The delay in L-Area startup provided an opportunity to obtain valuable data on the Emergency Cooling System (ECS) which will permit reactor operation at the highest safe power level. ECS flow is a major input to the FLOOD code which calculates reactor ECS power limits. The FLOOD code assesses the effectiveness of the ECS cooling capacity by modeling the core and plenum hydraulics under accident conditions. Presently, reactor power is not limited by the ECS cooling capacity (power limit). However, the manual calculations of ECS flows had been recently updated to include piping changes (debris strainer, valve changes, pressure release systems) and update fitting losses. Both updates resulted in reduced calculated ECS flows. Upon completion of...
Severe accident analysis for Small Break (SB), Middle Break (MB), and Large Break (LB) Loss-Of-Coola...
Thermal-hydraulics is one of the fundamental disciplines for the design and the operation of water c...
International audienceThe aging of the reactor pressure vessel (RPV) can be a limiting factor for th...
This paper discusses calculations related to hydraulics in a loss of coolant reactor accident. Earli...
A power limit criterion was developed for a postulated Loss of Pumping Accident (LOPA) in one of the...
Emergency core cooling system (ECCS) has been studied extensively for reactor safety. Emergency core...
Currently Appendix K of 10 CFR 50 is used in the United States to evaluate models for the emergency ...
Evaluation of the adequacy of the emergency cooling water addition system (ECWA system) requires ana...
The paper describes the combination of safety and normal operation functions adopted in the AES-2006...
International audienceIn the context of improvement studies for pressurized water reactors (PWR) in ...
After the FUKUSHIMA accident, passive systems become an important issue for the new projects of Pres...
Emergency Core Coolant Bypass (ECC Bypass) has been regarded as an important phenomenon to peak clad...
This paper summarizes the results of TRAC-PF1/MOD3 calculations of Mark-22 fuel assembly of loss-of-...
The newly developed, advanced, light water reactor (LWR) simulation code, RELAP5, is used to analyze...
The PSB reviews reactor auxiliary cooling water systems (CWS) that are required for safe shutdown du...
Severe accident analysis for Small Break (SB), Middle Break (MB), and Large Break (LB) Loss-Of-Coola...
Thermal-hydraulics is one of the fundamental disciplines for the design and the operation of water c...
International audienceThe aging of the reactor pressure vessel (RPV) can be a limiting factor for th...
This paper discusses calculations related to hydraulics in a loss of coolant reactor accident. Earli...
A power limit criterion was developed for a postulated Loss of Pumping Accident (LOPA) in one of the...
Emergency core cooling system (ECCS) has been studied extensively for reactor safety. Emergency core...
Currently Appendix K of 10 CFR 50 is used in the United States to evaluate models for the emergency ...
Evaluation of the adequacy of the emergency cooling water addition system (ECWA system) requires ana...
The paper describes the combination of safety and normal operation functions adopted in the AES-2006...
International audienceIn the context of improvement studies for pressurized water reactors (PWR) in ...
After the FUKUSHIMA accident, passive systems become an important issue for the new projects of Pres...
Emergency Core Coolant Bypass (ECC Bypass) has been regarded as an important phenomenon to peak clad...
This paper summarizes the results of TRAC-PF1/MOD3 calculations of Mark-22 fuel assembly of loss-of-...
The newly developed, advanced, light water reactor (LWR) simulation code, RELAP5, is used to analyze...
The PSB reviews reactor auxiliary cooling water systems (CWS) that are required for safe shutdown du...
Severe accident analysis for Small Break (SB), Middle Break (MB), and Large Break (LB) Loss-Of-Coola...
Thermal-hydraulics is one of the fundamental disciplines for the design and the operation of water c...
International audienceThe aging of the reactor pressure vessel (RPV) can be a limiting factor for th...