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Sakai, Takaaki; ; *; Oyama, Kazuhiro*
JNC TN9400 2001-052, 71 Pages, 2001/05
JNC-TN9400-2001-052.pdf:3.24MB
The feasibility study on future commercial fast breeder reactors in Japan has been conducted at JNC, in which various plant design options with a11 the possible coolant and fuel types are investigated to determine the conditions for the future detailed study. Lead-bismuth eutectic coolant has been selected as one of the possible coolant options. During the phase-I activity of the feasibility study in FY1999 and FY2000, several plant concepts, which were cooled by the heavy liquid metal coolant, were examined to evaluate the feasibility mainly with respect to economical competitiveness with other coolant reactors. A medium-scale (300
550MWe) plant, cooled by a lead-bismuth natural circulation flow in a pool type vessel, was selected as the most possible plant concept for the heavy liquid metal coolant. Thus, a conceptual design study for a lead-bismuth-cooled, natural-circulation reactor of 400MWe has been performed at JNC to identify remaining difficulties in technological aspect and its construction cost evaluation. In this report, a thermal-hydraulic analysis method for lead-bismuth-cooled, natural-circulation reactors is described. A Multi-dimensional steam Generator analysis code (MSG) was applied to evaluate the natural circulation plant by combination with a flow-network-type, plant dynamics code (Super-COPD). By using this combined multi-dimensional plant dynamics code, decay heat removals, ULOHS and UTOP accidents were evaluated for the 100MWe STAR-LM concept designed by ANL. In addition, decay heat removal by the Primary Reactor Auxilialy Cooling System (PRACS) in the 400MWe lead-bismuth-cooled, natural-circulation reactor, being studied at JNC, was analyzed. In conclusion, it becomes clear that the combined multi-dimensional plant dynamics code is suitably applicable to analyses of lead-bismuth-cooled, natural-circulation reactors to evaluate ...
Hazama, Taira; Mori, Tomoaki; ; Aihara, Nagafumi; ; Yoshida, Mamoru; *
JNC TN9400 2001-044, 136 Pages, 2001/05
JNC-TN9400-2001-044.pdf:3.97MB
A Subcriticality measurement technique was developed to improve safety and efficiency of criticality safety control in nuclear fuel processing facilities. In the development, two measurement techniques based on reactor noise analysis were selected as candidates of subcriticality measurement technique applicable to severe situations in FBR fuel reprocessing plants. The research activity was performed in Deuterium Critical Assembly (DCA) which was partly reconstructed from the original core of the advanced thermal reactor, so that light water and FBR type fuel could be used as in the FBR fuel reprocessing plants. Through the research, each technique was improved to satisfy criteria for subcriticality monitoring technique in FBR fuel reprocessing plant. Since the two techniques have basically different features while using common devices,thier combination would be a simple and reliable measurement system. This report summarizes processes and results of the research activity in DCA.
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JNC TN9400 2001-057, 36 Pages, 2001/03
JNC-TN9400-2001-057.pdf:1.15MB
A neutron flux monitor can be effective for early detection of reactivity anomalies in a FBR core. For instance, when a local blockage causes a bubble, there is a slight reactivity change in the high frequency band. It is important to understand the fluctuation characteristics of neutron flux monitors. Analysis of the JOYO reactor showed that flux-monitoring fluctuations are pink noise. These noise characteristics drive the requirements for the JOYO reactor anomaly detection system. The system requires fast time resolution and analog to digital resolution with sufficient dynamic range. A real time arithmetic unit is needed to compare the flux monitor signal with signals from normal conditions. This report describes the characteristics of neutron flux monitor signals and the anomaly detection system designed to analyze them.
