[en] A heat transfer package was developed for the core thermal-hydraulic design and analysis of the Japan Research Reactor-3 (JRR-3) which is to be remodeled to a 20 MWt pool-type, light water-cooled reactor with 20 % low enriched uranium (LEU) plate-type fuel. This paper presents the constitution of the developed heat transfer package and the applicability of the heat transfer correlations adopted in it, based on the heat transfer experiments in which thermal-hydraulic features of the new JRR-3 core were properly reflected. (author)

[en] JRR-3 is a light water moderated and cooled, beryllium and heavy water reflected pool type research reactor using low enriched uranium (LEU) plate-type fuels. Its thermal power is 20 MW. The core conversion program from uranium-aluminum (UAl_x-Al) dispersion type fuel (aluminide fuel) to uranium-silicon-aluminum (U₃Si₂-Al) dispersion type fuel (silicide fuel) is currently conducted at the JRR-3. This report describes about the steady-state thermal hydraulic analysis results and the flow channel blockage accident analysis result. In JRR-3, there are two operation mode. One is high power operation mode up to 20 MW, under forced convection cooling using the primary and the secondary cooling systems. The other is low power operation mode up to 200 kW, under natural circulation cooling between the reactor core and the reactor pool without the primary and the secondary cooling systems. For the analysis of the flow channel blockage accident, COOLOD code was used. On the other hand, steady-state thermal hydraulic analysis for both of the high power operation mode under forced convection cooling and low power operation under natural convection cooling, COOLOD-N2 code was used. From steady-state thermal hydraulic analysis results of both forced and natural convection cooling, fuel temperature, minimum DNBR etc. meet the design criteria and JRR-3 LEU silicide core has enough safety margin under normal operation conditions. Furthermore, flow channel blockage accident analysis results show that one channel flow blockage accident meet the safety criteria for accident conditions which have been established for JRR-3 LEU silicide core. (author)

[en] The Multiple Purpose Research Reactor MEX-15 is a 15 MW thermal power, swimming pool type research reactor and will be constructed by the National Institute of Nuclear Research (ININ) of Mexico. Demineralized light water will be used as coolant and moderator. The reactor core will be surrounding by graphite reflectors. The reactor will use 19.75% enriched U₃O₈-A1 plate-type fuel (MTR-type). The core thermal-hydraulic conceptual design of the MEX-15 was performed for two cooling modes, forced convection cooling and natural convection cooling. The key criteria are first to avoid the nucleate boiling anywhere in the core and second to have enough safety margin to the DNB for normal operation conditions. The results of the thermalhydraulic conceptual design and analysis show that the optimum coolant velocity in the standard fuel element is about 5.6 m/s with the minimum temperature margin against the ONB temperature of about 17degC and the minimum DNBR of 2.58 for the forced-convection cooling mode at a core power of 15 MW with the pressures of 1.43 kg/cm² at the core inlet and a core inlet coolant temperature of 35degC. It was also determined that the total core power up to about 300 kW can be removed by the natural convection cooling under the condition that nucleate boiling is not allowed anywhere in the core. The minimum temperature margin against ONB temperature and the minimum DNBR at 300 kW are 1.6degC and 6.31, respectively. The results obtained in this work establishes the preliminary technical specifications for the core thermal-hydraulic design of the Multiple Purpose Research Reactor MEX-15. (author)

[en] A mercury target is used in the spallation neutron source driven by a high-intensity proton accelerator. In this study, the effectiveness of the cross-flow type mercury target structure was evaluated experimentally and analytically. Prior to the experiment, the mercury flow field and the temperature distribution in the target container were analyzed assuming a proton beam energy and power of 1.5 GeV and 5 MW, respectively, and the feasibility of the cross-flow type target was evaluated. Then the average water flow velocity field in the target mock-up model, which was fabricated from Plexiglass for a water experiment, was measured at room temperature using the PIV technique. Water flow analyses were conducted and the analytical results were compared with the experimental results. The experimental results showed that the cross-flow could be realized in most of the proton beam path area and the analytical result of the water flow velocity field showed good correspondence to the experimental results in the case when the Reynolds number was more than 4.83x10⁵ at the model inlet. With these results, the effectiveness of the cross-flow type mercury target structure and the present analysis code system was demonstrated

[en] The COOLOD-N code provides a capability for the analysis of the steady-state thermal-hydraulics of research reactors in which plate-type fuel is employed. This code is revised version of the COOLOD code, and is applicable not only to a forced convection cooling mode, but also to a natural convection cooling mode. In the code, a function to calculate flow rate under a natural convection, and a heat transfer package which was a subroutine program to calculate heat transfer coefficient, ONB temperature and DNB heat flux, and was especially developed for the upgraded JRR-3, have been newly added to the COOLOD code. The COOLOD-N code also has a capability of calculating the heat flux at onset of flow instability as well as DNB heat flux. (author)

