Study on Measures During Loss of Normal Feedwater Accident for AP1000 NPP in Lower Power Operation

Abstract If an accident occurred in a plant in lower power operation, a series of problems could be caused by low residual heat and excessive removal capability of heat sinks. Taking a real accident as example, this paper discussed how to deal with loss of normal feedwater in AP1000 NPP during lower power operation. The level of Steam Generators (SG) will be rapidly reduced in case of Main Feedwater loss, and Startup Feedwater pump will start. If the reactor is not shut down immediately, the Startup Feedwater will not be enough to prevent SG level from falling, and automatic reactor trip will be inevitable. Therefore, in event of loss of Main Feedwater which cannot be recovered instantly, the reactor should be shut down manually to maintain SG level as high as possible. Appropriate measures should be taken to avoid excessive cooling after trip. Attention should be paid to SG level and feedwater flow to avoid Passive Residual Heat Removal Heat Exchanger (PRHR HX) actuation during regulating Startup Feedwater. Otherwise, PRHR HX may be actuated unexpectedly. With Reactor Coolant Pumps operating, PRHR HX heat removal capacity is strong. The Cold Leg temperature will drop rapidly because of PRHR HX and other heat sinks. In this case, the most effective measure to stabilize temperature of Cold legs is to isolate Streamline. Otherwise, structure integrity of the Reactor Pressure Vessel (RPV) will be challenged under Pressurized Thermal Shock (PTS) due to rapid drop of Cold leg temperature. In order to alleviate PTS, depressurization shall be performed and further cooling shall be suspended. Since PRHR HX is only connected to Loop 1, coolant in Hot Leg of Loop 2 is still kept at high temperature. Therefore, necessary measures shall be taken to prevent coolant flash in Loop 2 during the process of depressurization, which may fill the Pressurizer up with water. The RPV Head Vent valves perform safety-related function by preventing Pressurizer overfill in certain design basis events in AP1000 NPP. However, emergency letdown by opening the valves cannot effectively reduce Pressurizer level when coolant flashed in Loop 2. Because of mitigating PTS, opening RPV Head Vent valves, and steam condensation in Loop 2, the subcooling of the coolant at the core outlet is likely to drop below 6 °C or even 0 °C.

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A Simulation of Small Break Loss of Coolant Accident in Nuclear Heating Reactor Based on RELAP5

Volume 6A: Thermal-Hydraulics and Safety Analyses ◽

10.1115/icone26-81416 ◽

2018 ◽

Author(s):

Meng Lu ◽

Heng Xie

Keyword(s):

Thermal Power ◽

Heat Removal ◽

Injection System ◽

Water Volume ◽

Nuclear Heating Reactor ◽

Removal Capacity ◽

Loss Of Coolant ◽

Nuclear Heating ◽

Residual Heat

Nuclear heating reactor is integrated designed without main pump and safety injection system. The loss of coolant accidents are mainly in the form of small break LOCA. As no safety injection system is designed for coolant makeup, the water volume in the reactor vessel is critical since it determines whether the reactor will be submerged during the whole scenario. Therefore, the study on coolant loss in this pool system is indispensable. The RELAP5 code has been developed for best-estimate transient simulation of light water reactor coolant systems during postulated accidents. The long term effect in nuclear heating reactor is important. In this paper we investigated the influential factors on SBLOCA scenario and found the long term residual heat removal capacity is decisive in determining the loss of coolant. The residual heat removal capacity should be greater than 2% of reactor thermal power if ensuring the core submerged in the long run.

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AP1000 Passive Residual Heat Removal Heat Exchanger Confirmatory Analysis

Volume 2: Thermal Hydraulics ◽

10.1115/icone14-89524 ◽

2006 ◽

Cited By ~ 5

Author(s):

Richard F. Wright ◽

James R. Schwall ◽

Creed Taylor ◽

Naeem U. Karim ◽

Jivan G. Thakkar ◽

...

Keyword(s):

Heat Exchanger ◽

Cooling System ◽

Heat Removal ◽

Passive Safety ◽

Tube Heat Exchanger ◽

Removal Capacity ◽

Safety Margins ◽

Final Design ◽

Safety Features ◽

Residual Heat

The AP1000 is an 1100 MWe advanced nuclear power plant that uses passive safety features to enhance plant safety and to provide significant and measurable improvements in plant simplification, reliability, investment protection and plant costs. The AP1000 received final design approval from the US-NRC in 2004. The AP1000 design is based on the AP600 design that received final design approval in 1999. Wherever possible, the AP1000 plant configuration and layout was kept the same as AP600 to take advantage of the maturity of the design and to minimize new design efforts. As a result, the two-loop configuration was maintained for AP1000, and the containment vessel diameter was kept the same. It was determined that this significant power uprate was well within the capability of the passive safety features, and that the safety margins for AP1000 were greater than those of operating PWRs. A key feature of the passive core cooling system is the passive residual heat removal heat exchanger (PRHR HX) that provides decay heat removal for postulated LOCA and non-LOCA events. The PRHR HX is a C-tube heat exchanger located in the in-containment refueling water storage tank (IRWST) above the core promoting natural circulation heat removal between the reactor cooling system and the tank. Component testing was performed for the AP600 PRHR HX to determine the heat transfer characteristics and to develop correlations to be used for the AP1000 safety analysis codes. The data from these tests were confirmed by subsequent integral tests at three separate facilities including the ROSA facility in Japan. Owing to the importance of this component, an independent analysis has been performed using the ATHOS-based computational fluid dynamics computer code PRHRCFD. Two separate models of the PRHR HX and IRWST have been developed representing the ROSA test geometry and the AP1000 plant geometry. Confirmation of the ROSA test results were used to validate PRHRCFD, and the AP1000 plant model was used to confirm the heat removal capacity for the full-sized heat exchanger. The results of these simulations show that the heat removal capacity of the PRHR HX is conservatively represented in the AP1000 safety analyses.

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