Kanji Tasaka

The effects of break location on PWR small break LOCA: Experimental study at the ROSA-IV LSTF

Abstract This paper discusses the effect of break location on the break flow rate and break flow quality transitions during a small-break loss-of-coolant accident (LOCA) in a pressurized water reactor (PWR). Results from five experiments conducted at the ROSA-IV Large Scale Test Facility (LSTF) are compared for this purpose. These experiments simulated a 2-inch break at the lower plenum, upper head, pressurizer top, cold leg, and hot leg, respectively. The controlling phenomena for the break flow quality transitions in cold-leg and hot-leg break experiments are described.

Summary of ROSA-IV LSTF first-phase test program — Integral simulation of PWR small-break LOCAs and transients -

Abstract Significant experimental results obtained at the ROSA-IV Large-Scale Test Facility (LSTF) during the first phase of the test program (1985–1988) are summarized. The LSTF is a 1 48 volumetrically scaled, full-height, full-pressure simulator of a Westinghouse-type four-loop (3423 MW thermal power) pressurized water reactor (PWR). The LSTF first-phase program investigated the fundamental PWR thermal-hydraulic responses during small-break loss-of-coolant accidents (SBLOCAs) and transients. The test matrix included twenty-nine SBLOCA tests, three abnormal transient tests and ten steady-state natural circulation tests.

ROSA-III Double-Ended Break Test Series for a Loss-of-Coolant Accident in a Boiling Water Reactor

The Rig of Safety Assessment (ROSA) III facility is a volumetrically scaled (1/424) boiling water reactor (BWR) system with an electrically heated core designed for integral loss-of-coolant accident (LOCA) and emergency-core-cooling-system (ECCS) tests. Experimental results obtained so far confirm that the severest single failure assumption in ECCS is the high-pressure core spray system failure even in a large-break LOCA in a BWR. The measured peak cladding temperature was well below the present safety criterion of 1473 K, even with the single failure assumption in ECCS, and the effectiveness of ECCS for core cooling during a double-ended-break LOCA has been confirmed. The overall agreement between the results calculated by the RELAP4/MOD6/U4/J3 computer code and the experimental results is good. The similarity between the ROSA-III test and a BWR LOCA has been confirmed through the comparison of calculated results for the ROSA-III facility and a BWR system.

Flow Regime Transition to Wavy Dispersed Flow for High-Pressure Steam/Water Two-Phase Flow in Horizontal Pipe

A wavy-dispersed flow regime was observed between slug and annular-dispersed flow regimes in TPTF high-pressure steam/water horizontal pipe experiments, employing the video probe visual observation. The onset OF entrainment was identified to cause slug to wavy-dispersed flow transition. The wavy-dispersed flow regime extended towards lower gas flow rates as pressure was increased. Furthermore, it was found that the gas-liquid relative velocity for the onset of entrainment decreases significantly, resulting in decrease in the minimum void fraction. Consequently, the slug flow regime was found to disappear for pressures above 8.6MPa, as observed in the previous TPTF experiments. Applicability of available models and correlations on the onset of entrainment was assessed against the TPTF data. Steen-Wallis parameter correlated the data well when the superficial gas velocity term in this parameter 1s replaced by the gas-liquid relative velocity.

Removal of scattered neutrons in thermal neutron radiography using a multichannel collimator

Abstract The present paper proposes and examines an experimental procedure for removing the influences of scattered neutrons and gamma-rays on thermal neutron radiography (TNR). A multichannel collimator coated with a neutron-absorbing compound eliminates scattered neutrons that fall on an imaging device. In order to measure contamination of the gamma-rays, a supplementary detector is used. By evaluating neutron transmittance of several objects, the procedure is confirmed to effectively remove scattered neutrons and gamma-rays within the detectable transmittance range of conventional TNR. In addition, the effect of changing the beam spectrum with penetrating lead on the relationship between neutron transmittance and thickness is determined.

Analyses of ROSA-III break area spectrum experiments on a boiling water reactor loss-of-coolant accident

The ROSA-III program has conducted system effects tests on the thermal-hydraulic response of a boiling water reactor during a loss-of-coolant accident. The performance of the emergency core cooling systems was of particular interest. As part of this program, ten tests were conducted with (a) a simulated pipe rupture located at the recirculation pump suction line, (b) a spectrum of break area ranging from 0 to 200% of scaled pipe cross-sectional area, and (c) an unavailable high-pressure core spray (HPCS) system. In these tests the pressure vessel depressurized (a) due to the actuation of the automatic depressurization system for scaled break areas of 50%, and (c) due to both for the intermediate break areas between 5 and 50%. Vessel depressurization enabled the injection of emergency core coolant from the low-pressure core spray and low-pressure coolant injection system and thus led to safe recovery without HPCS. The behaviors of the vessel pressure, core mixture level, and core temperatures were fairly well reproduced by the THYDE-B1 code, based on simplified lumpedparameter models for the wide spectrum of break areas investigated.

Feedback control of a primary pump for safe and stable operation of a PIUS-type reactor

Abstract A new automatic pump speed control system by using a measurement of the temperature distribution in the lower density lock is proposed for the PIUS-type reactor. This control system maintains the fluid temperature at the axial center of the lower density lock at the average of the fluid temperatures below and above the density lock in order to prevent the poison water from penetrating into the core during normal operation. The effectiveness of this control system was successfully confirmed by a series of experiments such as start-up and power ramping tests for the stable normal operation and a loss-of-feedwater test for the safe shutdown in an accident condition, using a small scale atmospheric pressure test loop which simulated the PIUS principle.

Relevant thermalhydraulic aspects of new generation LWRs

Abstract The present paper deals with the evaluation of thermalhydraulic aspects retained of importance for the assessment of safety of the new generation nuclear plants. Following a survey of the reactor concepts proposed for the future, the attention will be focused toward SBWR, AP-600 and PIUS whose characteristics, under many respects, bound the features introduced in the largest part of the new reactors. Expected relevant phenomena typical of the mentioned plants will be discussed in the paper: on this basis a critical overview of the experimental activities planned or in progress is presented and a judgement about the suitability of available computer codes is formulated. Conclusions are drawn in relation to the assessment of the new design proposals from a thermalhydraulic point of view.

Fission Product Decay Power—AESJ Recommendation

A recommendation regarding the FP decay power was issued from the Research Committee on Standardization of the Decay Heat Power in Nuclear Reactors of Atomic Energy Society of Japan (AESJ). The recommendation is primarily based on the results of summation calculations, obtained through using the JNDC FP Nuclear Data Library, Version 2. It consists of decay power values, their estimated accuracy, and their β- and γ-ray components for five fissionable nuclides; U-235, -238, Pu-239, -240 and -241. A set of 33-term exponential polynomials given therein help the users in calculating the decay-power curve for any irradiation history and cooling time. The time evolution for the γ-ray component energy spectra is also given.