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Journal Articles

Detailed analyses of key phenomena in core disruptive accidents of sodium-cooled fast reactors by the COMPASS code

Morita, Koji*; Zhang, S.*; Koshizuka, Seiichi*; Tobita, Yoshiharu; Yamano, Hidemasa; Shirakawa, Noriyuki*; Inoue, Fusao*; Yugo, Hiroaki*; Naito, Masanori*; Okada, Hidetoshi*; et al.

Nuclear Engineering and Design, 241(12), p.4672 - 4681, 2011/12

 Times Cited Count:10 Percentile:31.15(Nuclear Science & Technology)

A five-year research project has been initiated in 2005 to develop a code based on the MPS (Moving Particle Semi-implicit) method for detailed analysis of key phenomena in core disruptive accidents (CDAs) of sodium-cooled fast reactors (SFRs). The code is named COMPASS (Computer Code with Moving Particle Semi-implicit for Reactor Safety Analysis). The key phenomena include (1) fuel pin failure and disruption, (2) molten pool boiling, (3) melt freezing and blockage formation, (4) duct wall failure, (5) low-energy disruptive core motion, (6) debris-bed coolability, (7) metal-fuel pin failure. Validation study of COMPASS is progressing for these key phenomena. In this paper, recent COMPASS results of detailed analyses for the several key phenomena are summarized. The present results demonstrate COMPASS will be useful to understand and clarify the key phenomena of CDAs in SFRs in details.

Journal Articles

COMPASS code development; Validation of multi-physics analysis using particle method for core disruptive accidents in sodium-cooled fast reactors

Koshizuka, Seiichi*; Morita, Koji*; Arima, Tatsumi*; Tobita, Yoshiharu; Yamano, Hidemasa; Ito, Takahiro*; Naito, Masanori*; Shirakawa, Noriyuki*; Okada, Hidetoshi*; Uehara, Yasushi*; et al.

Proceedings of 8th International Topical Meeting on Nuclear Thermal-Hydraulics, Operation and Safety (NUTHOS-8) (CD-ROM), 11 Pages, 2010/10

In this paper, FY2009 results of the COMPASS code development are reported. Validation calculations for melt freezing and blockage formation, eutectic reaction of metal fuel, duct wall failure (thermal-hydraulic analysis), fuel pin failure and disruption and duct wall failure (structural analysis) are shown. Phase diagram calculations, classical and first-principles molecular dynamics were used to investigate physical properties of eutectic reactions: metallic fuel/steel and control rod material/steel. Basic studies for the particle method and SIMMER code calculations supported the COMPASS code development. COMPASS is expected to clarify the basis of experimentally-obtained correlations used in SIMMER. Combination of SIMMER and COMPASS will be useful for safety assessment of CDAs as well as optimization of the core design.

Journal Articles

Detailed analyses of specific phenomena in core disruptive accidents of sodium-cooled fast reactors by the COMPASS code

Morita, Koji*; Zhang, S.*; Arima, Tatsumi*; Koshizuka, Seiichi*; Tobita, Yoshiharu; Yamano, Hidemasa; Ito, Takahiro*; Shirakawa, Noriyuki*; Inoue, Fusao*; Yugo, Hiroaki*; et al.

Proceedings of 18th International Conference on Nuclear Engineering (ICONE-18) (CD-ROM), 9 Pages, 2010/05

A five-year research project has been initiated in 2005 to develop a code based on the MPS (Moving Particle Semi-implicit) method for detailed analysis of specific phenomena in core disruptive accidents (CDAs) of sodium-cooled fast reactors (SFRs). The code is named COMPASS (Computer Code with Moving Particle Semi-implicit for Reactor Safety Analysis). The specific phenomena include (1) fuel pin failure and disruption, (2) molten pool boiling, (3) melt freezing and blockage formation, (4) duct wall failure, (5) low-energy disruptive core motion, (6) debris-bed coolability, and (7) metal-fuel pin failure. Validation study of COMPASS is progressing for these key phenomena. In this paper, recent COMPASS results of detailed analyses for the several specific phenomena are summarized.

Journal Articles

Validation for multi-physics simulation of core disruptive accidents in sodium-cooled fast reactors by COMPASS code

Koshizuka, Seiichi*; Morita, Koji*; Arima, Tatsumi*; Zhang, S.*; Tobita, Yoshiharu; Yamano, Hidemasa; Ito, Takahiro*; Naito, Masanori*; Shirakawa, Noriyuki*; Okada, Hidetoshi*; et al.

