Ohshima, Hiroyuki; Tanaka, Nobuatsu*; Eguchi, Yuzuru*; Nishimura, Motohiko*; Kunugi, Tomoaki*; Uchibori, Akihiro; Ito, Kei; Sakai, Takaaki
Nippon Genshiryoku Gakkai Wabun Rombunshi, 11(4), p.316 - 328, 2012/12
It is of importance for stable operations of sodium-cooled fast reactors (SFRs) to prevent gas entrainment (GE) phenomena due to free surface vortices. Therefore, the authors have been developing an evaluation method for GE based on computational fluid dynamics (CFD) methods. In this study, we determine the suitable CFD method for GE phenomena from several candidates through some numerical benchmarks. As the results, we obtain the following guideline for the vortex-induced gas entrainment. Free vortex flow around the vortex core can be correctly evaluated by using the appropriate numerical models such as enough mesh resolution, suitable advection solver, suitable turbulence and free surface modeling.
Ohshima, Hiroyuki; Sakai, Takaaki; Kamide, Hideki; Kimura, Nobuyuki; Ezure, Toshiki; Uchibori, Akihiro; Ito, Kei; Kunugi, Tomoaki*; Okamoto, Koji*; Tanaka, Nobuatsu*; et al.
JAEA-Research 2008-049, 44 Pages, 2008/06
Japan Atomic Energy Agency has conducted a conceptional design study of a sodium-cooled fast reactor in a frame work of the FBR feasibility study. The plant system concept for a commercial step is intended to minimize a vessel diameter to achieve an economical competitiveness. Therefore, the coolant in the vessel has relatively higher velocity and gas entrainment (GE) prevention from a liquid surface in the reactor vessel becomes one of important issues for the thermal-hydraulic design. In order to establish a design criteria for the GE prevention, the GE from vortex dimples at the liquid surface was investigated by a working group. The 1st proposal of "Design Guideline for Gas Entrainment Prevention Using CFD Method" was established based on the knowledge gained from the working group activities. This report introduces each study in the working group to clarify the basis of the design guideline.
Nishimura, Motohiko*; Nonaka, Yoshiharu*; Maekawa, Isamu*
JNC-TJ9400 2005-002, 55 Pages, 2005/07
Sodium cooled FBR reactor in a feasibility study on commercialized fast reactor system has been designed so compact that flow velocity is much higher in the reactor than in the reactors so far. This design requires an evaluation method for gas entrainment from reactor free surface and effective countermeasures. Focusing on the establishment of gas entrainment evaluation method based on CFD, CFD simulation has been carried out with analyses of an existing gas entrainment experiment. In the analyses, four turbulence models such as standard k-e, RNG k-e, non-linear k-e and k-w have been applied and evaluated for their analytical capability in addition to laminar model. Main results have been obtained as follows;1) Laminar model shows faster vortices behavior accompanied with strong flow and pressure fluctuations including random variations than any other turbulence models. RNG model shows the solutions more similar to the laminar model than the other turbulence models.2) k-w, non-linear k-e and Standard k-e models show milder vortices behavior in this order due to excess turbulence viscosity.3) Main characteristics of major turbulence models have been obtained when they are applied to gas entrainment analysis.
Sato, Manabu*; Nishimura, Motohiko; Ohshima, Hiroyuki
JNC-TN9400 2003-100, 66 Pages, 2003/08
This report describes a new program for the plant dynamic response analysis of helium gas cooled fast reactors, which has the following calculation functions as point reactor kinetics, multi-dimensional reactor vessel thermal-hydraulics, volume junction gas turbine system dynamics and these interactions, and its application to the analysis of a particle-fuel-type helium gas cooled fast reactor design that was proposed in 2002 fiscal year to clarify plant dynamic characteristics under several accident conditions.
