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Okabe, Ayao; Onuki, Koji; Kikuchi, H.; M.Uchihashi; Nishibayashi, Yohei; Ikeda, Makinori; Miyake, Osamu
JNC TN2400 2003-005, 62 Pages, 2004/03
JNC-TN2400-2003-005.pdf:2.41MB
Sodium combustion analyses were performed using ASSCOPS version 2.1 in order to obtain background data for evaluating validity of the mitigation system against secondary sodium leak of MONJU. The analytical results of floor temperature and hydrogen concentration were summarized in this report.In the sodium combustion analyses under the detailed design conditions, it was confirmed that the temperature rise of the floor liner was reduced In addition, as for the hydrogen concentration in sodium leak process which is formed by the reaction of sodium and moisture, it was confirmed that it is restricted under 4% of the hydrogen burn criterion. Concerning the hydrogen concentration due to the reaction with sodium and sodium hydroxide in the sodium pool after the storing, in the same way, it was confirmed that it is restricted under 4% of the criterion.
Miyake, Osamu; Hamada, Hirotsugu; Tanabe, Hiromi; Wada, Yusaku; Miyakawa, Akira; Okabe, Ayao; Nakai, Ryodai; Hiroi, Hiroshi
JNC TN2400 2003-003, 225 Pages, 2004/02
JNC-TN2400-2003-003.pdf:40.45MB
The model has been developed for the assessment of the overheating tube failure in an event of sodium-water reaction accident of fast breeder reactor's steam generators (SGs). The model has been applied to the Monju SG studies.
Okabe, Ayao; Ohno, Shuji; Nakai, Ryodai; Ebashi, Masaaki
JNC TN2400 2003-002, 49 Pages, 2004/02
JNC-TN2400-2003-002.pdf:0.72MB
Sodium combustion analyses were performed using ASSCOPS version 2.1 in order to obtain background data for evaluating the validity of the mitigation system against secondary sodium leak of MONJU.
Miyake, Osamu; Hamada, Hirotsugu; Tanabe, Hiromi; ; Miyakawa, Akira; Okabe, Ayao;
JNC TN9400 2001-130, 235 Pages, 2002/03
JNC-TN9400-2001-130.pdf:7.05MB
The model has been developed for the assessment of the overheating tube failure in an event of sodium-water reaction accident of fast breeder reactor's steam generators (SGs). The model has been applied to the Monju SG studies. Major results obtained in the studies are as follows: (1)To evaluate the structural integrity of tube material, the strength standard for 2.25Cr-1Mo steel was established taking account of time dependent effect based on the high temperature (700-1200
C) creep data. This standard has been validated with the tube rupture simulation test data. (2)The conditions for overheating by the high temperature reaction were determined by use of the SWAT-3 experimental data. The realistic local heating conditions (reaction zone temperature and related heat transfer conditions) for the sodium-water reaction were proposed as the cosine-shaped temperature profile. (3)For the cooling effects inside of target tubes, LWR's studies of critical heat flux (CHF) and post-CHF heat transfer correlations have been examined and considered in the model. (4)The model has been validated with experimental data obtained by SWAT-3 and LLTR. The results were satisfactory with conservatism. The PFR superheater leak event in 1987 was studied, and the cause of event and the effectiveness of the improvement after the leak event could be identified by the analysis. (5)The model has been applied to the Monju SG studies. It is revealed consequently that no tube failure occurs in 100%, 40%, and 10% water flow operating conditions when an initial leak is detected by the cover gas pressure detection system.
Miyake, Osamu; Hamada, Hirotsugu; Tanabe, Hiromi; Okabe, Ayao; Miyakawa, Akira
JNC TN9400 2001-099, 76 Pages, 2001/11
JNC-TN9400-2001-099.pdf:2.13MB
The JNC technical report "The Development and Application of overheating Failure Model of FBR Steam Generator Tubes" summarized the assessment method and its application for the overheating tube failure in an event of sodium-water reaction accident of fast breeder reactor's steam generators (SGs). This report describes the following items studied after the publication of the above technical report. (1)0n the basis of the SWAT-3 experimental data, realistic local heating conditions (reaction zone temperature and related heat transfer conditions) for the sodium-water reaction were proposed. New correlations are cosine-shaped temperature profiles with 1,170 C maximum for the 100% and 40% Monju operating conditions, and those with 1,110 C maximum for the 10% condition. (2)For the cooling effects inside of target tubes, LWR's studies of critical heat flux (CHF) and post-CHF heat transfer correlations have been examined and considered in the assessment. The revised assessment adopts the Katto's correlation for CHF, and the Condie-Bengston IV correlation for post-CHF. (3)Other additional examination for the assessment includes treatments of the whole heating effect (other than the local reaction zone) due to the sodium-water reaction, and the temperature-dependent thermal properties of the heat transfer tube material (2.25Cr-1Mo steel). The revised overheating tube failure assessment method has been applied to the Monju SG studies. It is revealed consequently that no tube failure occurs in 100%, 40%, and 10% operating conditions when an initial leak is detected by the cover gas pressure detection system. The assessment for the SG system improved for the detection and blowdown systems shows even better safety margins against the overheating tube failure.
