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Fujimura, Koji*; Itooka, Satoshi*; Oki, Shigeo; Takeda, Toshikazu*
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 6 Pages, 2017/04
Fujii, Tadashi; Chikazawa, Yoshitaka; Konomura, Mamoru; Kamide, Hideki; Kimura, Nobuyuki; Nakayama, Okatsu; Ohshima, Hiroyuki; Narita, Hitoshi*; Fujimata, Kazuhiro*; Itooka, Satoshi*
JAEA-Research 2006-017, 113 Pages, 2006/03
JAEA-Research-2006-017.pdf:14.98MB
A conceptual design study of the sodium-cooled fast reactor is in progress in the Feasibility Study on Commercialized Fast Reactor Cycle Systems. Reduced scale water experiments are being performed in order to clarify the flow pattern in the upper plenum of the reactor which has higher velocity condition than the past design. In this report, the hydraulic analyses of the water experiments using the general-purpose thermal hydraulic analysis program were executed; and the applicability to evaluation of flow pattern and vortex cavitations for the designed reactor was examined. (1) Steady-state analyses under the Froude number similar condition were carried out for the 1/10th reduced scale plenum experiments. Analyses results reproduced the characteristic flow patterns in the upper plenum, such as gushed flow from the inside of the upper internal structure to reactor vessel wall and the jet flow from the slit of the upper internal structure. Further, it was confirmed that the calculated flow pattern of a designed reactor system agreed with that of the water experiment qualitatively. Moreover, the influence which setting of numerical solution and boundary condition etc. in analyzing causes to flow pattern in the plenum became clear. (2) The distribution of the vortices under the dipped plate region in the 1/10th plenum model was evaluated using the prediction method of a submerged vortex which is based on the stretching vortex theory. In case of the same velocity condition as the reactor, it identified the two vortices which were sucked into the hot leg piping from the cold leg piping wall as the submerged vortex cavitations. From this analysis result, it confirmed that the submerged vortex cavitations, which may occur in the reactor upper plenum steadily, could be identified using this prediction method.
Itooka, Satoshi*; Fujimata, Kazuhiro*
JNC TJ9440 2003-001, 286 Pages, 2003/03
JNC-TJ9440-2003-001.pdf:9.23MB
In Japan Nuclear Cycle Development Institute, the improvement of analysis method for overheating tube rapture was studied for the accident of sodium-water reactions in the steam generator of a fast breeder reactor and the evaluation of heat transfer condition in the tube were carried out based on study of critical heat flux (CHF) and post-CHF heat transfer equation in Light Water Reactors. In this study, the improvement of blow down model for the LEAP code was carried out taking into consideration the above-mentioned evaluation of heat transfer condition. Improvements of the LEAP code were following items. Calculations and verification were performed with the improved LEAP code in order to confirm the code functions. (1)The addition of critical heat flux (CHF) by the formula of Katto and the formula of Tong. (2)The addition of post-CHF heat transfer equation by the formula of Condie-Bengston Ⅳ and the formula of Groeneveld 5.9. (3)The physical properties of the water and steam are expanded to the critical conditions of the water. (4)The expansion of the total number of section and the improvement of the input form. (5)The addition of the function to control the valve setting by the PID control model.
*; Sawada, Shusaku*; *; *; *
JNC TJ9400 2001-011, 159 Pages, 2001/03
JNC-TJ9400-2001-011.pdf:5.21MB
In order to improve the reliability of safety system of FBR employing passive safety functions, requirements on a self-actuated shutdown system (SASS) have been summarized, which employs Curie-point magnets in its safety rods, for a large homogeneous reactor core (1500MWe), which was designed by JNC in JFY1999 as one of candidates in the Feasibility Studies on Commercialized FBR System. The requirements were based on the sensitivity analyses, conducted by using the safety analysis code SAS4A, of uncertainty factors that affect the maximum coolant temperature in unprotected loss-of-flow events. The study has given the following requirements: (1)The detachment temperature of the SASS magnet is to be set above 638
C (911K) to avoid the possibility of unintended rod drop under normal operations, and to be set below 666C (939K) to prevent coolant boiling under ULOF conditions conservatively assuming a boiling temperature of 960C in the plant. (2)The parameter that has the largest sensitivity on the maximum coolant temperature under ULOF conditions is the exit coolant temperatures of the adjacent fuel assemblies to the SASS assembly, with a sensitivity coefficient of 20.5C/
(1= 6.7C for the exit coolant temperature). It has been concluded based on the parametric analyses that a SASS design even with a nominal detachment temperature of 666C, the SASS has a safety margin of about 3.4 for the exit coolant temperature of the most sensitive adjacent fuel assembly in preventing coolant boiling under ULOF conditions, if the incoherency of SASS detachments are appropriately taken into account together with a more realistic criterion on the coolant boiling temperature (1018C) under the ULOF conditions of this plant.
*; *; *; *; *
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.
*; *; *; *
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)
*; *; *; *; *
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.
*; *; *; *; *; *
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.
*; *; *; *; *; *
PNC TJ9124 98-003, 195 Pages, 1998/01
Experiments were carried out to understand burning behavior of a falling liquid sodium drop and to get verification data for an analytical program. The main results of this research were as follows. (1)Design and manufacturing of experimental devices. Devices were designed and manufactured, which generated liquid sodium drops with diameter of 3.5
1.0 mm and with temperature of 500 10C and dropped them in the atmosphere of inert gas and air. (2)Measurement of liquid sodium drop diameters of the inert gas experiments. Four experiments were conducted, in which 50 drops of liquid sodium were fell into inert gas atmosphere. An averaged diameter of liquid sodium drop was about 3.8 mm. (3)Measurement of burned weight and drop velocity of the air atmosphere experiments. Six experiments were conducted, in which 50 drops of liquid sodium fell into air atmosphere. An averaged burned weight of the liquid sodium drop was about 3.4 mg. The velocity after about 2.4 m drop was evaluated from photographs. Which indicated that the falling velocity of burned and un-burned single liquid sodium drop was 5.5 0.1 m/s.
*; *; *; *
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.
*; *; *; *; *
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.
*; *; *; *
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.