[en] In research reactors, plate-type fuel elements are generally adopted so as to produce high power densities and are cooled by a downward flow. A core flow reversal from a steady-state forced downward flow to an upward flow due to natural convection should occur during operational transients such as Loss of the primary coolant flow'. Therefore, in the thermal hydraulic design of research reactors, critical heat flux (CHF) under a counter-current flow limitation (CCFL) or a flooding condition are important to determine safety margins of fuel against CHF during a core flow reversal. The authors have proposed a CHF correlation scheme for the thermal hydraulic design of research reactors, based on CHF experiments for both upward and downward flows including CCFL condition. When the CHF correlation scheme was proposed, a subcooling effect for CHF correlation under CCFL condition had not been considered because of a conservative evaluation and a lack of enough CHF data to determine the subcooling effect on CHF. A too conservative evaluation is not appropriate for the design of research reactors because of construction costs etc. Also, conservativeness of the design must be determined precisely. In this study, therefore, the subcooling effect on CHF under the CCFL conditions in vertical rectangular channels heated from both sides were investigated quantitatively based on CHF experimental results obtained under uniform and non-uniform heat flux conditions. As a result, it was made clear that CHF in this region increase linearly with an increase of the channel inlet subcooling and a new CHF correlation including the effect of channel inlet subcooling was proposed. The new correlation could be adopted under the conditions of the atmospheric pressure, the inlet subcooling less than 78K, the channel gap size between 2.25 to 5.0mm, the axial peaking factor between 1.0 to 1.6 and L/De between 71 to 174 which were the ranges investigated in this study. (author)

[en] The COOLOD-N2 code provides a capability for the analyses of the steady-state thermal-hydraulics of research reactors. This code is revised version of the COOLOD-N code, and is applicable not only for research reactors in which plate-type fuel is adopted, but also for research reactors in which rod-type fuel is adopted. In the code, subroutines to calculate temperature distribution in rod-type fuel have been newly added to the COOLOD-N code. The COOLOD-N2 code can calculate fuel temperatures under both forced convection cooling mode and natural convection cooling mode as well as COOLOD-N code. In the COOLOD-N2 code, a 'Heat Transfer package' is used for calculating heat transfer coefficient, DNB heat flux etc. The 'Heat Transfer package' is subroutine program and is especially developed for research reactors in which plate-type fuel is adopted. In case of rod-type fuel, DNB heat flux is calculated by both the 'Heat Transfer package' and Lund DNB heat flux correlation which is popular for TRIGA reactor. The COOLOD-N2 code also has a capability of calculating ONB temperature, the heat flux at onset of flow instability as well as DNB heat flux. (author)

[en] This report describes the results about thermo-hydraulic behavior in the accident of flow blockage to coolant channels of upgraded JRR-3. Analysis was carried out using EUREKA-2 code. Flow blockage to coolant channels accident occur by some extraneous things which come from outside of the reactor pool, may block the coolant flow channels of the core. If flow blockage to coolant channels would occur, fuel temperature will increase due to flow rate decrease of coolant channels. And at last, fission products will be released from inside of fuel plates to the primary cooling system due to failure of fuel plates. In the analysis, one standard type fuel element was supposed as flow blockage channels, in the same way sa one of credible accidents, which postulated in the JRR-3 safety assessment. From the results, it was shown that about 16.7 % of the fuel element which was supposed as flow blockage channels, would fail, assuming that fuel plates might fail when the fuel meat temperatures riseover 400 deg C. (author)

[en] JRR-3 is a light water moderated and cooled, beryllium and heavy water reflected pool type research reactor using low enriched uranium (LEU) plate-type fuels. Its thermal power is 20 MW. The core conversion program from uranium-aluminum (UAl_x-Al) dispersion type fuel (aluminide fuel) to uranium-silicon-aluminum (U₃Si₂-Al) dispersion type fuel (silicide fuel) is currently conducted at the JRR-3. This report describes about reactivity initiated events analysis for the safety assessment of JRR-3 silicide core which have been carried out as a part of JRR-3 silicide fuel project. The following five cases for the anticipated operational transients have been selected and analyzed for the safety assessment. (1) Uncontrolled control rod withdrawal from zero power, (2) Uncontrolled control rod withdrawal during steady-state operation, (3) Reactivity insertion by removal of irradiation samples, (4) Reactivity insertion by cold water insertion, (5) Reactivity insertion by light water insertion to heavy water reflector, All analyses have been carried out by a point kinetics computer code EUREKA-2. The results show that all cases meet the safety criteria for anticipated operational transients which have been established for the JRR-3 silicide fueled core. (author)

[en] A detailed understanding of critical heat flux (CHF) for vertical rectangular channels is required for the thermohydraulic design and safety analysis of research nuclear reactors in which flat-plate-type fuel is employed. In this study differences in CHF between upflow and downflow were investigated, focussing especially on channel outlet subcooling, because differences in CHF between upflow and downflow under the subcooled condition at the channel exit had not been fully clarified systematically. With the investigation of the existing CHF data, it was elucidated that there was a systematic tendency in CHF between upflow and downflow under the subcooled condition at the channel exit with dimensionless subcooling and coolant mass flux, and a new CHF correlation was proposed for the vertical rectangular channels based on investigation results. The condition that CHF differs between downflow and upflow was also made clear within the range investigated in this study with pressure of less than 4 MPa and dimensionless mass flux of less than 2200. (author)