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Journal Articles

Development of computational method for predicting vortex cavitation in the reactor vessel of JSFR

Hamada, Noriaki*; Shiina, Koji*; Fujimata, Kazuhiro*; Hayakawa, Satoshi*; Watanabe, Osamu*; Yamano, Hidemasa

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles (FR 2009) (CD-ROM), 10 Pages, 2012/00

In a sodium-cooled fast reactor, a vortex cavitation evaluation methodology was developed to predict a possible cavitation generated by vortex at the center of accelerating flow. This methodology was applied to a scaled model experiment, leading to the prospect that the cavitation can be predicted.

Journal Articles

Study on flow-induced-vibration evaluation of large-diameter pipings in a sodium-cooled fast reactor, 1; Sensitivity analysis of turbulent flow models for unsteady short-elbow pipe flow

Aizawa, Kosuke; Nakanishi, Shigeyuki; Yamano, Hidemasa; Kotake, Shoji; Hayakawa, Satoshi*; Watanabe, Osamu*; Fujimata, Kazuhiro*

Proceedings of 6th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-6) (USB Flash Drive), 7 Pages, 2008/11

To evaluate the flow-induced vibration in the actual-sized pipings of JSFR, computer simulation is necessary. In this study, as the first step, sensitivity analysis of turbulence flow models for unsteady short-elbow pipe flow has been carried out with the STAR-CD thermal-hydraulic simulation code. Through the sensibility analysis, the objective of this study is to propose the best analysis models which can reproduce the unsteady characteristics obtained in the 1/3-scale test results with 9.2 m/s of main flow. In this study, to take into account anisotropic characteristics of turbulence, two turbulent flow models were used: large eddy simulation (LES) and Reynolds stress model (RSM). The both validated simulations have reproduced flow separation region and periodic vortex shedding. The simulation results with both models were compared with power spectrum densities of pressure fluctuations which were used in the pipe vibration evaluation. Only the RSM simulation with the best combination has reproduced the pressure-fluctuation power spectrum densities, which were characterized by a peak frequency of 10 Hz in the 1/3 test with 9.2 m/s.

Journal Articles

Numerical calculation of fluid flow within a large-diameter piping with a short-radius elbow in JSFR

Aizawa, Kosuke; Yamano, Hidemasa; Kotake, Shoji; Fujimata, Kazuhiro*

Proceedings of 7th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operation and Safety (NUTHOS-7) (CD-ROM), 14 Pages, 2008/10

The present study has numerically investigated using the STAR-CD code the fluid-flow characteristics within a large-diameter piping with a short-radius elbow, which is adopted in an advanced large-sized sodium-cooled fast reactor (named JSFR). This study reports the result of numerical steady-state calculations of the 1/3-scale experiment with 9.2m/s of velocity performed at the first step. Since the experiments have revealed that dominant fluctuating pressures were generated at the boundary of flow separation and reattachment point on the pipe wall, this study focused on the flow separation size as one of the flow characteristics. The calculation has reproduced the flow characteristics, such as the measured velocity profile in the flow separation region, by specifying appropriate analytical models and conditions. With the validated models, the effect of the coolant viscosity has also been investigated as well as the piping scale. In order to examine the disturbance at the piping inlet, the flow dynamics within the reactor vessel were also calculated by modeling an entire upper sodium plenum region including various components within the reactor vessel in the JSFR design. This upper plenum calculation had to reduce the spatial resolution within the hot-leg piping because of numerous computational meshes needed in this calculation. The plenum calculation has shown several vortexes and flow distortion at the hot-leg inlet. The hot-leg inlet flow condition obtained in the plenum calculation was interpolated for the calculation simulating the hot-leg piping, where the spatial resolution was better than in the plenum calculation. The numerical calculation under the reactor condition involving the inlet disturbance has indicated the flow separation size became smaller than that in no disturbance case. This calculation implies that the inlet disturbance may play an important role to mitigate the flow-induced vibration force in the flow separation region.

JAEA Reports

Study of hydraulic behavior for reactor upper plenum in sodium-cooled fast reactor; Verification analysis of water experiment and applicability of vortex prediction method

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.

JAEA Reports

Improvement of blow down model for LEAP code

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.

