Investigation of Pre- and Post-Dryout Heat Transfer in Upward Steam-Water Two-Phase Flow at Low Flow Rate

A heat transfer experiment was performed on steam-water two-phase flow in an annular flow path with a uniformly heated rod under the conditions of the mass flow rates from 0.2xlO⁶ to l.Ox 10⁶ kg/m²-h, inlet qualities from 0.5 to 1.0, heat fluxes below 4.7x 10⁵ W/m² and pressure of 31 bar. Dryout of the heater rod surface was observed resulting in the sharp rise of the heater rod surface temperature. Measured heat transfer coefficients were compared with the several empirical and semi-empirical correlations with the emphasis on the applicability of the correlations to the present test conditions being important in the analysis of the thermal hydraulic behavior during a LOCA of a nuclear reactor. The measured heat transfer coefficient in the pre-dryout region is lower than the existing correlations. The cooling of the heat transfer surface by the liquid phase in the post-dryout region is significant, which is neglected in the existing correlations. The heat transfer coefficients calculated for the post-dryout region by the Groeneveld correlation show good agreement with the presently measured results within the accuracy of 0~27%.     

Thermal-Hydraulics in Uncovered Core of Light Water Reactor in Severe Core Damage Accident, (III)

The SEFDAN is a computer program to analyze the one-dimensional thermal-hydraulics of a partially uncovered core of a light water reactor in a severe degraded-cooling event. In order to verify the code and to obtain better understanding of the severe core damage process, SEFDAN has been applied to analyses of the thermal response of fuel rods in the Power Burst Facility Severe Fuel Damage 1-1 Test.

The calculated results are in good agreement with the experimental results. The analysis indicates that fuel cladding temperature in a portion of the lowest one third of the test bundle would have reached the melting point of the ZrO₂ during a rapid temperature excursion driven by the zirconium-water reaction. The result is consistent with the result of metallographic examination. The crucibilization effect of the ZrO₂ layer played an important role in the reaction. Steam starvation condition would have occurred in contrast to the situation of the Scoping Test of the same test series. Zirconium-water reaction on the inner surface of the fuel cladding was found to have made a strong effect on the fuel rod temperature in the upper part of the test bundle.     

Thermal-Hydraulics in Uncovered Core of Light Water Reactor in Severe Core Damage Accident, (II)

The SEFDAN is a computer program to analyze the one-dimensional thermal-hydraulics of a partially uncovered core of a light water reactor in a severe degraded-cooling event. In order to verify the code and to obtain better understanding of the severe core damage process, SEFDAN has been applied to analyses of the thermal response of fuel rods in the Power Burst Facility Severe Fuel Damage Scoping Test. This paper presents the calculated results and discusses, based on the results, on the phenomena that are important for prediction of the thermal response of fuel rods to a severe accident under the partially un-covered core condition. The calculated results are in good agreement with the experimental results. Namely the dry-out time of each elevation and the temperature behavior in both the slow heat-up and rapid temperature excursion processes are well simulated. The analysis indicates that fuel cladding temperature of the upper part of the test bundle would have reached the melting point of ZrO₂ and fuel center line temperature would have reached the melting point of UO₂ during a rapid temperature excursion which was caused by rapid decreasing of the dry-out level and accelerated by zirconium-water reaction in the lower part.     

New Theoretical Model for Two-Phase Flow Discharged from Stratified Two-Phase Region through Small Break

A theoretical and experimental study was conducted to understand two-phase flow discharged from a stratified two-phase region through a small break. This problem is important for an analysis of a small break loss-of-coolant accident (LOCA) in a light water reactor (LWR). The present theoretical results show that a break quality is a function of h/h_b , where h is the elevation difference between a bulk water level in the upstream region and the break and b the suffix for entrainment initiation. This result is consistent with existing experimental results in literature. An air-water experiment was also conducted changing a break orientation as an experimental parameter to develop and assess the model. Comparisons between the model and the experimental results show that the present model can satisfactorily predict the flow rate and the quality at the break without using any adjusting constant when liquid entrainment occurs in a stratified two-phase region. When gas entrainment occurs, the experimental data are correlated well by using a single empirical constant.     

Safety Concerning Transportation of Spent Fuel

The Central Research Institute of Electric Power Industry (CRIEPI) has been conducting, under contract with the Science and Technology Agency of Japan, the spent fuel transport cask reliability demonstration test since 1977 to verify the safety and reliability of spent fuel transport casks. The first phase of this test was completed in 1987.

In this demonstration test, both 50 t and 100 t class of casks, designed and manufactured by current techniques, were subjected to tests to verify the integrity and adequacy of the design and manufacturing techniques through observation of behavior of the cask under test conditions. The casks were subjected to tests under normal conditions and under the accident conditions specified in the Japanese regulations and the IAEA regulations, and also to pressure tests, which were performed from the viewpoint of safety in shipping, although by sea, this is not specified in the Japanese regulations.