Ohshima, Hiroyuki; Nishimura, Motohiko*
JNC-TN9400 2002-072, 97 Pages, 2002/08
A feasibHity study has been carried out at JNC to construct new design concepts of commercialized fast reactors. This report describes two kinds of numerical investigations related to thermal-hydraulics of gas-cooled fast reactors of which design studies are being performed as part of the feasibility study. A series of thermal hydraulic analyses was carried out using multi-dimensional analysis program AQUA in order to confirm the heat removal capability of the helium-gas-cooled fast reactor with a coated-particle-type fuel assembly design under a rated power operation, a low power/low flow and an accident conditions. The calculation results indicates that the lateral gas flow which is indispensable for normal heat removal in the fuel particle region is kept under each calculation condition and the maximum temperature does not exceed the tentative design limitation as far as the inlet surface permeability is uniform. Only in the case of high pressure and very low flow rate, a possibility that local high temperature region exceeding the design limitation appears may not be denied due to upward flow driven by buoyancy force. Improvement on knowledge of gas property functions in high temperature and pressure drop correlations are required for more accurate analysis. Natural circulation decay heat removal characteristics of the CO gas cooled fast reactor are examined using a one-dimensional nuclear-thermal-hydraulics network analysis code, MR-X equipped with correlations for the core thermal-hydraulics of gas cooled fast reactors. Simulation parameters are the shutdown time of steam generators (SGs), restart time of gravitational water feed to the SGs, and flow rate of the feed water. It was predicted that the reactor satisfied limitation of the maximum cladding temperature under the realistic operation condition of SGs with the shutdown time of 30 s and the restart time of 20 min. Moreover, the cladding temperature still satisfied the limitation, even ...
PNC-TN9410 98-027, 40 Pages, 1998/04
To predict thermal-hydraulic phenomena in actual plant under various conditions acculately, adequate simulation of laminar-turbulent flow transition is of importance. A low Reynolds number turbulence model is commonly used for a numerical simulation of the laminar-turbulent transition. The existing low Reynolds number turbulence models generally demands very thin mesh width between a wall and a first computational node from the wall, to keep accuracy and stability of numerical analyses. There is a criterion for the distance between the wall and the first computational node in which non-dimensional distance y must be less than 0.5. Due to this criterion the suitable distance depends on Reynolds number. A liquid metal sodium is used for a coolant in first reactors therefore, Reynolds number is usually one or two order higher than that of the usual plants in which air and water are used for the work fluid. This makes the load of thermal-hydraulic numerical simulation of the liquid sodium relatively heavier. From above context, a new method is proposed for providing wall boundary condition of turbulent kinetic energy dissipation rate . The present method enables the wall-first node distance 10 times larger compared to the existing models. A function of the wall boundary condition has been constructed aided by a direct numerical simulation (DNS) data base. The method was validated through calculations of a turbulent Couette flow and a fully developed pipe flow and its laminar-turbulent transition. The predicted critical Reynolds number lay between 2300 and 2500, where commonly known value 2,320 was included. The error of the calculated Nusselt number and friction factor were comparable to uncertainties of empirical correlations. Thus the present method and modeling are capable of predicting the laminar-turbulent transition with less mesh numbers i.e. lighter computational loads.
