calculation procedures used for the conceptual design study of the advanced core with 127 pin bundle for MONJU that has been carried out in our group. The topics include effective cross section preparation calculations, two-dimensional depletion calculations, three-dimensional diffusion calculations, reactivity coefficient calculations, and control rod worth calculations. The calculational methods shown in this report are the standard neutronics calculation methods employed in our group at the moment. However, the improvement of calculation codes, the reduction of correction factors and uncertainties for design using the nuclear data obtained in the start-up test for MONJU and so on, and the update of nuclear data file will be planned in order to improve evaluation accuracies. Those may change the neutronic design calculational methods, but we decided to describe the present standard calculational methods in our group from the viewpoint of sharing information in JNC.
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.
