Emergency Core Cooling Mode of the European Pressurized Water Reactor: The Evolution from Combined Injection to Cold-Leg Injection

1998 ◽  
Vol 124 (1) ◽  
pp. 65-81
Author(s):  
Florin Curca-Tivig
Author(s):  
Shinya Miyata ◽  
Satoru Kamohara ◽  
Wataru Sakuma ◽  
Hiroaki Nishi

In typical pressurized water reactor (PWR), to cope with beyond design basis events such as station black out (SBO) or small break loss of coolant accident with safety injection system failure, injection from accumulator sustains core cooling by compensating for loss of coolant. Core cooling is sustained by single- or two-phase natural circulation or reflux condensation depending on primary coolant mass inventory. Behavior of the natural circulation in PWR has been investigated in the facilities such as Large Scale Test Facility (LSTF) which is a full-height and full-pressure and thermal-hydraulic simulator of typical four-loop PWR. Two steady-state natural circulation tests were conducted in LSTF at both high and low pressure. These two tests were conducted changing the primary mass inventory as a test parameter, while keeping the other parameters such as core power, steam generator (SG) pressure, and steam generator water level as they are. Mitsubishi Heavy Industries (MHI) plans new natural circulation tests to cover wider range of core power and pressure as test-matrix (including the previous LSTF tests) to validate applicability of the model in wider range of core power and pressure conditions including the SBO conditions. In this paper, the previous LSTF natural circulation tests are reviewed and the new test plan will be described. Additionally, MHI also started a feasibility study to improve the steam generator tube and inlet/outlet plenum model using the M-RELAP5 code [4]. Newly developed model gives reasonable agreement with the previous LSTF tests and applies to the new test conditions. The feasibility findings will also be described in this paper.


2003 ◽  
Vol 143 (1) ◽  
pp. 37-56 ◽  
Author(s):  
Horst-Michael Prasser ◽  
Gerhard Grunwald ◽  
Thomas Höhne ◽  
Sören Kliem ◽  
Ulrich Rohde ◽  
...  

Author(s):  
Terry L. Schulz ◽  
Timothy S. Andreychek ◽  
Yong J. Song ◽  
Kevin F. McNamee

The AP1000 is a pressurized water reactor with passive safety features and extensive plant simplifications that provides significant and measurable improvements in safety, construction, reliability, operation, maintenance and costs. The design of the AP1000 incorporates a standard approach, which results in a plant design that can be constructed in multiple geographical regions with varying regulatory standards and expectations. The AP1000 uses proven technology, which builds on more than 2,500 reactor years of highly successful Westinghouse PWR operation. The AP1000 received Final Design Approval by the Nuclear Regulatory Commission in September 2004. The AP1000 Nuclear Power Plant uses natural recirculation of coolant to cool the core following a postulated Loss Of Coolant Accident (LOCA). Recirculation screens are provided in strategic areas of the plant to remove debris that might migrate with the water in containment and adversely affect core cooling. The approach used to avoid the potential for debris to plug the AP1000 recirculation screens is consistent with the guidance identified in Regulatory Guide 1.82 Revision 3, the Pressurized Water Reactor (PWR) Industry Guidance of NEI 04–07, and the Nuclear Regulatory Commission’s Safety Evaluation on NEI 04–07. Various contributors to screen plugging were considered, including debris that could be produced by a LOCA, resident containment debris and post accident chemical products that might be generated in the coolant pool that forms on the containment floor post-accident. The solution developed for AP1000 includes three major aspects, including the elimination of debris sources by design, features that prevent transportation of debris to the screens and the use of large advanced screen designs. Measures were taken to design out debris sources including fibers, particles and chemicals. Available industry data from walkdowns in existing plants is used to determine the characteristics and amounts of the fibrous and particulate debris that could exist in containment prior to the LOCA. Materials used in the AP1000 containment are selected to eliminate post accident chemical debris generation. Large, advanced screen designs that can tolerate significant quantities of debris have been incorporated. Testing has been performed which demonstrates that the AP1000 screens will have essentially no head loss considering the debris that could be transported to them. Testing has also been performed on an AP1000 fuel assembly that demonstrates that it will also have essentially no head loss considering the debris that could be transported to it.


Author(s):  
Xu Caihong ◽  
Shi Guobao ◽  
Fan Pu

The Advanced Core-cooling Mechanism Experiment (ACME) is conducted to investigate the performance of passive core-cooling system (PXS) for the advanced CAP1400 Pressurized Water Reactor (PWR). The small-break LOCA experiments conducted at ACME integrated test facility are simulated with a SNERDI modified version of RE-LAP5/MOD3 code. Several typical SBLOCA test cases are simulated and one case (2 inch cold leg break) is presented in this paper. And the predicted results are compared with the test data to assess the performance of the modified code. The calculated results agree reasonably well with the test data.


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