Tanaka, Masaaki; Kobayashi, Jun; Isozaki, Tadashi; Nishimura, Motohiko; Kamide, Hideki
PNC-TN9410 98-024, 94 Pages, 1998/03
In the liquid metal cooled Fast Breeder Reactor, Local Fault incident is recognized as a key issue of the local subassembly accident. In terms of the reactor safety assessment, it is important to predict the velocity and temperature distributions not only in the fuel subassembly but also in the blockage accurately to evaluate the location of the hottest point and the maximum temperature. In this study, the experiment was performed with the 4 sub-channel geometry water test facility. Dimension is five times larger than that of a real FBR. The porous blockage is located at the center sub-channel in the test section and surrounded with three unplugged sub-channels. The blockages used in this study were (1)the solid metal, (2)the porous medium consisted of metal spheres, (3)the porous blockage with end plates covering the side or top faces of the blockage to prevent the horizontal and axial flows into the blockage. The experimental parameters were the heater output provided by the electrical heater in the simulated fuel pins and the flow rate. Temperature of the fluid was measured inside/outside the blockage and velocity profiles outside the blockage were measured. From the comparison of velocity profiles, the flow field inside the blockage depended remarkably on the blockage conditions. Such variation of flow fields affected the temperature distributions. Efficient heat transportation by horizontal flow existed in the upper part of the porous blockage. While, in the lower part of the blockage, the axial flow from the bottom face of the blockage was pre-dominated for the heat removal. Nusselt number defined by the temperature difference between the heater pin surface and the bulk temperature of the unplugged sub-channel was proportional to the power of 0.50.6 of Reynolds number. This result shows that the dependency of the Nusselt number to the Reynolds number was decided by the heat transfer from the blockage matrix to the coolant at the side of porous blockage.
Iitsuka, Toru; Oki, Yoshihisa; Kawashima, Shigeyo*; Nishimura, Motohiko; Isozaki, Tadashi; Kamide, Hideki
PNC-TN9410 98-022, 58 Pages, 1998/03
Assessment of the maximum temperature and the position of the hot spot is the most important issues on the reactor safety when the local subchannel porous blockage is occurred. From these background, authors are going to perform a sodium experiment with 37-pin bundle test rig simulating the porous blockage, to understand the phenomena and acquire data for thermal-hydraulic analysis code validation. Before the execution of sodium test, one basic experiment and some using subchannel analysis code ASFRE-III had been done. The basic experiment was a water test to examine the pressure loss characteristics of the porous blockage. The pressure loss correlation derived from the water test was applied to the subsequent subchannel analysis of the 37-pin bundle sodium test rig. The analysis such predicted that the difference between the maximum temperature and the inlet temperature would be in propotion to the power to flow rate ratio, within the condition of the power=100400 W/cm and the flow rate =200480 /min. And it was also shown that the maximum subchannel temperature would not over the operational limit temperature 650 C, if the power to flow rate ratio were kept lower than 0.75(W/cm)//min). The map was made to predict the maximum temperature from the experimental conditions.
Kimura, Nobuyuki; Nishimura, Motohiko; Momoi, K.; Hayashi, Kenji; Kamide, Hideki
PNC-TN9410 97-046, 69 Pages, 1997/04
To enhance reliability and safety of FBR, taking advantage of its inherent properties is of importance. From this point of view, natural circulation decay heat removal (NC/DHR) has been studied in which no active components such as pumps are used. DRACS (Direct Reactor Auxiliary Cooling System) is an option of NC/DHR systems. which causes cold coolant flow from DHX (Direct Heat Exchanger) penetrating into inter wrapper gaps: IWF (Inter-Wrapper Flow). Another option for NC/DHR is to use PRACS (Primary Reactor Auxiliary Cooling System) in which no remarkable IWF occurs. Thermal-hydraulic behavior in the core depends on interactions among auxiliary reactor cooling system, IHX (Intermediate Heat Exchanger), and the secondary loop during NC/DHR. Such interactions have been studied with the test rig called PLANDTL-DHX equipped with DRACS and PRACS. In this study, influence of operating condition of the auxiliary cooling systems and the secondary loop of IHX were examined on the core thermal-hydraulic bchaviors. In the present paper, one-dimensional network analyses using LEDHER code are reported. The analyses were performed on steady tests using two models: a model taking account of an inter-subassembly heat transfer only, and a model simulated both IWF and the inter-subassembly heat transfer. The calculation method was validated through comparisons with the experimental results. In the cases cooled by PRACS or IHX, two calculated models showcd good agreements with the experiments regarding the natural circulation flow rate, the temperature distribution in the core and the temperature at the inlet/outlet of the heat exchangers. However, in the case cooled by DRACS operated, the model without flow pass of IWF could not simulate the experiments with respect to the natural circulation flow rate (12% larger than experiment) and temperature profiles in the inter wrapper gaps. On the other hand, the model taking account of IWF simulated the experiments in good ...
