Shinichi Morooka

Void fraction estimation within rod bundles based on three-fluid model and comparison with x-ray CT void data

An interfacial shear stress equation in the dispersed-annular two-phase flow regime has been developed, which is based on a three-fluid model consisting of a liquid film on a rod, vapor and entrained liquid associated with a vapor flow. It is an extension of J.G.M. Andersens procedure that provides a two-fluid interfacial shear stress equation using the drift flux parameters C0 and Vgj. This interfacial shear stress equation can take into account a phase and velocity distribution through an equivalence between the drift flux parameters and the interfacial shear stress. Using the three-fluid subchannel analysis code TEMPO with the three-fluid interfacial shear stress model, the capability of a three-fluid calculation using the drift flux parameters C0 and Vgj that reproduce a measured void fraction is demonstrated. A comparison was made with advanced X-ray computed tomography (CT) void fraction data within a 4×4 rod bundle in diabatic 1 MPa pressure conditions. The three-fluid velocity field was estimated to be in good agreement with the experimental result of a void fraction.

An experimental study on rewetting phenomena in transient conditions of bwrs

Abstract It is known that rod temperature rise after boiling transition (BT) is not excursive and that the peak cladding temperature (PCT) is suppressed by rewetting to return to nucleate boiling, even if BT occurs under severe conditions exceeding abnormal operational transients for a BWR. The purpose of this study is to develop and verify the rewetting correlation. The rewetting correlation was developed based on single rod data, as a function of quality, mass flux, pressure and heat flux. The transient thermal-hydraulic code used in the BWR design analysis (SCAT) with this rewetting correlation was compared with transient rod temperature result after the occurence of BT obtianed by the 8×8 and 4×4 rod bundle. It is concluded that the transient code with the developed rewetting correlation predicts the PCT conservatively, and the rewetting time well.

Analytical study on effects of BWR fuel spacer on droplet deposition

In a boiling water nuclear reactor (BWR), liquid film dryout may occur on a fuel rod surface when the fuel assembly power exceeds the critical power. The spacers supporting fuel rods affect on the thermal-hydraulic performance of the fuel assembly. The spacer is designed to enhance critical power significantly. If spacer effects for two-phase flow could be estimated analytically, the cost and time for the development of the advanced BWR fuel would be certainly decreased. The final goal of this study is to be able to analytically predict the critical power of a new BWR fuel assembly without any thermal-hydraulic tests. Initially, we developed the finite element code to estimate spacer effects on the droplet deposition. Then, using the developed code, the spacer effects were estimated for various spacer geometries in a plane channel and one subchannel of BWR fuel bundle. The estimated results of the spacer effects showed a possibility to analytically predict the critical power of a BWR fuel assembly.

Study on a heat pipe

Abstract The objective of this study was to obtain an understanding of heat pipe operating limits. Sintered powders were used as the wick, and pure water and Freon 113 as the working fluid. In this study, two types of experiments were undertaken. The first involved independent studies of wick characteristics, friction losses and capillary properties. The second involved the measurement of maximum heat transfer rates. The simplified model was developed for predicting the maximum heat transfer rates of capillary limits. The agreement between predicted and experimental maximum heat-transfer rates was excellent.

374. 燃料集合体信頼性実証試験―BWR燃料集合体熱水力試験

Nuclear Power Engineering Corporation (NUPEC) has conducted a proving test for thermal-hydraulic performance of BWR fuel (high-burnup 8×8, 9×9) assemblies entrusted by the Ministry of Economy, Trade and Industry (NUPECTH-B Project). The high-burnup 8×8 fuel (average fuel assembly discharge burnup: about 39.5GWd/t), has been utilized from 1991. And the 9×9 fuel (average fuel assembly discharge burnup: about 45GWd/t), has started to be used since 1999. There are two types (A-type and B-type) of fuel design in 9×9 fuel assembly. Using an electrically heated test assembly which simulated a BWR fuel bundle on full scale, flow induced vibration,

