Hiroyasu Mochizuki

Liquid metal heat transfer in heat exchangers under low flow rate conditions

The present paper describes the liquid metal heat transfer in heat exchangers under low flow rate conditions. Measured data from some experiments indicate that heat transfer coefficients of liquid metals at very low Péclet number are much lower than what are predicted by the well-known empirical relations. The cause of this phenomenon was not fully understood for many years. In the present study, one countercurrent-type heat exchanger is analyzed using three, separated countercurrent heat exchanger models: one is a heat exchanger model in the tube bank region, while the upper and lower plena are modeled as two heat exchangers with a single heat transfer tube. In all three heat exchangers, the same empirical correlation is used in the heat transfer calculation on the tube and the shell sides. The Nusselt number, as a function of the Péclet number, calculated from measured temperature and flow rate data in a 50 MW experimental facility was correctly reproduced by the calculation result, when the calculated result is processed in the same way as the experiment. Finally, it is clarified that the deviation is a superficial phenomenon which is caused by the heat transfer in the plena of the heat exchanger.

Study on Detailed Calculation and Experiment Methods of Neutronics, Fuel Materials, and Thermal Hydraulics for a Commercial Type Japanese Sodium-Cooled Fast Reactor

This paper discusses the objectives and results of a multiyear R&D project to improve the modeling accuracy for the detailed calculation of the Japanese Sodium-cooled Fast Reactor (JSFR), although the preliminary design of JSFR is prepared using conventional methods. For detailed design calculations, new methods are required because the JSFR has special features, which cannot be accurately modeled with existing codes. An example is the presence of an inner duct in the fuel assemblies. Therefore, we have developed new calculational and experimental methods in three areas: (1) for neutronics, we discuss the development of methods and codes to model advanced FBR fuel subassemblies, (2) for fuel materials, modeling and measurement of the thermal conductivity of annular fuel is discussed, and (3) for thermal hydraulics, we describe advances in modeling and calculational models for the intermediate heat exchanger and the calculational treatment of thermal stratification in the hot plenum of an FBR under low flow conditions. The new methods are discussed and the verification tests are described. In the validation test, measured data from the prototype FBR Monju is partly used.

Network Analysis of Turbine and Feedwater Systems of the ‘Fugen’ Nuclear Power Plant

The present study describes the thermal-hydraulic network analysis of the turbine and feedwater systems of the ‘Fugen’ reactor. Turbines, feedwater heaters, and corresponding piping systems are modeled using the network calculation code NETFLOW++ and thermal-hydraulic conditions are calculated using the coupled numerical model. As a result of the calculation, distributions of important characteristics of the single-phase flow and two-phase flow in the piping such as pressure and void fraction are clarified. Flow patterns in the piping were investigated using the calculated result. It was found that the state of the coolant in the drainpipe changes from saturated liquid at the inlet to a two-phase flow with a large void fraction at the connection to the feedwater heaters. This is attributed to the pressure difference between the inlet and outlet of the drainpipes. Even the drainpipe from the moisture separator to the shell of feedwater heater #4 shows a similar behavior, and the flow pattern changes from single phase to slug flow. The steam quality in the extraction line is very high, although a large number of droplets are contained in the flow. Contrary to expectation, these droplets do not completely evaporate in spite of the low-pressure conditions.

Best estimate approach for the evaluation of critical heat flux phenomenon in the boiling water reactors

Abstract Because of the uncertainties associated with the definition of Critical Heat Flux (CHF), the best estimate approach should be used. In this paper the application of best-estimate approach for the analysis of CHF phenomenon in the boiling water reactors is presented. At first, the nodalization of RBMK-1500, BWR-5 and ABWR fuel assemblies were developed using RELAP5 code. Using developed models the CHF and Critical Heat Flux Ratio (CHFR) for different types of reactors were evaluated. The calculation results of CHF were compared with the well-known experimental data for light water reactors. The uncertainty and sensitivity analysis of ABWR 8 × 8 fuel assembly CHFR calculation result was performed using the GRS (Germany) methodology with the SUSA tool. Finally, the values of Minimum Critical Power Ratio (MCPR) were calculated for RBMK-1500, BWR-5 and ABWR fuel assemblies. The paper demonstrate how, using the results of sensitivity analysis, to receive the MCPR values, which covers all uncertainties and remains best estimated.

Analysis of the EBR-II SHRT-45R Unprotected Loss of Flow Experiment with ERANOS and RELAP

This paper presents the results of the analysis of the Unprotected Loss of Flow (ULOF) experiment SHRT-45R performed in the EBR-II fast reactor. These experiments are being analyzed in the scope of a benchmark exercise coordinated by the IAEA. The SHRT-45R benchmark contains a neutronic and a thermal-hydraulic part and results are presented for both. Neutronic calculations are performed with the ERANOS2.0 code in combination with various sets of nuclear data. The thermal-hydraulic evaluation is done with RELAP5-3D. The results are that the major neutronic parameters are well predicted with error margins on the order of 1%. The thermal-hydraulic results are less favourable: a consistent overestimation of the outlet temperature occurs in combination with erroneous flow distribution. Observed differences with measured data cannot be explained easily. The work presented in this paper was undertaken to investigate and validate the effectiveness of the calculational tools and data that are commonly used in our lab for the design and analysis of liquid metal cooled fast reactors.

A thermal hydraulic calculation method of an intermediate heat exchanger of a loop type FBR

This paper describes a thermal-hydraulic calculation of an intermediate heat exchanger (IHX) with the computational fluid dynamics (CFD) code CFX. The motivation of this paper is to clarify a heat transfer degradation phenomenon in the IHX through three-dimensional calculation. The whole IHX of the “Monju” reactor is modeled with three parts, i.e., the primary side, the secondary side and the heat transfer region. Through a partial calculation using these models, the flows on the primary side and the secondary side are shown to be axisymmetric. Therefore, a sector model is adopted for the calculation model in the heat transfer region. The calculated temperatures in the IHX are compared with the measured results using the IHX in the “Monju” reactor. Good agreement is obtained for the predicted outlet temperatures and temperatures on the shell surface. As a result of the CFD calculation, it is evaluated that a heat transfer in the lower plenum on the secondary side is dominant under the low flow rate conditions. This fact contributes to lower the heat transfer coefficient in the IHX when the standard heat exchanger theory is applied to calculate the overall heat transfer coefficient between the primary and the secondary sides.