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JNC TJ9440 2000-010, 132 Pages, 2000/03
JNC-TJ9440-2000-010.pdf:14.85MB
In selecting the reasonable DBL on steam generator (SG), it is necessary to improve analytical method for estimating the sodium temperature on failure propagation due to overheating. Improvement on sodium-water reaction (SWR) jet code (LEAP-JET ver.1.30) and application analysis to the water injection tests for confirmation of code propriety were performed. On the improvement of the code, a gas-liquid interface area density model was introduced to develop a chemical reaction model with a little dependence on calculation mesh size. The test calculation using the improved code (LEAP-JET ver.1.40) were carried out with conditions of the SWAT-3
Run-19 test and an actual scale SG. It is confirmed that the SWR jet behavior on the results and the influence to analysis result of a model are reasonable. For the application analysis to the water injection tests, water injection behavior and SWR jet behavior analyses on the new SWAT-1 (SWAT-1R) and SWAT-3(SWAT-3R) tests were performed using the LEAP-BLOW code and the LEAP-JET code. In the application analysis of the LEAP-BLOW code, parameter survey study was performed. As the results, the condition of the injection nozzle diameter needed to simulate the water leak rate was confirmed. In the application analysis of the LEAP-JET code, temperature behavior of the SWR jet was investigated.
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JNC TJ9440 99-024, 142 Pages, 1999/03
Behavior of over heating tube rupture in sodium-water reaction have to be evaluated practically in order to confirm the propriety of DBL(Design Basis Leak) on steam generator of next large LMFBR. Over heating tube rupture is closely concerned with water / steam condition in tubes, sodium-water reaction temperature and high temperature strength of tube wall. Therefore, it is very important to precisely evaluate water / steam conditions in blow down event especially. On the other hand, as work for MONJU safety general inspection, blow down behavior was analyzed by BLOOPH code and RELAP5/Mod.2 code. LEAP-BLOW code (Ver.1.20) has been developed reflecting the acknowledgment of the MONJU blow down analysis in the blow down code for LEAP. In this code heat transfer model of sodium side of the downcommer was improved. And, using LEAP-BLOW code (Ver.1.20) MONJU blow down characteristic on the following cases was analyzed and compared with the analysis results of RELAP5/Mod.2 code and BLOOPH code. Then, it has been confirmed that there are no meaningfull difference in the results of these code, and the propriety of analysis result of LEAP-BLOW code has been confirmed. (1)Blow-down from 100% power in MONJU. (1 channel model and 2 channel model) (2)Blow-down from 100% power in MONJU. (Improvement equipment model) (3)Blow-down from Partial power in MONJU, (40% power and start-up)
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JNC TJ9440 99-023, 218 Pages, 1999/03
In selecting the reasonable DBL on steam generator (SG), it is necessary to improve analytical method for estimating the sodium temperature on failure propagation due to overheating. Improvement on the basic code for sodium-water reaction (SWR) jet was performed for an actual scale SG. The improvement points of the code are as follows; (1)introduction of advanced model such as heat transfer between the jet and structure (tube array), cooling effect of the structure, heat transfer between analytic cells, and (2)model improvement for heat transfer between two-phase flow and porous-media. The test calculation using the improved code (LEAP-JET ver.1.30) were carried out with conditions of the SWAT-3 Run-19 test and an actual scale SG. It is confirmed that the SWR jet behavior on the results is reasonable and Influence to analysis result of a model. Code integration with the blow down analytic code (LEAP-BLOW) was also studied. It is suitable that LEAP-JET was improved as one of the LEAP-BLOW's models, and it was integrated into this. In addition to above, the improvement for setting of boundary condition and the development of the interface program to transfer the analytical results of LEAP-BLOW have been performed in order to consider the cooling effect of coolant in the tube simply. However, verification of the code by new SWAT-1 and SWAT-3 test data planned in future is necessary because LEAP-JET is under development. And furthermore advancement needs to be planned.
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PNC TJ9124 98-002, 180 Pages, 1998/03
In selecting the reasonable DBL on steam generator, it is necessaly to improve analytical method of estimating the sodium-water temperature for the evaluation of failure propagation due to overheating. Using basic code for sodium-water reaction (SWR) jet, the code improvement for calculation speed up and practical analyses for functional check were carried out. The speed up methods are (1)the code improvement of time integral calculus (application of implicit method of SIMPLE) and (2)simplification of chemical reaction model (the materials properties estimation). As for calculating speed and affection on the results, the results of the improved code on the practical analyses were compared with that of the previous code. The analytical conditions was based on the case 1 (100% load conditions, normal SG pressure and non sodium flow). It is confirmed that the behavior of SWR jet on the results; distributions of void fraction and temperature is reasonable. On this improved code, the speed up options are also available. It is confirmed that the improved code is able to be speeded up in the implicit method or simplification of the properties calculation.
