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Yokobayashi, Masao; Oikawa, Tetsukuni; Muramatsu, Ken
Nihon Genshiryoku Gakkai Wabun Rombunshi, 1(1), p.95 - 105, 2002/03
no abstracts in English
Yokobayashi, Masao; Kondo, Masaaki*
JAERI-Tech 2001-007, 90 Pages, 2001/03
JAERI-Tech-2001-007.pdf:4.02MB
no abstracts in English
*; *; *; *
JNC TJ8400 2000-005, 71 Pages, 2000/05
In this research, simulations with some parameters which characterize ground water flow and the reliability evaluation for the expansion of the calculation method of groundwater flow were carried out by using the radionuclide transport computations in nearfield heterogeneous porous media. Concretely contents are follows: (1)With the series of calculation method for three-dimensional saturated/unsaturated groundwater flow and one-dimensional radionuclide transport, the computational analyses with the parameters used in JNC report in 2000 was carried out and the influence of the different input flux was evaluated. (2)The examination of the application for the different ways of inverse laplace transformation which is used in one-dimensional radionuclide transport analysis code "MATRICS" was carried out. (3)The examination of the application of multi-element "MATRICS" (m-MATRICS) for radionuclide transport computations in nearfield heterogeneous porous media was carried out. (4)The series of calculation methods from three-dimensional saturated/unsaturated ground water flow simulation code to one-dimensional radionuclide transport simulation code was integrated.
Yokobayashi, Masao; Kinoshita, Naoki*; Tamura, Kazuo*
JAERI-Data/Code 2000-015, p.102 - 0, 2000/03
JAERI-Data-Code-2000-015.pdf:13.92MB
no abstracts in English
Yokobayashi, Masao; *
JAERI-Data/Code 95-013, 99 Pages, 1995/10
JAERI-Data-Code-95-013.pdf:3.04MB
no abstracts in English
Yamato, Aiji; Sasaki, Noriaki; ; Miyahara, Kaname;
PNC TN1100 94-003, 355 Pages, 1993/11
Radioactive waste management research programs inevitably include laboratory solubili and sorption studies to provide data for radionuclide transport model. Estimation of lubility strongly depends on the reliability of thermodynamic data (e.g., carbonato-colexes) and may also depend on kinetic data on alteration of solubility limiting solid ases. Existing sorption data may include some kind of retardation mechanisms to be excded (e.g., precipitation). To develop these fundamental data, we must also consider a rge number of radioactive elements, a large number of factors (e.g., pH, Eh, complexinligands) in the repository environmentg, and numerous solid and aqueous species of radnuclides along with the many absorbents. Therefore, a systematic approach and researchlan are needed for obtaining and evaluation thermodynamic and sorption constants. The cus of this session was on thermodynamic data for aqueous species and solid phases imptant to the geological disposal system, on kinetic data
Watanabe, Norio; *; *; *
Dai-6-Kai Kakuritsuronteki Anzen Hyoka (PSA) Ni Kansuru Kokunai Shimpojiumu Rombunshu (IAE-9206), p.159 - 164, 1993/01
no abstracts in English
Kondo, Masaaki
Dai-6-Kai Kakuritsuronteki Anzen Hyoka (PSA) Ni Kansuru Kokunai Shimpojiumu Rombunshu (IAE-9206), p.53 - 59, 1993/01
no abstracts in English
PNC TN9410 91-286, 117 Pages, 1991/08
A conventional type of RSS in a large scale FBR was designed and its unavailability was analyzed with fault-tree. Reliability of logic circuits of the reaetor protection system is relatively high when compared to that of the control rod insertion. Contributing factors to the unavailabity are multiple failures of detection systems, and failure to insert rods such as failure to deratch or rod jamming. Then the new concept of control rod release mechanism was introduced in the RSS design. The thermal-hydraulic characteristics of the mechanism was analyzed using computer codes SSC-L and AQUA. Further, qualitative analysis of the common cause failure for the RSS was tried with the generic cause approach. The reactor protection systems of the backup RSS are diversified by the self actuated control rod release mechanism. With such a mechanism, the number of common cause factors were decreased for postulated LOF event.
*; *; *; *; Nakanishi, Seiji; *
PNC TN9410 88-062, 82 Pages, 1988/06
In a safety design of Fast Breeder Reactor (FBR), high reliability for the decay heat removal systems (DHRSs) is required as well as the reactor shutdown systems in order to define core disruptive accident as the Beyond Design Basis Accidents (BDBAs). Therefore, it had been desired to establish the quantitative reliability evaluation techniques for DHRSs. In the Key Technological Design Study(II), it was represented that a usual fault-tree analysis was insufficient as a reliability evaluation technique for DHRSs and the development of a dynamic analysis tool considering the state transition of the reference plant was required. On the basis of suggestions, the dynamic reliability analysis code DRAC, to which a Markov model is applied, has been developed. In this code, following two time- dependent effects in reliability evaluation are considered. (1)Transition of required removal heat according to the time-dependency of decay heat. (using the Phased Mission method) (2)Transition of plant coolability considering the failure and repair of plant components. (using the Markov model) To verify the analytical capability of this code, the reliability analysis of a typical loop type plant with four IRACSs designed in the Key Technological Design Study(II) was performed as an example analysis for the practical FBR plant. The results of the analysis showed that the reliability of the plant can be easily estimated in cases of all different systems on the design concepts and the dependence of the equipments, and so the applicability of this code for the practical plant was confirmed. After this, we will utilize this code for the reliability analysis in the practical FBR plants design study through both the estimation of the decay heat removal capability by the plant dynamic analysis codes and the application of the failure and repair rate data of the components in the CREDO data-base, etc.
Iguchi, Yukihiro
PNC TN3410 88-007, 100 Pages, 1988/04
In "Fugen", we started the project which evaluates the importance of the components of the plant in 1985, in order to improve the reliability of the plant effectively. The data base of the evaluation are mostly based on the disclosed databases of other reports and partly on the Maintenance Management System (MMS) data base. As a method of the evaluation, Fault Tree Analysis (FTA) is adopted. In 1987, we complete the fault tree of the plant shut-down, and coded all the data as FT data base. Moreover, we collected new data for future use such as maintenance items use and coded them also. In order to analyze the FT data base, we introduce SETS and FTD codes which are used in PNC often. And we developed a transforming program to apply FT database to SEST. This program can manipulate the trains of the components and their logic combination (e.g. 1 out of 2 twice) easily. And we also modified FTD so that it can plot out Japanese characters, because the event names of FT data base are written in Japanese. In future, these program will be used and improved as ATR Maintenance Instruction System (AMIS).
Suzuki, Kazuhiko*; *; *; Watanabe, Norio; ;
Shisutemu To Seigyo, 30(2), p.109 - 119, 1986/00
no abstracts in English
*; *; Watanabe, Norio; ;
World Congress III of Chemical Engineering, 2, p.1100 - 1103, 1986/00
no abstracts in English
Watanabe, Norio; ; ; *; *
Proc.of ANS/ENS Int.Topical Meeting on Probabilistic Safety Methods and Applications, p.152 - 1, 1985/00
no abstracts in English
Kubo, Kotaro*; Takito, Kiyotaka; Choi, B.; Nishida, Akemi; Muramatsu, Ken; Takada, Tsuyoshi
no journal, ,
no abstracts in English
