Yoshio Murao

Scale effects on countercurrent gas-liquid flow in a horizontal tube connected to an inclined riser

The scale effects of a flow path under countercurrent flow limitation (CCFL) (air/water or steam/saturated water) in a horizontal tube connected to an inclined riser have been studied. The studied geometry simulates that of a PWR hot leg. An analytical model with a two-fluid model was developed based on flow observation results in small scale experiments and then assessed with various scale experiments under various pressures to evaluate the scale effects. The assessments with small scale experiments revealed that the region controlling the flow limitation was shifted from the horizontal tube near the bend to the inclined riser part as the length of inclined riser increased. The degree of the shift became weaker for larger scale experiments under the assumption that the ratio of interfacial friction factor to wall-to-gas friction factor was maintained to be the same as that for small scale experiments. The degree of the shift was not affected by the change of pressure (0.3 MPa → 1.5 MPa).

Experimental Modeling of Core Hydrodynamics during Reflood Phase of LOCA

The over-all core hydrodynamics during the reflood phase is quantitatively discussed. The following model was proposed for the core hydrodynamics: (1) When the liquid mass flux and the steam mass flux exist at the quench front, dispersed flow appears. (2) The void fraction downstream of the quench front gradually decreases with time. (3) Finally, the void fraction approaches the value predicted by the modified Cunningham & Yeh and the modified Lockhart-Martinelli correlations within about ±30% error. (4) Even after the concerning region is quenched, the void fraction can be predicted with the same correlations. Both correlations have been recently developed.

A concept and safety characteristics of JAERI passive safety reactor (JPSR)

Abstract A JAERI passive safety reactor (JPSR) system has been developed for reduction of manpower in operation and maintenance and increase of safety. The major features are as follows; an inherent matching nature of core heat generation and heat removal, in-vessel control rod drive mechanism units, once-through steam generators, a large volume pressurizer, passive heat removal system, and passive safety injection system. The passive heat removal system consists of residual heat exchangers, gravity coolant injection pools and air coolers. Steady-state analysis clarified that core residual heat during the whole period from shutdown to refueling phase can be transferred by natural circulation to the gravity coolant injection pool and finally to the atmosphere. In case of a large break loss-of-coolant accident (LOCA), steam is discharged and condensed in the gravity coolant injection pool. It was evaluated that the pool water temperature remains below boiling temperature under the large break LOCA condition. In the course of the design study, probabilistic safety assessment (PSA) was used to clarify the safety features and weak points. The results indicated that loss of offsite power is the dominant initiating event. Based on the analyses, improvements of safety system design were proposed.

Analytical Study of Thermo-Hydrodynamic Behavior of Reflood-Phase during LOCA

The objective of this study is the establishment of the thermo-hydrodynamic model of the reactor core during reflood phase of LOCA. Based on the quench model proposed by the author, and assuming a reflood model including a flow model and a set of the thermo-hydrodynamic correlations, a reflood analysis code named “REFLA-1D” was developed. Considerably close agreement between PWR-FLECHT tests and the results calculated by REFLA-1D code for the critical Weber number Wec= 1 was obtained for fuel clad temperature histories and the quench time and the quench temperature except for the quenching from the top of the fuel rod. It was found that the errors of calculated quench time and temperature are within ±20% under the following conditions: (1) pressure 4.5–1.5 kg/cm2·a or core inlet velocity 15–4.8 cm/s, (2) inlet subcooling more than 30°C. In the transition flow region, the calculated tendency of the temperature histories is different from the measured. This reflood model appears to be reasonable but some mo...

Critical heat flux experiments under steady-state and transient conditions and visualization of CHF phenomenon with neutron radiography

Abstract Steady-state and transient critical heat flux (CHF) experiments were performed using triangular-pitch 7-rod assemblies with non-uniform power distributions under the maximum pressure of 15.5 MPa. The onset of steady-state CHF was predicted within an uncertainty of 10% with the KfK correlation using local flow conditions calculated by the subchannel analysis code COBRA-IV-I. However, existing mechanistic CHF models did not agree with the steady-state CHF data. The transient CHFs under the conditions of flow reduction, power increase or flow and power simultaneous variation were predicted with the quasi-steady-state method within approximately the same uncertainty as the steady-state cases. In order to clarify the CHF phenomenon, a real-time neutron radiography technique was used to visualize boiling flow inside a stainless-steel shroud under a pressure of 1.0 MPa. Various two-phase flow patterns and flow behaviors just before and after the onset of CHF were observed by this method.