Kakehi, Isao; Yoshiuji, Takahiro*
JNC TN9400 2001-054, 144 Pages, 2001/03
JNC-TN9400-2001-054.pdf:25.03MB
This report describes accomplishment of simulations of maintenance methods for the pyrochemical process by using virtual engineering models. The pyrochemical process using molten salt electrorefining would introduce new technologies for new fuels of particle oxide, particle nitride and metallic fuels. This system is a batch treatment system of reprocessing and re-fabrication, which transports products of solid form from a process to next process. The maintenance of this process needs remote control, because the process equipments are arranged in a large hot cell. In this study, a simulation code system has been prepared, which provides virtual engineering environment to evaluate the maintenance methods, which we have planned for the pyrochemical process. The simulation study has been conducted to evaluate the required system functions of the maintenance, including methods procedure, handling system, equipments, and remote control. The simulation study has been conducted in the case of the maintenance of molten salt electrorefiner. As a result of simulation of the process maintenance, which we have designed, the realistic maintenance system for the pyrochemical process have been shown. The issues for the system development have been pointed out.
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JNC TN9400 2001-051, 38 Pages, 2001/03
JNC-TN9400-2001-051.pdf:3.46MB
The study on the passive safety test by using the Experimental Fast Reactor Joyo was performed to demonstrate the inherent safety of fast breeder reactors. An analysis code: Mimir-N2, which has been developed to analyze Joyo plant kinetics, was selected as a standard code for this study. In order to increase the reliability of the calculation, Mimir-N2 code was adjusted based on the data obtained through several plant characteristics tests carried out in Joyo. Throughout an operational data obtained in Joyo, it is supposed that the burn-up dependency observed on the power reactivity coefficient might be coming fiom the reactivity shift caused by a depression of a thermal expansion of fuel pellet. Based on the relationship between the measured power reactivity coefficient and the core averaged burn-up, the burn-up dependency mentioned above was estimated and introduced to Mimir-N2. As a result, calculated core and plant dynamics during the step reactivity response test, such as the response of the power range neutron monitor and the coolant temperature at the core inlet / outlet, corresponded with the measured value, Especially, it was confirmed that Mimir-N2 can simulate the perturbation caused by the thermal expansion of the core support plate. In addition, Mimir-N2 was modified to be enable to take into account for the core bowing reactivity, which is calculated by the core bowing reactivity analysis system developed for Joyo. The preliminary analysis of the plant dynamics during the ATWS events in MK-III core were carried out by using modified Mimir-N2. As a result, it was confirmed that the core bowing reactivity should not be neglected because it sometimes shows positive feedback characteristics.
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JNC TJ9520 2001-001, 81 Pages, 2001/03
Reactor Physics Gr., System Engineering Technology Division, O-arai Engineering Center has consolidated the nuclear design database to improve analytical methods and prediction accuracy for large fast breeder cores such as demonstration or commercial FBRs from the previous research. The up-to-date information about usage of the nuclear characteristics analysis code-set was compiled as a part of the improvement of basic design data base for FBR core. The outlines of the electronic manual are as follows: (1)The electronic manual includes explanations of following codes: (JOINT : Code Interface Program) (SLAROM, CASUP : Effective Cross Section Calculation Code) (CITAT10N-FBR : Diffusion Analysis Code) (PERKY : Perturbative Diffusion Analysis Code) (SNPERT, SNPERT-3D : Perturbative Transport Analysis Code) (SAGEP, SAGEP-3D : Sensitivity Coefficient Calculation Code) (NSHEX : Transport Analysis Code using Nodal Method) (ABLE : Cross Section Adjustment Calculation Code) (ACCEPT : Predicting Accuracy Evaluation Code) (2)The electronic manual is described using HTML file format and PDF file for easy maintenance, updating and for easy referring through JNC Intranet. User can refer manual pages by usual Web browser software without any special setup. (3)Many of manual pages include link-tags to jump to related pages. String search is available in both HTML and PDF documents. (4)User can download source code, sample input data and shell script files to carry out each analysis from download page of each code (JNC inside only). (5)Usage of the electronic manual and maintenance/updating process are described in this report and it makes possible to enroll new code or new information in the electronic manual. Since the information has been taken into account about modifications and error fixings, added to each code after the last consolidation in 1994, the electronic manual would cover most recent status of the nuclear characteristics analysis code-set. One of other ...