From the test results, it was confirmed that the 1001 class cask maintained its integrity and characteristics in conformity with regulations even after accident condition tests. It is clear that the design concept and manufacturing procedure employed for this cask is adequate.     

Transient Behavior of Low Enriched Uranium Silicide Plate-Type Fuel for Research Reactors during Reactivity Initiated Accident Conditions

Abstract

This paper describes the results of transient experiments using a low enriched uranium silicide miniplate fuel for research reactor. The pulse irradiation was performed in the Nuclear Safety Research Reactor (NSRR) at the Japan Atomic Energy Research Institute (JAERI).

The results obtained in this study are summarized as follows :

(1) The tested fuel plates were damaged with energy depositions above 94 cal/g·fuel, but remained intact below 82cal/g·fuel. A failure threshold should therefore exist between these two values.

(2) Four of the fuel plates that showed peak cladding surface temperatures <330°C were damaged by the thermal stress during quenching. These damaged fuel plates revealed small intergranular cracks that propagated perpendicularly to the axial direction of the plate, from the Al cladding surface to the fuel core, without significant dimensional changes. On the other hand, when peak cladding surface temperatures were >400°C, the test fuel plates were damaged mainly by melting of the Al cladding, accompanying significant dimensional changes.

(3) The thermal stress of the damaged fuel plates calculated on the basis of the maximum transient temperature drop during quenching was greater than the tensile stress that occurred during fabrication.     

SCDAP/RELAP5 Analysis of Station Blackout with Pump Seal LOCA in Surry Plant

During a station blackout of PWR, the pump seal will fail due to loss of the seal cooling. This particular transient-LOCA sequence designated as S3-TMLB' analyzed by SNL with MELPROG/TRAC for Surry plant showed that the depressurization due to the pump seal LOCA would result in early accumulator injection and subsequent core cooling which lead to the delay of reactor pressure vessel (RPV) meltthrough. The present analysis was performed with SCDAP/RELAP5 to evaluate this scenario shown in the MELPROG/TRAC analyses. Addition-ally, the calculated results were compared with the similar experimental studies of JAERI's ROSA-IV program.

The present analyses showed that: (1) During S3-TMLB', the loop seal clearing would occur and cause a slight delay of accident progression. (2) It is unlikely that the accumulator injection, which leads to the delay of RPV meltthrough by approximately 60 min, is initiated automatically during S3-TMLB'. Accordingly, an intentional depressurization using PORVs is recommended for the mitigation of the accident consequences. (3) The present SCDAP/RELAP5 analyses did not show significant delay of accident progression. It was found that non-realistic lower heat generation and higher core cooling models used in the MELPROG/TRAC analysis are attributed to this discrepancy.     

Experimental and Analytical Study on the Behavior of Cesium Iodide Aerosol/Vapor Deposition onto Inner Surface of Pipe Wall under Severe Accident Conditions

The WAVE experiments have been performed at JAERI to investigate the CsI deposition onto the inner surface of pipe wall under typical severe accident conditions. It was shown that relatively large amount of CsI was deposited at the upstream floor of the pipe and that larger amount of CsI was deposited on the ceiling than the floor at the downstream. Analyses of the experiments have also been conducted with the three-dimensional thermohydraulic code, SPRAC, and the radionuclide transport analysis code, ART. The experimental results were well reproduced with ART by using peripherally subdivided pipe cross section and associated representative thermohydraulic information from SPRAC prediction. It was clarified through the present experiment and analyses that major deposition mechanisms for the chemical form of CsI are thermophoresis and condensation. Accordingly, the coupling of the FP behavior and the detailed thermohydraulic analyses was found to be essential in order to accurately predict the CsI deposition in the pipe, to which little attention has been paid in the previous studies.     

Fission Gas Induced Cladding Deformation of LWR Fuel Rods under Reactivity Initiated Accident Conditions

Behavior of irradiated fuel rods under power burst conditions by accidental reactivity insertion in light water reactors (LWRs) has been studied in the Nuclear Safety Research Reactor (NSRR). In the experiments, cladding hoop deformation, which reached up to about 10%, was much larger than that of the fresh rods. The current LWR fuel behavior analysis codes, which only take account of the thermal expansion of the fuel pellets for the deformation calculation, under-predicted the plastic deformation of the cladding to be less than about 1%. Fission gas release during the pulse irradiation tests reached as high as 22% in the NSRR irradiated fuel tests. In order to describe these test results, a model of grain boundary fission gases to cause the cladding deformation has been developed and installed in a fuel behavior simulation code, FRAP-T6. In the model, the over-pressurized gases by the pulse irradiation cause grain boundary separation and stress the cladding during the tests. The model assumes that the gases remain in the fuel during the early part of pulse irradiation and are released to the open volume in the rod after the cladding deformation. The model, in combination with a fuel thermal expansion model, GAPCON, which was validated through fresh fuel tests, reproduces the NSRR test results reasonably well.