SCTF-III test plan and recent SCTF-III test results☆

Abstract A test plan of the Slab Core Test Facility with Core-III (SCTF-III) has been clarified. The previous SCTF-III tests simulating PWRs with combined-injection-type ECCS indicated the following results on the thermal-hydraulics in the full-radius core: 1. (1) two-region separation and multidimensional thermal-hydraulics both before and after reflood initiation. 2. (2) practically no core cooling in the region far from the water downflow region before reflood initiation in spite of good core cooling in the water downflow region. 3. (3) good core cooling even in the two-phase upflow region after bottom reflood initiation. 4. (4) large fall-back flow rate in comparison with the prediction by a typical one-dimensional correlation.

Experimental study of effects of upward steam flow rate on quench propagation by falling water film.

Emergency core cooling (ECC) mater is carried up to the upper plenum and falls down again into the core during the reflood phase in PWR-LOCA. Therefore the quench front also propagates downward from the top of the core. The effect of upward steam flow rate on the top-down quench propagation was experimentally investigated. It was found that top-down quench velocity was delayed by upward steam flow. This effect is more significant when rod surface temperature is low and the falling water flow rate is small. The effect of the flow rate and the rod temperature on the quench velocity was correlated based on the experimental results under the conditions of atmospheric pressure, saturation temperature for water and steam, rod surface temperature of 350–600°C, down-ward water velocity of 0.01–0.1 m/s and upward steam velocity of 0–20 m/s.

Large-scale multi-dimensional phenomena found in CCTF and SCTF experiments

Abstract Thermal-hydraulic behavior in the pressure vessel during the reflood phase of a PWR-LOCA is discussed based on the results from tests with Cylindrical Core and Slab Core Test Facilities, which model a 1100-MWe-class PWR with a scaling ratio of about 1/20. Major findings on core thermal-hydraulics are: (i) substantial water accumulation in the upper part of the core, (ii) radially uniform water accumulation, and (iii) fluid circulation and/or concentration to high power bundles. These multi-dimensional phenomena cause better core cooling in the high power bundle than expected from an evaluation model for licensing based on one-dimensional reflood experiments. After comparison with data from the FLECHT low-flooding tests the substantial water accumulation is considered to be caused by recirculation flow. Based on small scale reflood experiments, effect of local mass flow on heat transfer enhancement in film boiling was correlated and all phenomena except the flow concentration effect were numerically well-simulated by using an analytical model with this correlation. The flow concentration effect on core heat transfer was empirically correlated with local power ratio. It was observed that the core coolability was enhanced by non-uniform fluid mixing between core flow and subcooled water from the hot legs for PWRs with combined-injection ECCS. This is a multi-dimensional phenomena. The heat transfer enhancement was also well-simulated with the above-mentioned model.

Numerical simulation of reflooding behavior in tight-lattice rod bundles

To evaluate the applicability of the reflood analysis code REFLA for ordinal pressurized water reactors to the analysis of reflooding phenomena in light water high conversion reactors (LWHCRs) with tight-lattice cores, a numerical simulation of the NEPTUN LWHCR test was performed with the REFLA code. The NEPTUN LWHCR test was performed at the Swiss Federal Insitute for Reactor Research with a test section simulating the tight-lattice core of an LWHCR. The results indicate no potential problems in the use of REFLA for the simulation of reflooding behavior in tight-lattice rod bundles. To improve the code, however, it is recommended to modify models of core heat transfer at a high flooding rate and core water distribution (integration of droplet flow) in the axial direction, and to investigate core pressure drop and horizontal cross flow.

Experimental Study of Effect of Initial Clad Temperature on Reflood Phenomena during PWR-LOCA

Integral system tests with the Cylindrical Core Test Facility (CCTF) were performed to investigate the effect of the initial clad temperature on the reflood phenomena in a PWRLOCA. The initial peak clad temperatures in these three tests were 871, 968 and 1,047 K, respectively. The feedback of the system on the core inlet mass flow rate was estimated to be little influenced by the variation of the initial clad temperature except for the first 20 s in the transient. The observed temperature rise from the reflood initiation was lower with the higher initial clad temperature. This qualitatively agreed with the results of the small scale forced feed reflood experiments. However, the magnitude of the temperature rise in CCTF was significantly low due to the high initial core inlet mass flow rate. Also observed were the multi-dimensional thermal behaviors for the three cases in the CCTF wide core. The analysis codes REFLA and TRAC reasonably predicted the effect of the initial clad temperature on the core thermo...