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Takahashi, Fumiaki; Yamaguchi, Yasuhiro; Saito, Kimiaki; Iwasaki, Midori*; Miyazawa, Chuzo*; Hamada, Tatsuji*
KEK Proceedings 2000-20, p.48 - 55, 2000/12
no abstracts in English
*; Aizawa, Takao*; *
JNC TJ7420 2000-006, 54 Pages, 2000/03
JNC-TJ7420-2000-006.pdf:16.6MB
no abstracts in English
Nishio, Gunji*; *; Kono, Koji*; *; Murazaki, Minoru*
JAERI-Data/Code 98-033, 235 Pages, 1998/11
JAERI-Data-Code-98-033.pdf:9.87MB
no abstracts in English
Nishio, Gunji*; *; Kono, Koji*; *; *; *
JAERI-Data/Code 98-017, 279 Pages, 1998/03
JAERI-Data-Code-98-017.pdf:10.17MB
no abstracts in English
Lund
PNC TN3410 98-001, 14 Pages, 1998/01
At the FUGEN Power Station a system for online monitoring of selected process component behavior, CONFU (CONdition monitoring at FUgen) has been implemented. This system is based on MOCOM (Model Based Condition Monitoring System), developed at IFE/OECD Halden Reactor Project. The system is currently monitoring the heat exchangers for the Reactor Auxiliary Cooling Water System. These heat exchangers has shown a slowly degrading performance over time due to fouling, i.e. accumulation of a heat resisting layer of organic material on the sea water side. This slow degradation, which is not detected by the conventional control and alarm systems, is not an operational, but rather a maintenance problem. CONFU is using dynamically updated mathematical models to compute the performance degradation of the heat exchangers, expressed in overall heat transfer, heat transfer coefficients or heat exchanger efficiency. The results of testing CONFU on real plant data identify the expected degradation trends. The data from CONFU can, in addition to give the plant operator a good impression of the component's operational state, be utilized by the maintenance planning personnel for determination of the most optimal maintenance schedule. Furthermore, the process models in CONFU have been used for simulation purposes.
PNC TN9410 97-039, 187 Pages, 1997/05
A numerical evaluation system, which is consisted of four codes, AQUA, DINUS-3, THEMIS and BEMSET has been developed for thermal striping phenomena. To validate the system for the phenomena, thermally fluid - structure interaction analysis was carried out using a existing sodium experiment of parallel impinging jet simulating the outlet region of an LMFBR core. Calculational results on the RMS values of temperature fluctuation, the histograms of temperature amplitudes and frequencies, the auto-power spectral density distributions of temperature fluctuations and the damping characteristics of temperature fluctuations showed good agreement with the measured values under the test conditions of various flow velocity. From the comparisons with the experimental data, it was concluded that the numerical evaluation system is applicable to the evaluation of thermally fluid - structure interaction phenomena related to the thermal striping.
Jinno, Kenji*; Nakagawa, Kei*; *; ; Ijiri, Yuji*; Watari, Shingo; Webb, E. K.*; Kanazawa, Yasuo*; Uchida, Masahiro
PNC TY1606 97-001, 44 Pages, 1997/03
no abstracts in English
Otaka, Masahiko; Ohshima, Hiroyuki
PNC TN9410 96-118, 26 Pages, 1996/04
We have started to develop a whole core thermmal-hydraulic analysis code ACT(Analysis program of whole Core Thermal-hydraulics) for the purpose of evaluating detailed in-core thermal-hydraulic phenomena under various operation conditions, e,9., the normal operation and the transition from forced to natural circulation, of fast reactors. For the high accurate predictivity of the in-core thermal-hydraulics, key phenomena such as inter-wrapper flow (convection through the gaps between fuel subassemblies) and core-plenum thermal-hydraulic interaction should be accounted for. Therefore, ACT consists of four kinds of programs, i.e., intra-subassembly, inter-subassembly, upper plenum and primary loop (including intermediate heat exchanger) analysis programs, which will be made based on several thermal-hydraulic codes that have been developed at PNC and taken the verification and validation. The latter two programs are inevitable parts to give the proper boundary conditions of the in-core thermal-hydraulic analysis, especially in the natural circulation decay heat removal operation mode. These four programs will be coupled with each other and be calculated simultaneously by using parallel computers. In this report, the code development strategy and inter-wrapper flow analysis program which we developed as the first stage of the code development are presented. This program analyzes sodium single phase flow phenomena in inter-subassembly gap at whole core. The finite differential method is applied and the governing equations for fluid continuity, energy and momentum are solved simultaneously. The basic function of program was confirmed through the interwrapper flow analysis of a core consist of 37 fuel subassemblies. This program will be coupled with inter-subassembly analysis program at next stage.