Proceedings of 13th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-13) (CD-ROM), 11 Pages, 2009/09

Dispersion and freezing of molten core material was calculated by the COMPASS code to compare with the experimental data of GEYSER. Molten core material flowed up with freezing on the pipe inner surface. As a molten pool behavior, CABRI-TPA2 experiment was analyzed, where a sphere of solid steel was surrounded by solid fuel. Power was injected to cause melting and boiling of the steel sphere. SCARABEE-BE+3 test was analyzed by COMPASS as a validation of failure of duct walls.

Journal Articles

Next generation safety analysis methods for SFRs, 3; Thermal hydraulics models of COMPASS code and experimental analyses

Yamamoto, Yuichi*; Hirano, Etsujo*; Oue, Masaya*; Shimizu, Sensuke*; Shirakawa, Noriyuki*; Koshizuka, Seiichi*; Morita, Koji*; Yamano, Hidemasa; Tobita, Yoshiharu

Proceedings of 17th International Conference on Nuclear Engineering (ICONE-17) (CD-ROM), 10 Pages, 2009/06

The COMPASS code is designed to analyze multi-physics problems involving thermal hydraulics, structure and phase change, in a unified framework of MPS method. In FY2006 and 2007, development of the basic functions of COMPASS was completed and fundamental verification calculations were carried out. In FY2007, the integrated verification program using available experimental data for key phenomena in CDAs was also started. In this paper, we show the basic verification calculations for the phase change model of COMPASS and the results of experimental analyses, together with the outline of the formulation of MPS method and the conceptual design of the COMPASS code.

Journal Articles

Next generation safety analysis methods for SFRs, 6; SCARABEE BE+3 analysis with SIMMER-III and COMPASS codes featuring duct-wall failure

Uehara, Yasushi*; Shirakawa, Noriyuki*; Naito, Masanori*; Okada, Hidetoshi*; Yamano, Hidemasa; Tobita, Yoshiharu; Yamamoto, Yuichi*; Koshizuka, Seiichi*

Proceedings of 17th International Conference on Nuclear Engineering (ICONE-17) (CD-ROM), 10 Pages, 2009/06

A mesoscopic approach with the COMPASS code is expected to advance the understanding of key phenomena during event progression in core disruptive accidents. In this paper, the overall analysis of SCARABEE-BE+3 test with the SIMMER-III is described as well as the simulation with COMPASS, focusing on the duct wall failure in a small temporal and spatial window cut from the SIMMER-III analysis results.

Journal Articles

COMPASS code development and validation; A Multi-physics analysis of core disruptive accidents in sodium-cooled fast reactors using particle method

Koshizuka, Seiichi*; Liu, J.*; Morita, Koji*; Arima, Tatsumi*; Zhang, S.*; Tobita, Yoshiharu; Yamano, Hidemasa; Ito, Takahiro*; Naito, Masanori*; Shirakawa, Noriyuki*; et al.

Proceedings of 2009 International Congress on Advances in Nuclear Power Plants (ICAPP '09) (CD-ROM), 1 Pages, 2009/05

A computer code, named COMPASS, is developed for multi-physics analysis of core disruptive accidents of sodium-cooled fast reactors (SFRs). A meshless method, called MPS method, is employed since complex thermal-hydraulics and structural problems with various phase change processes have to be analyzed. Verification for separeted basic processes and validation for practical phenomena are carried out. COMPASS is also expected to investigate molten fuel discharge to avoid re-criticality in large size SFR cores. Both MOX and metal fuels are considered. Eutectic reactions between the metal fuel and the cladding material are investigated by phase diagram calculation, classical and first-principles molecular dynamics. Basic studies relevant to the numerical methods support the code development of COMPASS. Parallel processing is implemented by OpenMP to treat large-scale problems. A visualization tool is also prepared by using AVS.

Journal Articles

Code development for multi-physics and multi-scale analysis of core disruptive accidents in fast reactors using particle methods

Koshizuka, Seiichi*; Morita, Koji*; Arima, Tatsumi*; Zhang, S.*; Tobita, Yoshiharu; Yamano, Hidemasa; Ito, Takahiro*; Shirakawa, Noriyuki*; Naito, Masanori*; Okada, Hidetoshi*; et al.

