Fuel Cladding Temperature Research Articles

Thermal-hydraulics in uncovered core of light water reactor in severe core damage accident. (II) Analysis of Power Burst Facility severe fuel damage scoping test with SEFDAN code.

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.

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 modifications on the low flooding quench model and the transition flow are necessary.

Loss-of-Coolant Accidents in Small Compact Nuclear Reactors

The maximum credible accident for which a nuclear reactor must be analyzed is a loss-of-coolant accident (LOCA) due to a major rupture in the primary system. Such an accident has been analyzed for a conceptual design of a consolidated nuclear steam system (CNSS) using the thermal-hydraulic computer code RELAP4/MOD5.The results of the analysis show that the maximum fuel cladding temperature during the accident is ~344°C (652°F), which is sufficiently low to preclude any damage to the reactor core. Based on the results of this analysis, it appears that a LOCA in a CNSS may be less severe than in typical loop-type pressurized water and boiling water reactors. This result is expected because of the smaller piping connected to the CNSS reactor vessel

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 modifications on the low flooding quench model and the transition flow are necessary.

A study of the sensitivity of loca heat transfer analysis for a water-cooled reactor system

Abstract Results are given of computer calculations, using the reactor thermal analysis code THETA1-B, to determine the significance and relative importance of various heat transfer regimes in predicting maximum fuel cladding temperature for the blowdown phase of a postulated loss-of-coolant accident (LOCA) in a pressurized water reactor system. The factors considered include the choice of heat transfer correlation for a particular heat transfer regime, the method of delineating the boundaries between regimes, and core inlet coolant flow conditions. For a hot-leg rupture, the maximum surface temperature is sensitive to a number of factors, including choices of critical heat flux correlation, flow boiling transition heat transfer correlation, and in particular, stable film flow boiling correlation. However, for a LOCA resulting from a double-ended rupture of an inlet feeder, these factors have only marginal effects on maximum cladding temperature. In this case the importance of heat transfer to dry steam coolant at low net flow rate conditions is demonstrated, indicating a need for further information.

581. BWR非常用炉心冷却系統のフラッディングによる冷却効果

The effectiveness of emergency core cooling under loss-of-coolant accident conditions has been one of the most important safety issues that have interested the regulatory authorities in their review of applications for the construction of light-water-cooled nuclear power plants. Devices for flooding, to be adopted in conjunction with spray cooling, are included in emergency core cooling systems of Boiling Water Reactor (BWR) plants.The effectiveness of flooding heat transfer was tested with a full scale BWR fuel bundle simulator over the range of conditions expected in loss-of-coolant accidents.The results of the test prove the adequacy of the BWR flooding technique:(1) Core flooding effectively terminates core heat-up transients predicted to occur immediately following a loss-of-coolant accident. The post-accident peak fuel cladding temperature strongly depends on the fuel cladding temperature at the time of flooding initiation. Above 50mm/sec, differences in flooding rate are found to have no effect on the post-accident peak fuel cladding temperature.(2) Core decay heat is effectively removed for long periods of time after a loss-of-coolant accident when the core is flooded. Continued long term core cooling capability of BWR-ECCS is thus demonstrated.It should be noted that the water level required to cool the core decreases with increasing core power.

The determination of reactor fuel cladding temperature by recording krypton-85 release during post-irradiation heating

The temperature of evolution of fission product 86Kr from the inside surface of reactor canning material during controlled post-irradiation heating has been shown to agree to within 25 °C with the irradiation temperature, estimated by more direct means. The temperature found in this way is the maximum temperature of the specimen during the last period of the irradiation where this period is of the order of the time for half the atoms in the specimen to be displaced by the high energy particle flux in the reactor. The accuracy of determining by this method a maximum temperature set by a temperature excursion in the laboratory and found by a subsequent temperature rise experiment, has been shown to be ±7 °C.