Study on post-BT heat transfer in a full scale BWR (8×8) rod bundle

Post-BT heat transfer experiments on a full scale BWR (8 × 8) rod bundle were conducted at the Nuclear Power Engineering Test Center (NUPEC), which is sponsored by the Ministry of International Trade and Industry (MITI). Demineralized water was used as a test fluid. The experimental conditions were: Mass flux: 284 to 1562 kg/m2s, Pressure: 7.15 MPa, Inlet subcooling: 50 kJ/kg. Test results indicate that the rod temperature rise after boiling transition (BT) is not “excursive”. Comparisons with several existing post-BT correlations indicate that the Groeneveld, The Dougall-Rohsenow and the Condie-Bengston correlations tend to be conservative in the intermediate range before the onset of fully developed film boiling, while the Koizumi and the Sugawara correlations predict well over the complete Post-BT range.

Recent activities in the field of thermal hydraulic researches

In the field of thermal hydraulics, substantial progress has been made in research on single and two-phase heat transfer. Grid-enhanced convection heat transfer has been studied [1,2]. Moon et al. [1] performed an experimental study in a 6 × 6 rod bundle to investigate the effects of spacer grids on the single-phase convection heat transfer enhancement. The experimental data showed that the Reynolds number has a significant impact on the heat transfer enhancement only when the Reynolds numbers are lower than about 10,000. They suggested more systematic experiments should be performed using various spacer rids with large blockage ratios at low Reynolds numbers, considering an early phase of the re-flood conditions.Miller et al. [2] reported a two-phase dispersed droplet flow investigation of the grid-enhanced heat transfer augmentation using a 7 × 7 rod bundle heat transfer facility. It was found that a second-stage augmentation occurs under wet grid conditions at a distance of 10 diameters downstream of the grid. This second-stage augmentation was not observed under dry-grid conditions, nor was it observed in single-phase steam cooling tests [2]. Schlegel et al. [3] performed extensive experiments in pipes with diameters up to 0.304 m to collect area-averaged void fraction data using electrical impedance void meters for the purpose of remedying an inability of current drift-flux models to accurately predict the void fraction in churn-turbulent flows in large diameter pipes. They obtained a distribution parameter modified for churn-turbulent flows. It has been evaluated through comparison of the void fraction predicted by the drift-flux model and the measured void fraction. Experimental data bases are important for models’ assessment and verification. Heat transfer and flow experiments using a mercury flow system were carried out by Kinoshita et al. [4] to clarify the validity and predictability of existing experimental correlations. They obtained a result that the heat transfer coefficients agreed well with the Subbotin correlation and analytical results with the STAR-CD code. Conner et al. [5] reported hydraulic benchmark data on Westinghouse PWR mixing vane grids at Texas A&M University. The data acquisition of interest is from an advanced particle image velocimetry (PIV) technique which can attain the high spatial and temporal resolution of the velocity vectors. The data obtained provided amuchmore thorough benchmark of computational fluid dynamics (CFD) results than were available before. Use of this data can not only help in benchmarking steady-state CFD simulations, but can also be used in benchmarking transient CFD simulations such as large eddy simulation [5]. To assess the safety at nuclear facilities and to respond to emergencies against accidental or intentional release of radioactive materials, a LOcal-scale Highresolution atmospheric DIspersion Model using LargeEddy simulation (LOHDIM-LES) has been developed by Nakayama et al. [6]. It was extended to turbulent flows and plume dispersion in various building arrays, and successfully simulated the unsteady behaviors of turbulent flows and plume dispersion in urban-type surface geometries. The CUPID code and TAPINS code were also developed for the analyses of transient twophase flows in nuclear reactor components and transient analysis of an integral reactor, REX-10 [7,8]. Lee and Park [8] compared the calculation results of TAPINS with the experimental data obtained from a series of integral effects tests using a scaled apparatus of REX-10. It was concluded that TAPINS can provide the reasonable prediction on the thermal-hydraulic responses of REX-10 during the transient and accident conditions [8].Moreover, Onder andLeung [9] evaluated the ASSERT-PV subchannel code using boiling-lengthaverage (BLA) critical heat flux (CHF) values for the CANFLEX bundle at cross-sectional average subcooled conditions. Severe accident analyses are also carried out. Kawahara et al. [10] proposed a method for identifying the success criteria regarding alternative water injection in long-term station blackout (SBO) of a BWR5 model plant by summarizing the sensitivity analysis results using RELAP5/SCDAP mod 3.5. They found that preventing core damage was almost equivalent to