Proceedings of 16th Pacific Basin Nuclear Conference (PBNC-16) (CD-ROM), 6 Pages, 2008/10

A computer code, named COMPASS, is being developed for various complex phenomena of core disruptive accidents (CDAs) in sodium-cooled fast reactors (SFRs). The COMPASS is designed to analyze multi-physics problems involving thermal hydraulics, structure and phase change, in a unified framework of the MPS (Moving Particle Semi-implicit) method. The project has been carried out by six organizations for five years from FY2005 to FY2009. In this paper, the outcomes of the project in FY2007 are presented. Three validation calculations were completed by following the validation plan: melt freezing and blockage formation, molten pool boiling, and duct wall failure. The COMPASS code development was supported by basic studies of the numerical method, material science for eutectic reaction of the metal fuel, and SIMMER-III analyses.

Journal Articles

Code development for core disruptive accidents in sodium-cooled fast reactors

Koshizuka, Seiichi*; Liu, J.*; Morita, Koji*; Arima, Tatsumi*; Zhang, S.*; Tobita, Yoshiharu; Yamano, Hidemasa; Ito, Takahiro*; Naito, Masanori*; Shirakawa, Noriyuki*; et al.

Proceedings of IAEA Topical Meeting on Advanced Safety Assessment Methods for Nuclear Reactors (CD-ROM), 9 Pages, 2007/10

A computer code, named COMPASS (Computer Code with Moving Particle Semi-implicit for Reactor Safety Analysis), is being developed for various complex phenomena of core disruptive accidents (CDAs) in sodium-cooled fast reactors (SFRs). Theoretical studies are performed about a unified algorithm for compressible and incompressible flows, fluid flow with solid debris, and algorithm improvement for free surface flows. Code verification and validation procedures are established by exploiting the past experiences in those of SIMMER-III code. COMPASS will be used for separated phenomena in CDAs, while the whole core will be analyzed by SIMMER-III. COMPASS is expected to clarify the detailed process in duct wall failure and fuel discharge to avoid re-criticality during CDAs in large size SFRs.

Journal Articles

Multi-physics and multi-scale simulation for core disruptive accidents in fast breeder reactors

Koshizuka, Seiichi*; Liu, J.*; Morita, Koji*; Arima, Tatsumi*; Tobita, Yoshiharu; Yamano, Hidemasa; Ito, Takahiro*; Shirakawa, Noriyuki*; Hosoda, Seigo*; Araki, Kazuhiro*; et al.

Proceedings of 5th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-5), p.472 - 479, 2006/11

A 5-year research project started in FY2005 in the framework of Innovative Nuclear Research and Development Program funded by the Ministry of Education, Culture, Sports, Science and Technology in Japan. A computer code, named COMPASS (Computer Code with Moving Particle Semi-implicit for Reactor Safety Analysis), is being developed using the Moving Particle Semi-implicit (MPS) method for various complex phenomena of severe accidents in fast breeder reactors. Both MOX and metal fuels are considered. Eutectic reactions between the metal fuel and the cladding material are being investigated by molecular dynamics and molecular orbital methods. The molten metal flow with solidification was analyzed by MPS. The elastic analysis of a hexagonal wrapper tube was analyzed by the MPS method as well. The results were compared with an experiment and an calculation using an commercial code. Eutectic reactions were calculated by molecular dynamics and compared with the references. We found that the combination of the above numerical methods was useful for multi-physics and multi-scale phenomena of core disruptive accidents in fast breeder reactors.

JAEA Reports

Evaluation of the correlation model for the contact areas concentration between sodium and steam with the particle interaction method

Horie, Hideki*; Yamamoto, Yuichi*; Oue, Masaya*; Shirakawa, Noriyuki*

JNC-TJ9400 2005-007, 135 Pages, 2004/02

JNC-TJ9400-2005-007.pdf:7.19MB

In a LMFR steam generator, liquid sodium flows through the component vessel, which has a manifold of heat transfer tubes through which water or steam flows under very high pressure. If the water or steam is issued as a jet into sodium pool by the high pressure due to tube failure, sodium-water reaction occurs and affects the component integrity. The phenomena are strongly nonlinear processes in multi-phase flow. To evaluate contact areas concentration between fluid components is essential to analyze the phenomena. In this work, the correlation model for the contact areas concentration between two different liquids developed with the particle interaction method, which method is capable of evaluating the mixing layer between two kinds of fluid, was applied to a gasjet issued into a liquid pool to investigate its applicability. This investigation involves the analysis to consider the mole change of steam and generated gas due to the sodium-water chemical reaction and the analysis of the effect of a rod on the correlation model.

