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Journal Articles

Numerical simulation on self-leveling behavior of mixed particle beds using multi-fluid model coupled with DEM

Phan, L. H. S.*; Ohara, Yohei*; Kawata, Ryo*; Liu, X.*; Liu, W.*; Morita, Koji*; Guo, L.*; Kamiyama, Kenji; Tagami, Hirotaka

Proceedings of 12th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation and Safety (NUTHOS-12) (USB Flash Drive), 12 Pages, 2018/10

Self-leveling behavior of core fuel debris beds is one of the key phenomena for the safety assessment of core disruptive accidents (CDAs) in sodium-cooled fast reactors (SFRs). The SIMMER code has been developed for CDA analysis of SFRs, and the code has been successfully applied to numerical simulations for key thermal-hydraulic phenomena involved in CDAs as well as reactor safety assessment. However, in SIMMER's fluid-dynamics model, it is always difficult to represent the strong interactions between solid particles as well as the discrete particle characteristics. To solve this problem, a new method has been developed by combining the multi-fluid model of the SIMMER code with the discrete element method (DEM) for the solid phase to reasonably simulate the particle behaviors as well as the fluid-particle interactions in multi-phase flows. In this study, in order to validate the multi-fluid model of the SIMMER code coupled with DEM, numerical simulations were performed on a series of self-leveling experiments using a gas injection method in cylindrical particle beds. The effects of friction coefficient on the simulation results were investigated by sensitivity analysis. Though more extensive validations are needed, the reasonable agreement between simulation results and corresponding experimental data preliminarily demonstrates the potential ability of the present method in simulating the self-leveling behaviors of debris bed. It is expected that the SIMMER code coupled with DEM is a prospective computational tool for analysis of safety issues related to solid particle debris bed in SFRs.

Journal Articles

Modeling of eutectic reaction between molten stainless steel and B$$_{4}$$C for severe accident simulations

Liu, X.*; Morita, Koji*; Yamano, Hidemasa

Proceedings of 12th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation and Safety (NUTHOS-12) (USB Flash Drive), 12 Pages, 2018/10

On the basis of experimental results, growth of the eutectic material is modeled by the parabolic rate law. Heat and mass transfer processes are also modeled considering both the equilibrium and non-equilibrium phase changes of eutectic material. Thermophysical properties of eutectic material obtained from the experimental measurements are also included in the analytic thermophysical property model and analytic equation-of-state model.

Journal Articles

Preliminary analysis of the post-disassembly expansion phase and structural response under unprotected loss of flow accident in prototype sodium cooled fast reactor

Onoda, Yuichi; Matsuba, Kenichi; Tobita, Yoshiharu; Suzuki, Toru

Mechanical Engineering Journal (Internet), 4(3), p.16-00597_1 - 16-00597_14, 2017/06

Journal Articles

Improvements to the simmer code model for steel wall failure based on EAGLE-1 test results

Toyooka, Junichi; Kamiyama, Kenji; Tobita, Yoshiharu; Suzuki, Toru

Proceedings of 10th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-10) (USB Flash Drive), 7 Pages, 2016/11

Journal Articles

Preliminary analysis of the post-disassembly expansion phase and structural response under unprotected loss of flow accident in prototype sodium cooled fast reactor

Onoda, Yuichi; Matsuba, Kenichi; Tobita, Yoshiharu; Suzuki, Toru

Proceedings of 24th International Conference on Nuclear Engineering (ICONE-24) (DVD-ROM), 10 Pages, 2016/06

Journal Articles

Numerical simulations of gas-liquid-particle three-phase flows using a hybrid method

Guo, L.*; Morita, Koji*; Tobita, Yoshiharu

Journal of Nuclear Science and Technology, 53(2), p.271 - 280, 2016/02

 Times Cited Count:4 Percentile:31.55(Nuclear Science & Technology)

Journal Articles

A Numerical study on local fuel-coolant interactions in a simulated molten fuel pool using the SIMMER-III code

Cheng, S.; Matsuba, Kenichi; Isozaki, Mikio; Kamiyama, Kenji; Suzuki, Toru; Tobita, Yoshiharu

Annals of Nuclear Energy, 85, p.740 - 752, 2015/11

 Times Cited Count:13 Percentile:7.42(Nuclear Science & Technology)

Journal Articles

First analysis of local fuel-coolant interactions in a molten pool by SIMMER-III using reactor materials

Cheng, S.; Matsuba, Kenichi; Isozaki, Mikio; Kamiyama, Kenji; Suzuki, Toru; Tobita, Yoshiharu

Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 9 Pages, 2015/05

Journal Articles

A Preliminary evaluation of unprotected loss-of-flow accident for a prototype fast-breeder reactor

Suzuki, Toru; Tobita, Yoshiharu; Kawada, Kenichi; Tagami, Hirotaka; Sogabe, Joji; Matsuba, Kenichi; Ito, Kei; Ohshima, Hiroyuki

Nuclear Engineering and Technology, 47(3), p.240 - 252, 2015/04

 Times Cited Count:11 Percentile:10.49(Nuclear Science & Technology)

Journal Articles

SIMMER-III analyses of local fuel-coolant interactions in a simulated molten fuel pool; Effect of coolant quantity

Cheng, S.; Matsuba, Kenichi; Isozaki, Mikio; Kamiyama, Kenji; Suzuki, Toru; Tobita, Yoshiharu

Science and Technology of Nuclear Installations, 2015, p.964327_1 - 964327_14, 2015/00

 Times Cited Count:1 Percentile:81.41(Nuclear Science & Technology)

JAEA Reports

An Evaluation study of measures for prevention of Re-criticality in sodium-cooled large FBR with MOX fuel

JNC-TN9400 2000-038, 98 Pages, 2000/04

JNC-TN9400-2000-038.pdf:7.49MB

As an effort in the feasibility study on commercialized Fast Breeder Reactor cycle systems, an evaluation of the measures to prevent the energetic re-criticality in sodium-cooled large MOX core, which is one of the candidates for the commercialized reactor, has been performed. The core disruptive accident analysis of Demonstration FBR showed that the fuel compaction of the molten fuel by radial motion in a large molten core pool had a potential to drive the severe super-prompt re-criticality phenomena in ULOF sequence. ln order to prevent occurrence of the energetic re-criticality, a subassembly with an inner duct and the removal of a part of LAB are suggested based on CMR (Controlled Material Relocation) concept. The objective of this study is the comparison of the effectiveness of CMR among these measures by the analysis using SIMMER-III. The molten fuel in the subassembly with inner duct flows out faster than that from other measures. The subassembly with inner duct will work effectively in preventing energetic re-criticality. Though the molten fuel in the subassembly without a part of LAB flows out a little slower, it is still one of the promising measures. However, the UAB should be also removed from the same pin to prevent the fuel re-entries into the core region due to the pressurization by FCl below the core, unless it disturbs the core performance. The effect of the axial fuel length of the center pin to CMR behavior is small, compared to the effect of the existence of UAB.

JAEA Reports

SIMMER-III Analysis, of Gas-Liquid Flow with Large Liquid Density

Suzuki, Toru; Tobita, Yoshiharu

JNC-TN9400 2000-019, 35 Pages, 2000/03

JNC-TN9400-2000-019.pdf:1.79MB

The transition phase analysis code SIMMER-III has been developed to appropriately evaluate the core disruptive accident in a fast breeder reactor. The momentum exchange model used in the fluid dynamics portion of the code uses the conventional correlation based on ordinary flows such as air-water flows. lt has already been confirmed that this code can represent the experimental results of ordinary flows. However, more detailed research is needed to confirm that this code is applicable to two-phase flow with large liquid density, which would be formed in an actual molten core pool. In addition, since the shapes of bubbles affect their drag in the bubbly flow where the liquid and the gas form continuous and dispersed phases, respectively, it is necessary to take this effect of bubble shape into account to improve SIMMER-III's analytical precision. ln this study, using experimental results obtained through a joint research program with Kyoto university, the momentum exchange model of SIMMER-III is assessed with regard to the bubbly flow regime of two-phase flow with large liquid density, on which experimental data and information on bubble shapes had been lacking. This study suggests that the original SIMMER-III can appropriately represent the characteristics of bubbly flows containing ellipsoidal bubbles with relatively small gas flux. Moreover, this study shows that the precision of SIMMER-III for bubbly flows containing cap bubbles with relatively large gas flux is much improved by using Kataoka-lshii's correlation to determine the drag coefficient of bubbles in the momentum exchange model.