Nishimura, Motohiko; Kamide, Hideki; Miyake, Yasuhiro*
PNC-TN9410 97-044, 55 Pages, 1997/04
Temperature distributions in fuel subassemblies of fast reactors interactively affect heat transfer from center to outer region of the core (inter-subassembly heat transfer) and cooling capability of an inter-wrapper flow, as well as maximum cladding temperature. The prediction of temperature distribution in the subassembly is, therefore one of the important issues for the reactor safety assessment. Mixing factors were applied to multi-dimensional thermal-hydraulic code AQUA to enhance the predictive capability of simulating maximum cladding temperature in the fuel subassemblies. In the previous studies, this analytical method had been validated through the calculations of the sodium experiments using driver subassembly test rig PLANDTL-DHX with 37-pin bundle and blanket subassembly test rig CCTL-CFR with 61-pin bundle. The error of the analyses were comparable to the error of instrumentation's. Thus the modeling was capable of predicting thermal-hydraulic field in the middle scale subassemblies. Before the application to large scale real subassemblies with more than 217 pins, accuracy of the analytical method have to be inspected through calculations of sodium tests in a large scale pin bundle. Therefore, computations were performed on sodium experiments in the relatively large 169-pin subassembly which had heater pins sparsely within the bundle. The analysis succeeded to predict the experimental temperature distributions. The errors of temperature rise from inlet to maximum values were reduced to half magnitudes by using mixing factors, compared to those of analyses without mixing factors. Thus the modeling is capable of predicting the large scale real subassemblies.
Momoi, K.; Hayashi, Kenji; Kamide, Hideki; Nishimura, Motohiko; Kokaki, Nobuhisa
PNC-TN9410 97-047, 93 Pages, 1997/03
The evaluation of core thermohydraulics under natural circulation conditions is of significance in order to utilize passive safety features of fast reactors. When the heat exchangers of the decay heat removal system are operated in the upper plenum of a reactor vessel, cold sodium provided by the heat exchangers can penetrate into the gap regions between fuel subassemblies; thiS natural convection phenomenon is called inter-wrapper flow (IWF). During natural circulation decay heat removal, IWF will significantly modify the flow and temperature distributions in the subassemblies. IWF can decrease the temperature in the subassemblies. On the other hand, the natural circulation head will be reduced by temperature reduction in the upper non-heated section of subassemblies due to the IWF cooling. These positive and negative effects of IWF are our main concerns in this report. Sodium experiments were carried out to investigate these phenomena. In a test section, seven subassemblies are bundled and connected to an upper plenum with a heat exchanger. The experiments were carried out under steady state conditions. Experimental parameters were power in the core and flow resistance in the primary loop. Decrease of natural circulation flows in the subassemblies were recognized. Inter-subassembly flow redistribution was also seen due to larger cooling in outer 6 subassemblies and smaller cooling in the center subassembly. In the extremely low flow conditions (large flow resistance in the primary loop), reverse flow was registered in 2 or 3 outer subassemblies. Cooling effect of IWF was also observed. It consisted of direct cooling through the wrapper tube, flow redistribution among subassemblies (higher flow rate for hotter subchannel), and cold reverse flow from the upper plenum. When the flow resistance was small in the primary loop, i.e., flow rate was larger than 1% of reactor rated conditions (based on subassembly averaged flow velocity), the cooling effects and ...