JAEA Reports

Advanced modeling of the contact area correlation between sodium and water with the particle interaction method

Shirakawa, Noriyuki*; Yamamoto, Yuichi*; Horie, Hideki*

JNC-TJ9400 2005-005, 103 Pages, 2003/02

JNC-TJ9400-2005-005.pdf:6.41MB

In a LMFR steam generator, liquid sodium flows through the component vessel, which has a manifold of heat transfer tubes through which water or steam flows in very high pressure. If the water or steam is issued in a jet into sodium pool by the high pressure due to tube failure, sodium-water reaction occurs and affects the component integrity. The phenomena are strongly nonlinear processes in multi-phase flow. To evaluate contact areas concentration between fluid components is essential to compute the phenomena. In our previous investigations, mesoscopic and direct analysis method has been developed by using the particle interaction method in order to compute multi-phase, multi-component, and chemically interactive fluids. With this method, flow regime and contact areas concentration were investigated and a correlation model for a liquid jet flow issued into another liquid was made. In this fiscal year, the correlation model was investigated in detail to give higher accuracy prior to developing the correlation for a gas jet flow issued into liquid. Furthermore, the experiment of gas jet flow issued into liquid was analyzed with the particle interaction method to confirm applicability to gas-liquid system.

JAEA Reports

Analysis of the flow with sodium-water chemical reaction by the particle interaction method

Shirakawa, Noriyuki*; Horie, Hideki*; Yamamoto, Yuichi*

JNC-TJ9400 2005-006, 183 Pages, 2002/02

JNC-TJ9400-2005-006.pdf:10.61MB

To evaluate the effect of accidents induced by sodium-water chemical reaction on a LMFR component, the numerical thermo-hydraulic analysis should involve the whole boundary of the component. Therefore, the thermo-hydraulic code is required to model the chemically reactive fluids dynamics with constitutive correlations.Both thermal and chemical reaction rate largely depends on the binary contact areas between components such as continuous liquids, droplets, solid particles, and bubbles. The contact areas change sharply according to the interface state between components. Since no experiment to investigate the jet flow with sodium-water chemical reaction has been done, the purpose of this study is to develop the evaluation method for flow regimes and contact areas by analyzing the fluid dynamics of multi-phase and reactive components mechanistically with the particle interaction method. In this fiscal year, following works were performed:(1) investigation to link the mesoscopic information of contact areas obtained by the particle interaction method with the macroscopic fluid dynamics code, (2) development of the correlation of contact areas, (3) investigation of the effect of the water-leak conditions on contact areas, and (4) analysis of contact areas for a slug flow.

JAEA Reports

Analysis of the flow with phase change and chemical reaction with the paticle interaction method (Report under the contract between JNC and toshiba Corporation)

Shirakawa, Noriyuki*; *; *

JNC-TJ9400 2001-006, 93 Pages, 2001/02

JNC-TJ9400-2001-006.pdf:3.27MB

The numerical thermohydraulic analysis of a LMFR component should involve its whole boundary in order to evaluate the effect of chemical reaction within it. Therefore, it becomes difficult mainly due to computing time to adopt microscopic approach for the chemical reaction directly. Thus, the thermohydraulic code is required to model the chemically reactive fluid dynamics with constitutive correlations. The reaction rate depends on the binary contact areas between components such as continuous liquids, droplets, solid particles, and bubbles. The contact areas change sharply according to the interface state between components. Since no experiments to study the jet flow with sodium-water chemical reaction have been done, the goal of this study is to obtain the knowledge of flow regimes and contact areas by analyzing the fluid dynamics of multi-pahse and reactive components mechanistically with the particle interaction method. In this fiscal year, following works were performed: (1)Development and coding of the interfacial area model, (2)Development and coding of the phase change model, (3)Verification of the fundamental functions of the models, and (4)Literature investigation of the related experiments.

