Residual Heat Exchanger Research Articles

Recent progress of CFD applications in PWR thermal hydraulics study and future directions

CFD method has the potential to simulate the detailed three-dimensional flow and heat transfer features in nuclear reactors, and is promising to play a more important role in the future reactor design and thermal hydraulics analysis. In recent years, series of research achievements in nuclear engineering based on the CFD method, including the single-phase application, two-phase model development and multi-scale and multi-physics coupling, are accomplished around the world. XJTU-NuTheL has also been committed to the development and application of high fidelity thermal–hydraulic models using CFD method. The three-dimensional CFD models of key equipments in nuclear power plants, such as RPV, SG, valves, T-junctions and passive residual heat exchanger, are developed. The mathematical models of complicate two-phase boiling phenomena and thermal hydraulic features under the motion conditions are established. In addition, with the high fidelity simulation requirement of the whole reactor system, the nuclear reactor multi-scale and multi-physics coupling platforms are developed on the basis of CFD codes. In this paper, the latest progress of nuclear reactor thermal–hydraulic research using CFD method is outlined, especially at XJTU-NuTheL. The major challenges and promising directions of CFD method in the nuclear reactor engineering are proposed, which would be beneficial for the promotion of its further applications.

Development of a thermal–hydraulic analysis software for a passive residual heat removal system

A series of reasonable mathematical and physical models for the thermal–hydraulic characteristic analysis of a passive residual heat removal system (PRHRS) in an integral pressurized water reactor were established. These models mainly include the core, once-through steam generator, nitrogen pressurizer, main coolant pump, and flow and heat transfer models. The flow and heat transfer models are suitable for the core with plate-type fuel element and the once-through steam generator with annular channel, respectively. A transient analysis code with Visual Fortran 6.5 has been developed to analyze the thermal–hydraulic characteristics of the PRHRS. Visual input, real-time processing and dynamic visualization output were achieved with Microsoft Visual Studio.NET 2008, which greatly facilitate applications in the engineering. The software was applied to analyze the effects of the heat transfer area of the PRHRS, the height difference between the residual heat exchanger center and the steam generator center, and the main steam valve turn-off time on the transient thermal–hydraulic characteristics of PRHRS. The obtained analysis results are significant to the improvement design of the PRHRS and the safety operation of the integral pressurized water reactor.

Thermal Hydraulic Analysis of a Passive Residual Heat Removal System for an Integral Pressurized Water Reactor

A theoretical investigation on the thermal hydraulic characteristics of a new type of passive residual heat removal system (PRHRS), which is connected to the reactor coolant system via the secondary side of the steam generator, for an integral pressurized water reactor is presented in this paper. Three‐interknited natural circulation loops are adopted by this PRHRS to remove the residual heat of the reactor core after a reactor trip. Based on the one‐dimensional model and a simulation code (SCPRHRS), the transient behaviors of the PRHRS as well as the effects of the height difference between the steam generator and the heat exchanger and the heat transfer area of the heat exchanger are studied in detail. Through the calculation analysis, it is found that the calculated parameter variation trends are reasonable. The higher height difference between the steam generator and the residual heat exchanger and the larger heat transfer area of the residual heat exchanger are favorable to the passive residual heat removal system.

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

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.

New fabrication techniques for components

This paper presents the high quality level of nuclear technology in the Federal Republic of Germany, which has been achieved by the consistent application of advanced manufacturing techniques. Therefore the synchronous development and sophistication of manufacturing process coupled with the utilization of optimized materials is demonstrated by two typical examples. The forged prechamber of the residual heat exchangers as well as the inductive bending of thick-walled pipes show that the use of complex forgings and of L- and Z-shaped piping may guarantee perfectly satisfactory fabrication and testing conditions while at the same time a reduction in the fabrication and testing costs may be achieved.