Application of Probabilistic Safety Analysis in Design and Maintenance of the ABWR

A brief history of the development of nuclear reactor in Japan is summarized in Fig. 1.1s. In the 1960s, nuclear reactor technology was introduced mainly from the United States. But in this era, the capacity factor of Japanese boiling water reactors (BWRs) is low because of initial problems such as stress corrosion cracking (SCC). A program to improve the nuclear reactor performance was started. In the 1970s, phases-I and -II of this program was carried out for the purpose of improvement, standardization, and localization of conventional light water reactors (LWRs). The final stage of this program was carried out in the 1980s to develop advanced reactors (both ABWR and APWR), which had to meet the following objectives.

Study on Condensation Heat Transfer Under High-Temperature, High-Pressure Conditions

Knowing the predicted overpressure rate under anticipated operational occurrences (AOOs) is very important when evaluating the integrity of a BWR reactor pressure vessel. One of the factors that influence the overpressure rate is the wall condensing performance. Many condensing studies have been done under low-pressure conditions without vapor flow, but few condensing test results under BWR conditions have been reported. Therefore, the purposes of this study were to extend the vapor condensing data base for high-pressure, high-temperature conditions, to improve the heat transfer correlation and to evaluate the condensing effect on the overpressure rate. Condensation heat transfer tests have been performed with the pressure range from 0.5 to 8 MPa under upward and downward vapor flow. The test section consisted of a condensing tube and a water-cooling jacket. The condensing tube was a circular tube. The test results showed that the condensation heat transfer coefficient increased with the velocity of vapor flow due to enhancement of heat transfer caused by turbulence of the liquid film. We obtained a new correlation for condensation heat transfer that considered vapor shear force and condensate film Reynolds number. This new correlation agreed well with experimental data over a wide range of pressure. New correlation was incorporated into TRACG02modT1 code. When the condensation heat transfer tests were analyzed using this modified TRACG02modT1 code, the calculated condensation heat transfer coefficients were found to be in considerable agreement with the measured data. Furthermore, when the main steam isolation valve AOO (safety relief valve capacity design) of the BWR plant was evaluated by this modified TRACG02modT1 code, we found that the vapor condensation effect appeared under relatively high-pressure conditions and the pressure with improved condensation model was lower than that without vapor condensation. In summary, the condensation heat transfer model of TRACG02modT1 code has been improved based on high-pressure, high-temperature condensation test data with vapor flow. The vapor condensation effect was found to be strong, especially in the pressure increase AOO of the actual plant.Copyright

Application of the Post-Boiling-Transition Standard to Licensing Analysis

Abstract To ensure fuel integrity, light water reactor cores are designed to avoid the onset of boiling transition (BT) inside the fuel assembly that leads to a deterioration of the heat transfer characteristics and subsequent excessive rise of the fuel-cladding temperature in the anticipated operational occurrences (AOOs). However, some boiling water reactors’ AOO events result in immediate scram or suppression of the reactor power due to an increase in the reactor coolant void fraction. Recent studies show that a short duration of dryout inside the fuel assembly only leads to a small rise in the fuel-cladding temperature and thus does not pose a threat to fuel integrity. Many tests on BT and an improved comprehension of its mechanism have led to the development of a methodology to appropriately assess the fuel-cladding temperature after BT has been reached. The Standards Committee of the Atomic Energy Society of Japan has therefore proposed a cladding temperature criterion after BT. Applying the post-BT standard enables the value of the operating limit minimum critical power ratio (OLMCPR) to be decreased by allowing for a short duration of dryout. We calculated the fuel-cladding temperature and dryout duration in the load rejection condition without a bypass event. The calculated results show that both the fuel-cladding temperature and dryout duration meet the post-BT standard in the case of a small OLMCPR, which is determined by the loss of feedwater heating. This enables a more efficient reactor core to be designed by applying the post-BT standard to licensing analysis. The possibility of applying a post-BT standard is demonstrated from the results of this work.