Hayashi, Kenji; Momoi, K.; Nishimura, Motohiko; Kamide, Hideki
PNC-TN9410 97-045, 68 Pages, 1997/03
Steady state sodium experiments were performed to investigate interactions between the heat transport systems, i.c., the primary system, the secondary system, and the decay heat removal system, during the natural circulation decay heat removal in FBRs. The PLANDTL-DHX test rig was used for the experiments. The core model has seven subassemblies; the center assembly simulates pin bundle geometry of a core fuel subassembly in a large scale FBR and consists of 37 pins, six outer subassemblies consists of 7 pins. As the decay heat removal system, Dirtct Reactor Auxiliary Cooling System (DRACS) and Primary Reactor Auxiliary Cooling System (PRACS) can be selected. Experiments were carried out under natural circulation conditions in the primary loop and force convection conditions in the decay heat removal system. In cases using DRACS, natural circulation flow rate in the primary loop was smaller by 20% than that in cases using PRACS due to the low temperature in the upper plenum and also in the upper non-heated section of the core. When natural circulation was allowed in the secondary heat transport system, the natural circulation flow rate in the primary system increased in spite of the operation of DRACS. In cases using DRACS, inter-subassembly flow redistribution occurred; the center subassembly had larger flow rate than those in outer subassemblies due to the low natural circulation head in the outer subassemblies which were cooled by the inter-wrapper flow (IWF). The highest temperature in the core was reduced by IWF via not only the direct cooling effect but also the inter-subassembly flow redistribution. Temperature fluctuations around the PRACS cooling coil installed in the IHX were registered under the natural circulation conditions in the primary system. The amplitude of fluctuation was less than 20C and small on the points of structural integrity.
Nishimura, Motohiko; Kamide, Hideki; Ohshima, Hiroyuki
PNC-TN9410 96-289, 158 Pages, 1996/10
Temperature distributions in fuel subassemblies of fast reactors interactively affect heat transfer from center to outer region of the core (inter-subassembly heat transfer) and cooling capability of an inter-wrapper flow, as well as maximum cladding temperature, The prediction of temperature distribution in the sub-assembly is, therefore one of the important issues for the reactor safety assessment. To treat the complex phenomena in the core, a multi-dimensional thermal hydraulic analysis is the most promising method. From the studies on the multi-dimensional thermal hydraulic modeling for the fuel sub-assemblies, the followings have been recommended through the analysis of sodium experiments using driver subassembly test rig PLANDTL-DHX and blanket subassembly test rig CCTL-CFR. (1)Staggered half pin mesh arrangements for the pin bundles. (2)MIT's distributed resistance correiation for the predictions of axial pressure drop. (3)MIT's Mixing factors for inter-subchannel mixing due to wire sweeps, turbulence and thermal plumes. Computations of steady states experiments in the test rigs using the above modeling showed quite good agreement to the experimental data. In the present study, the use of this modeling was extended to transient analyses, and its applicability was examined, Firstly, non-dimensional parameters used to determine the mixing factors were modified from the ones based on bundle-averaged values to the ones by local values. Sccondly, a new threshold function was derived and introduced to cut off the mixing factor of thermal plumes under inertia force dominant conditions. This function goes 1 to 0 when Richardson number decreasingly passes through value of 0.1. Since no thermal plumes were observed under the condition below Richardson numbcr equal 0.1, in the results of existing studies. A comparison or the multi-dimensional analysis to a subchannel analysis was also made to cxamine the consistency of formulations between the two analytical ...