JAEA Reports

SIMMER-III Analytic Equation-of-State Model

Morita, Koji; Tobita, Yoshiharu; kondo, Satoru; E.A.Fischer*

JNC-TN9400 2000-005, 57 Pages, 1999/05

JNC-TN9400-2000-005.pdf:2.92MB

An improved analytic equation-of-state (EOS) model using flexible thermodynamic functions is developed for a reactor safety analysis code, SIMMER-III. The present EOS model is designed to have adequate accuracy in describing thermodynamic properties of reactor-core materials over wide temperature and pressure ranges and to consistently satisfy basic thermodynamic relationships without deterioration of the computing efficiency. The fluid-dynamic algorithm for pressure iteration consistently coupled with the EOS model is also described in the present report. The EOS data of the basic core materials, uranium dioxide, mixed-oxide fuel, stainless steel, and sodium, are developed up to the critical point by compiling the most up-to-date and reliable sources using basic thermodynamic relationships. The thermodynamic consistency and accuracy of the evaluated EOS data are also discussed by comparison with the available sources.

JAEA Reports

SIMMER-III Analytic Thermophysical Property Model

Morita, Koji; Tobita, Yoshiharu; kondo, Satoru; E.A.Fischer*

JNC-TN9400 2000-004, 38 Pages, 1999/05

JNC-TN9400-2000-004.pdf:1.11MB

An analytic thermophysical property model using general function forms is developed for a reactor safety analysis code, SIMMER-III. The function forms arc designed to represent correct behavior of properties of reactor-core materials over wide temperature ranges, especially for the thermal conductivity and the viscosity near the critical point. The most up-to-date and reliable sources for uranium dioxide, mixed-oxide fuel, stainless stee1, and sodium available at present are used to determine parameters in the proposed functions. This model is also designed to be consistent with a SIMMER-III model on thermodynamic properties and equations of state for reactor-corc materials.

JAEA Reports

PNC Technical Review No.103

PNC-TN1340 97-003, 101 Pages, 1997/09

PNC-TN1340-97-003.pdf:12.06MB

no abstracts in English

JAEA Reports

PNC Technical Review No.101

PNC-TN1340 97-001, 154 Pages, 1997/03

PNC-TN1340-97-001.pdf:21.55MB

no abstracts in English

JAEA Reports

SIMMER-II Analysis of simulated core expansion experiments at purdue university

*; *

PNC-TN941 85-44, 75 Pages, 1985/03

PNC-TN941-85-44.pdf:1.83MB

In the analysis of a core expansion (or postdisassembly expansion) phase by the SIMMER-II code, it was shown that there exist various thermo-hydraulic phenomena available for mitigating effectively the mechanical energy released in a hypothetical core disruptive accident. To utilize SIMMER-II as a standard tool in future safety assessment, the experimental validation of the code is crucial especially on the energetics-mitigating effects. In this study, a series of simulated core expansion experiments performed at Purdue University was analyzed by SIMMER-II as the first effort of the code validation program in Japan. In the experiments, either the nitrogen gas at room temperature or the flashing water at high temperature was injected and expanded into the water pool simulating the outlet plenum of the reactor vessel (a 1/7-scaled model of the Clinch River Breeder Reactor vessel). In the analysis of the nitrogen expansion experiments, SIMMER-II could reproduce the experimentally measured slug impact time without adjusting input parameters. This means that the overall fluid-dynamics model of SIMMER-II is valid. In the flashing water expansion experiments, on the other hand, SIMMER did not reproduce the experimental data very well due to the presence of complex rate-limited processes including heat transfer and phase transiton. This discrepancy is ascribed to lack of modeling the entrainment phenomenon occurring at the interface of a vapor bubble. The effect of the entrainment is very important since the entrained cold liquid efficiently enhances the vapor condensation and hence reduces the slug kinetic energy. It was shown that this effect can be approximated by increasing the heat transfer coefficient between liquid components. Obviously, this result cannot be directly extrapolated to the reacor condition, but implys that the nominal SIMMER parameters are conservative from the energetics point of view because of underestimation of the vapor condensation. ...

Oral presentation

Oral presentation

Study on In-Vessel Retention (IVR) of unprotected accident for fast reactor, 4; Assessment of transition phase in ULOF

Tobita, Yoshiharu; Suzuki, Toru; Tagami, Hirotaka

no journal, , 

The event progression in the transition phase of ULOF (Unprotected Loss of Flow), which is the representative event in ATWS (Anticipated Transient without Scram) of fast breeding reactor. The existing experimental knowledges on the important phenomena, which dominates the event progression, were adopted in the nominal case. The reactivity lowered with accordance to the fuel discharge through control rod guide tube and the event terminated without prompt criticality. If the uncertainty was considered in the analysis, the reactivity slightly exceeded the prompt criticality, but no mechanical energy was produced.

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