Takase, Hiroyasu*; PeterGr*; Kathery*; *
PNC TJ7281 96-002, 168 Pages, 1996/03
no abstracts in English
Hamada, Hirotsugu; Tanabe, Hiromi
PNC TN9410 96-027, 41 Pages, 1995/12
If a sodium-water reaction jet was formed due to water leakage in an FBR steam generator(SG), neighboring tubes would suffer from overheating. On the safety aspect of the SGs, it is important to confirm that the neighboring tubes would not fail under such a severe overheating condition. So far, an analytical model using the structural integrity analysis code, FINAS, has been prepared and validated by the explosive torch overheating test data. This report presents the results on the overheating failure analysis of the under-sodium leak in the PFR superheater(SH), 1987. In the SH with slow steam dump system in 1987, neighboring overheated tubes are failed about 3 seconds after the SH isolation, which is shown both by the leak in the PFR and its analysis. For the SH in which a fast steam dump system was installed after the leak of 1987, the analysis shows no tube failure due to the fast steam depression and cooling effect inside. These results indicate that the FINAS model adequately predicts the overheating failure and the specific SH design and operation possibly result in further growth of the leak. It is concluded that steam blow effect is extremely important and the analysis model presented here is useful for the overheating failure evaluation of the SGs.
; ; Adachi, Tetsuya; Sato, Toshinori;
PNC TN7410 95-049, 47 Pages, 1995/10
Excavation of a shaft or a horizontal drift in a rock mass probably affects the rock mass around the underground openings. It is necessary in the design, construction and safety assessment of underground facilities to consider the properties and extent of the EDZ (Excavation Disturbed Zone; the zone where rock properties and rock conditions have been changed due to excavation). In-situ experiment on excavation disturbance has been carried out in the Tono mine and the controlling factors of properties and extent of the EDZ due to blasting has been studied. In order to evaluate dependence of the change of properties and extent of the EDZ on excavation method, Machine Excavation Effects Experiment has been carried out. In FY 1992, a horizontal drift for measurements was excavated. A horizontal drift parallel to the measuring drift is scheduled to be excavated by a machine in FY 1995. The investigations and numerical analysis before excavation of the test drift were carried out in FY 1993 and FY 1994. The objectives of the investigations and numerical analysis carried out in FY 1993 and 1994 are as follows: (1)to measure and evaluate the rock properties and the rock conditions around the test drift before excavation, and (2)to predict the displacements and stress change during excavation of the test drift. The investigations and numerical analysis in FY 1994 consist of the following items: (1)core logging and borehole wall observation, (2)installation of extensometers, and (3)numerical analysis with the Finite Element Method. This report describes the details of the investigations and numerical analysis carried out in FY 1994.
Hamada, Hirotsugu
PNC TN9410 95-262, 35 Pages, 1995/09
Neighboring tubes in an FBR Steam Generator (SG) would suffer from overheating if a sodium-water reaction jet were formed due to water leakage in the SG. On the safety aspect of the SGs, it is important to confirm that the neighboring tubes would not fail under such an overheating condition. An analytical model using the structural integrity analysis code, FINAS, has been prepared to evaluate the overheating failure and here an explosive torch overheating test was analyzed to validate the FINAS model. These experiments and analysis indicate that the overheating failure is closely associated with heat transfer coefficients (HTCs) of outer and inner tube wall and that the FINAS model conservatively predicts the overheating failure within acceptable accuracy. For making progress in further tests like an explosive torch test and its code validation, it would be required that sodium-water reaction experiments should be performed to provide the data on the HTCs, high pressurized and superheated steam should be supplied in the explosive torch test, and that a multidimensional analytical model should be developed to closely predict the temperature distribution in the axial(z-) and circumferential(
-) directions on the tube wall.
*; *; *; *
PNC TJ9124 94-009, 164 Pages, 1994/03
In selecting the reasonable DBL on steam generator, it is indicated that the possibility of failure propagation due to overheating should be evaluated. In this study, the general plan for the next models to evaluate the reasonable DBL have been designed; a)overheating tube bursting models (structural/fractural dynamics), b)unsteady heat conduction analysis models, c)blow down analysis models and d)reaction zone temperature distribution analysis models. Then blow down analysis models were developed to evaluate the overheating tube bursting and analysis code was preliminarily designed in which the module construction of this code and link of each modules were described. Furthermore, easy coupling of this code and LEAP in future was fully considered.