JAEA Reports

Summary of requirements on SASS for a large FBR Core

*; 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$$^{circ}$$C (911K) to avoid the possibility of unintended rod drop under normal operations, and to be set below 666$$^{circ}$$C (939K) to prevent coolant boiling under ULOF conditions conservatively assuming a boiling temperature of 960$$^{circ}$$C 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 20.5$$^{circ}$$C/$$sigma$$ (1$$sigma$$= 6.7$$^{circ}$$C 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 666$$^{circ}$$C, the SASS has a safety margin of about 3.4$$sigma$$ 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 (1018$$^{circ}$$C) under the ULOF conditions of this plant.

JAEA Reports

Improvement on reaction model for sodium-water reaction; Jet code and application analysis

*; *; *; *; *

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$$cdot$$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.

JAEA Reports

Verification of blow down code for LEAP code; Verification by RELAP5/Mod.2 and BLOOPH code

*; *; *; *

JNC TJ9440 99-024, 142 Pages, 1999/03

JNC-TJ9440-99-024.pdf:7.16MB

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)

JAEA Reports

Improvement and test calculation on basic code for sodium-water reaction jet

*; *; *; *; *

JNC TJ9440 99-023, 218 Pages, 1999/03

JNC-TJ9440-99-023.pdf:33.77MB

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 $$cdot$$ 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.

JAEA Reports

Speed up improvement on basic code for sodium-water reaction jet

*; *; *; *; *; *

PNC TJ9124 98-002, 180 Pages, 1998/03

PNC-TJ9124-98-002.pdf:3.8MB

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.

JAEA Reports

Improvement and verification of blow down code for LEAP code; Comparison with blow down test data of 50MWtSG

*; *; *; *

PNC TJ9124 97-006, 295 Pages, 1997/03

PNC-TJ9124-97-006.pdf:6.03MB

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.

JAEA Reports

Development of blow down model for the LEAP code

*; *; *; *; *

PNC TJ9124 95-003, 242 Pages, 1995/03

PNC-TJ9124-95-003.pdf:5.48MB

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.

JAEA Reports

Preliminary study on modification of LEAP

*; *; *; *

PNC TJ9124 94-009, 164 Pages, 1994/03

PNC-TJ9124-94-009.pdf:4.63MB

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.

JAEA Reports

Long-term thermo-hydraulic analysis in large-scale sodium-water reaction (Analysis of SWAT-3 Runs 4, 5, 6 and 7 by SWAC-13E); Large-scale sodium-water reaction analysis (Report No.14)

*; *; Kuroha, Mitsuo; *; *; *; *

PNC TN941 85-53, 144 Pages, 1985/03

PNC-TN941-85-53.pdf:3.01MB

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.

Oral presentation

R&D issues in structural design standard of fast reactor, 16; Application of inelastic analysis to piping design

Fujimata, Kazuhiro*; Nagashima, Hideaki*; Sukekawa, Masayuki*; Shibamoto, Hiroshi; Inoue, Kazuhiko*; Kasahara, Naoto

no journal, , 

no abstracts in English

Oral presentation

Consideration on applicability of turbulent flow model to flow dynamics in a short-elbow pipe

Aizawa, Kosuke; Yamano, Hidemasa; Uto, Nariaki; Kotake, Shoji; Watanabe, Osamu*; Fujimata, Kazuhiro*

no journal, , 

A conceptual design study of a large-scale sodium-cooled fast reactor adopts a two-loop primary cooling system with large-diameter piping in order to reduce plant construction cost. In this design, one of issues is a flow-induced vibration behavior of the piping under a high Reynolds number of 10$$^7$$ order. To evaluate the piping integrity, it is necessary to obtain power spectrum densities of pressure fluctuations on the piping wall by a numerical unsteady flow simulation. In this study, the numerical simulation capability of Reynolds stress model and large-eddy simulation in the STAR-CD code has been investigated using the 1/3-scale hot-leg test data. Through the sensitivity analysis, the Reynolds stress model with appropriate analytical models has shown the best applicability to flow dynamics simulation in the short-elbow pipe.

Oral presentation

Development of computational method for predicting vortex cavitation in the reactor vessel of JSFR

Hamada, Noriaki*; Fujimata, Kazuhiro*; Shiina, Koji*; Hayakawa, Satoshi*; Watanabe, Osamu*; Yamano, Hidemasa

no journal, , 

A computational method for predicting vortex cavitation based on the theory of vortex streching was developed to predict possible cavitation generated by vortex at the center of accelerating swirl flow in the reactor vessel in a sodium-cooled fast reactor. This method was applied to a scale model test of a commercial fast reactor, leading to feasibility of this method that can predict the cavitation.

Oral presentation

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