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PNC TJ9124 98-001, 315 Pages, 1998/01
In selecting the reasonable DBL on steam generator, it is necessary to develop analytical method for estimating the sodium temperature on failure propagation due to overheating. Using basic code for sodium-water reaction (SWR) jet, this work includes improvement of the analytical model, inspectional analyses for SWAT-3 experiments and practical analyses for Monju evaporator. The inspectional analyses were carried out for 2 cases of Run-19 & 17 on the SWAT-3. Behavior of SWR jet on the results is reasonable. Moreover, suitable value for reaction rate coefficient was evaluated. The practical analyses were carried out for 5 different conditions on the evaporator. The effects of running conditions, Na pressure and flow rate on the jet behavior were estimated for the first time. Improvement of calculating efficiency on the basic code were considered. An implicit method with SIMPLE are most suitable for applicable development and reduction of calculating load.
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PNC TJ9124 97-007, 189 Pages, 1997/03
In selecting the reasonable DBL on steam generator, it is indicated that the possibility of failure propagation due to overheating should be evaluated. This study is concerned with the design development of sodium-water reaction jet code for the LEAP in the overall development plan for the next models to evaluate the reasonable DBL; (a)blow down analysis models, (b)overheating tube bursting models (structural/ fractural dynamics) and (c)sodium-water reaction jet model for reaction zone temperature distribution analysis. In this study, basic code for sodium-water reaction jet were designed & developed with considering analytical models; two-fluid model and chemical reaction model. Reasonable results were obtained with test analyses for the two models. Basic code coupled with the models were developed. Test analyses with the code were carried out. This codes are considered to be useful for two-phase flow analysis with chemical reaction and, therefore, are most available for estimation of flow behavior on sodium-water reaction jet.
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PNC TJ9124 97-006, 295 Pages, 1997/03
In selecting the reasonable DBL(Design Basis Leak) on steam generator of next large LMFBR, it is indicated that the possibility of failure propagation due to overheating should be evaluated. Therefore, it is important to appropriately evaluate blow down behavior of water and steam in SG. For this purpose, cooling effect by water or steam in the tube should be considered appropriately or an evaluation of overheating tube rupture, and it is important adequately to select a heat transfer mode on the steam side. There, heat transfer models used in LOCA(Loss of Coolant Accident) analysis of a LWR(Light Water Reactor) have been investigated and applicable models have been employed in the LEAP-BLOW code. In addition, verification of analysis code and selection of a combination of the most suitable model has been performed using the test data of 50MWt SG. Then input and output functions have been improved.
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PNC TJ9124 96-005, 198 Pages, 1996/03
In selecting the reasonable DBL on steam generator, it is indicated that the possibility of failure propagation due to overheating should be evaluated. This study is concerned with the design development of sodium-water reaction jet code for the LEAP in the overall development plan for the next models to evaluate the reasonable DBL; a)blow down analysis models, b)overheating tube bursting models (structural/ fractural dynamics)and c)sodium-water reaction jet model for reaction zone temperature distribution analysis. In this study, sodium-water reaction jet code were designed with considering aniytical models; two-fluid model and chemical reaction model. In order to check the application of the models, test analyses for two-fluid model & chemical reaction model were carried out. These results were physically reasonable. Comparing some typical developed codes for two-phase flow analysis, SIMA/SMORC codes are considered to be useful for two-phase flow analysis with chemical reaction and, therefore, are most available for developing new sodium-water reaction jet code.
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PNC TJ9124 95-003, 242 Pages, 1995/03
In selecting the reasonable DBL on steam generator, it is indicated that the possibility of failure propagation due to overheating should be evaluated. This study is concerned with the development of blow down model for the LEAP code in the overall development plan for the next models to evaluate the reasonable DBL ; a)blow down analysis models, b)reaction zone temperature distribution analysis models and c)overheating tube bursting models (structural/fractural dynamics). In this study, blow down analysis models were developed and the analysis code was programmed. Benchmark analysis between the RELAP4 code and the developed code has been performed. Then the blow down code for the LEAP was applied to the evaluation of overheating tube bursting. Consequently, it was confirmed that the blow down code for the LEAP is applicable to the evaluation of cooling effect in the tube. Furthermore, easy coupling of this code and LEAP in future was fully considered.
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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.
*; *; Kuroha, Mitsuo; *; *; *; *
PNC TN941 85-53, 144 Pages, 1985/03
SWAC13E is a one-dimensional thermo-hydraulic computer program to analyze large scale sodium-water reaction accidents in an LMFBR steam generator. The code is the advanced version of SWAC13, the long-term hydraulic analysis module of SWACS; the energy conservation is taken into consideration in the new version to add the function to analyze the temperature behavior of the reaction. The present document covers the validation study of the code by using the large leak data of the Steam Generator Safety Test Facility (SWAT-3). The analytical parameters are as follows: (1)Model of relative velocity. (2)Void/droplet density. (3)The number of nodes where water leaks. (4)Reaction heat. It is concluded that the code can analyze the phenomena with a reasonable conservatism by choosing the proper value of the parameters.
Yamada, Fumiaki; Tabata, Hiroaki; Miyakawa, Akira; Okada, Mamoru; Okabe, Ayao; Misawa, Naoto*
no journal, ,
no abstracts in English