JAEA Reports

A feasibility study of the particle interaction method for the flow regimes with the chemical reaction; (Report under the contract between JNC and Toshiba Corporation)

Shirakawa, Noriyuki*; *; *; *

JNC-TJ9440 2000-008, 47 Pages, 2000/03

JNC-TJ9440-2000-008.pdf:1.96MB

The numerical thermohydraulic analysis of a LMFR component should involve its whole boundaly in order to evaluate the effect of chemical reaction within it. Therefore, it becomes difficult mainly due to computing time to adopt microscopic approach for the chemical reaction directly. Thus, the thermohydraulic code is required to model the chemically reactive fluid dynamics with constitutive correlations. The reaction rate denpends on the binary contact areas between components such as continuous liquids, droplets, solid particles, and bubbles. The contact areas change sharply according to the interface state between components. Since no experiments to study the jet flow with sodium-water chemical reaction have been done, the goal of this study is to obtain the knowledge of flow regimes and contact areas by analyzing the fluid dynamics of multi-pahse and reactive components mechanistically with the particle interaction method. For the first stage of the study, the applicability of this method to the nalysis of a liquid jet into the other liquid pool was investigated. Based on the literatures, we investigated the jet flow mechanisms and analyzed the experiment of a water jet into a gasoline pool. We also analyzed SWAT3/Run19 test, the jet flow in a rod bundle, to study the applicability of the method to a complicated boundary without a chemical reaction model. The calculated fluid dynamics was in good agreement with the experiment. Furthermore, we studied and formulated the paths of phase change and chemical reaction, and conceptually designed the adopting the heat-transfer-limited phase change model and the synthesizd reaction model with a water-hydrogen conversion ratio.

JAEA Reports

Improvement of the sodium-water chemical reaction analysis code SIMMER-SW (2); (Report under the contract between JNC and Toshiba Corporation)

Shirakawa, Noriyuki*; *; *; *

JNC-TJ9440 99-009, 195 Pages, 1999/03

JNC-TJ9440-99-009.pdf:5.93MB

It is necessary for the evaluation of the design base flow rate of water leakage out of the steam generator heat transfer tubes to evaluate the possibility of a tube-failure propagation quantitatively, which needs development of the followings, (1)Blow down model of the steam generator heat transfer tube, (2)Rupture model at high temperature of the steam generator heat transfer tube, and (3)Sodium-water chemical reaction model to analyze the temperature distribution around a leakage point. In this study, development effort is focused on the item (3)that is most important to the evaluation of failure propagation. The FBR safety aalysis code SIMMER-III is used as a reference to realize the analysis and is named SIMMER-SW. The followings were done in this work: (a)Coding of the chemical reaction model, and verification of the model functions, (b)Coding of the rod bundle pressure drop model, and verification of the model functions, (3)Coding of the preprocessor to prepare the input for a rod bundle, (4)Data production of the EOS(Equation-of-State) and TPP(Thermo Physical Properties), (5)Investigation of the interaction between sodium and jet out of a leak hole, (6)Modeling of the thermocouple, (7)SWAT1/P06 test analysis, and (8)SWAT3/Run19 test analysis. The following knowledge was obtained: (a)The results of the SWAT1/P06 test analysis which can be done in two-dimension agree well with that of the experiment at least qualitatively, (b)Thermocouple model is very useful in the analysis such that various kinds of components contact a thermocouple, (c)The constant K in the chemical reaction model is likely to be 0.01$$<$$K$$<$$0.1, and (d)The results of the SWAT3/Run19 test analysis which should be done in three-dimension do not agree with that of the experiment.

JAEA Reports

Development of data base for rational evaluation of safety design, 2

*; *; *; Shirakawa, Noriyuki*

PNC-TJ2164 87-005, 423 Pages, 1987/04

PNC-TJ2164-87-005.pdf:9.45MB

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Journal Articles

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; ; Morita, Koji; Shirakawa, Noriyuki

Donen Giho, , 

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Journal Articles

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Morita, Koji; ; Shirakawa, Noriyuki;

Int.Top.Mtg.on Sodium Cooled Fast React, , 

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Journal Articles

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Morita, Koji; ; ; Shirakawa, Noriyuki

International Conference on Multiphuse Flows'91TU, , 

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22 (Records 1-20 displayed on this page)