Momoi, K.; Hayashi, Kenji; Nishimura, Motohiko; Kamide, Hideki
PNC-TN9410 96-280, 146 Pages, 1996/10
The complicated thermal hydraulics which was not observed in forced circulation occurs in the core during natural circulation decay heat removal in Fast Breeder Reactors (FBRs). Especially, in a FBR which has the auxiliary cooling system of DRACS (Direct Reactor Auxiliary Cooling System) type with direct heat exchangers (DHXs) immersed in the reactor plenum, cold sodium provided by the DHXs penetrates into inter-wrapper gaps in the core. This phenomena called Inter-wrapper Flow (IWF) will influence the thermal hydraulics in the core. Further, thermal-interaction between the cooling systems makes the natural circulation flow rate in the primary loop change and will influence the thermal hydraulics in the core. The purpose of this study is to grasp the interaction between the cooling systems and its influence on the thermal hydraulics in the core during natural circulation decay heat removal in FBRs. Transient Sodium experiments which simulated transitions from forced to natural circulation in reactors were performed under several kinds of operating conditions of the IHX secondary system and of the auxiliary cooling system using PLANDTL-DHX test facility with DRACS and PRACS (Primary Reactor Auxiliary Cooling System). In the DRACS, the heat removed by natural circulation in the IHX secondary system influenced the natural circulation flow rate in the primary loop more than the operation conditions of DRACS. When natural circulation in the IHX secondary loop was stopped, the IHX outlet temperature rapidly increased and the natural circulation flow rate in the primary loop decreased. In addition, reverse flows were detected in the outer subassemblies. The temperature rise in the heated length of the center subassembly was reduced by the cooling effect of the inter-subassembly heat transfer to the reverse flow subassemblies and IWF. When the natural circulation in the IHX secondary loop was continued, the natural circulation flow rate of 1% level of rated ...
Kamide, Hideki; Nishimura, Motohiko; Hayashi, Kenji; Momoi, K.; Miyake, Yasuhiro*
PNC-TN9410 96-268, 79 Pages, 1996/09
It is considerably effective to utilize the natural circulation on advance of reliability of the decay heat removal systems of Fast Breeder Reactors. The natural circulation dose not depend on components which needs external power sources like pumps. This increases the reliability of the decay heat removal systems. However, thermohydraulics in the core have complex characteristics under the natural circulation. Under low flow conditions, buoyancy effects and heat transfer from high temperature subassemblies to low temperature subassemblies, i.e. inter-subassembly heat transfer, will significantly modify the flow and temperature distributions in the subassemblies. Thus, development of an evaluation method for the core thermohydraulic is significant to utilize the natural circulation. A multi-subassembly analysis method using the three-dimensional thermohydraulic analysis code, AQUA, was developed to predict thermohydraulics in the subassemblies with influence of the inter-subassembly heat transfer. In this method, each subassembly is modeled in individual mesh region of multi-region model of AQUA and the staggered half-pin mesh arrangement was applied in each subassembly. The heat transfer between the subassemblies was simulated by a thermal structure model. The analysis method was applied to two sodium experiments where three or seven subassemblies were modeled in simulated cores. The experimental analyses showed that the multi-subassembly analysis method could evaluate the thermohydraulics in the subassemblies.
Kamide, Hideki; Hayashi, Kenji; Gunji, Minoru; Hayashida, Hitoshi; Nishimura, Motohiko; Iitsuka, Toru; Kimura, Nobuyuki; Tanaka, Masaaki; Nakai, Satoru; Mochizuki, Hiroyasu; et al.
PNC-TN9410 96-279, 51 Pages, 1996/08
Large-scaled thermohydraulic tests are planned for some new key technologies in the heat transport systems of demonstration fast reactors, in which the reactor vessel, the primary system, the secondary system, water-steam system, and the decay heat removal systems are modeled. Thermohydraulic issues and structural integrity issues were discussed for the top entry piping systems with satellite pools of the intermediate heat exchangers and the pumps, the natural circulation decay heat removal using direct heat exchangers in a reactor hot pool, the reactor vessel wall cooling system, and the new type of steam generators in the demonstration reactor. Concepts of the experimental model for the reactor vessel and the primary system were created and compared with each other for the sodium test facility which enables to answer the thermohydraulic and structural integrity issues. Following items were considered in the creation and in the selection of the models; (1)solution of the issues for Demonstration First Reactor on total system characteristics, the reactor vessel wall cooling system, the decay heat removal system, and the steam generator, (2)balance between the thermohydraulic issues and the structural integrity issues, (3)simulations of compound phenomena and interactions between the components and the heat transport systems. Total system of test facility was specified based on the selected test model.