Heat-up Transients Research Articles

Recent improvements of probabilistic fracture mechanics analysis code PASCAL for reactor pressure vessels

A probabilistic fracture mechanics (PFM) analysis code PASCAL was developed in Japan for probabilistic integrity assessment of reactor pressure vessels (RPVs) in pressurized water reactors (PWRs) considering neutron irradiation embrittlement and pressurized thermal shock (PTS) transients. To strengthen the practical applications of PFM methodology in Japan, PASCAL has been upgraded to a new version, PASCAL5, which enables PFM analyses of RPVs in both PWRs and boiling water reactors (BWRs) subjected to a broad range of transients, including PTS and normal operational transients. In this paper, the recent improvements in PASCAL5 are described such as the incorporated stress intensity factor solutions and corresponding calculation methods for external surface cracks and embedded cracks near the RPV outer surface. In addition, the analysis conditions and evaluation models recommended for PFM analyses of Japanese RPVs in BWRs are investigated. Finally, PFM analysis examples for core region of a Japanese BWR-type model RPV subjected to two transients (i.e., low-temperature over pressure and heat-up transients) are presented using PASCAL5.

An Analytical Approach to Transient Room Temperature Analysis

For the past few years, nuclear utilities have been interested in the calculation of transient room temperatures at various station locations following loss of heating/ventilating/air conditioning and/or following station blackout. Transient room temperature analyses invariably involve the use of computer programs utilizing various finite difference schemes. A manual solution method is proposed for room heatup transients, thereby reducing the engineering time spent to obtain the results from tens (occasionally hundreds) of hours to a few hours in many cases.

Thermodynamic investigations of passive decay heat removal from HTR cores and component behaviour

This paper describes core heatup transients for the THTR-300 and HTR-Module and the knowledge obtained with the experimental facility LUNA-HTR for the study of natural convection characteristics of different HTR primary circuits. In consideration of thermodynamic similarity principles, transients with heat removal via natural convection from a pebble-bed into the primary circuit were simulated, as well as calculated with the code THERMIX-CRAY-2D. On the basis of experiments, the code is furthermore used for the analysis of natural decay heat dissipation in the depressurization case of the primary circuit. Studies on HTR safety have indicated the great significance of processes resulting during temperature stressing of the concrete pressure vessel for the sequence and consequences of accidents, particularly those with unrestricted core heat-up. In the course of the accident the vessel is slowly heated up to very high temperatures. Over the last ten years the behaviour of PCRVs of different HTR (THTR, HTR-500) plants during core heat-up accidents has been analyzed. Besides the PCRV itself, also the top reflector suspension is experimentally investigated.

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.