Emergency Operating Procedures (EOPs) for the AP1000 Simulator

Author(s):  
Yuichi Hayashi ◽  
Gianfranco Saiu ◽  
Richard F. Wright

The AP1000 is two-loop 1100 MWe advanced pressurized water reactor (PWR) 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 uses proven technology, which builds on over 30 years of operating PWR experience. The AP1000 final design certification was approved by the NRC in December, 2005. A total of 34 Emergency Operating Procedures (EOPs) for operation of the AP1000 simulator have been prepared based on the AP1000 Emergency Response Guidelines (ERGs), background information documents and detailed plant information. These include 28 EOPs at power and 6 EOPs during shutdown. The AP1000 ERGs were developed by using the generic ERGs for the low pressure reference PWR plant as a basis. The AP1000 design differences from the reference plant were reviewed and reflected in the process of developing operational steps in each ERG. The provisions of the AP1000 PRA were also reviewed and incorporated into the ERGs. Although the AP1000 design does not require operator actions for the first 72 hours after accidents, the operator actions with both safety-related and nonsafety-related equipment have an important role to mitigate the consequence of accidents. In the event of a steam generator tube rupture (SGTR), although the AP1000 is designed so that no operator actions are required to recover from the event, there are actions that can be taken by the operator to limit the release of radioactive effluents from the ruptured SG. These actions include isolation of the ruptured SG and depressurization of the reactor coolant system (RCS) to terminate primary-to-secondary leakage, restoring reactor coolant inventory to ensure adequate core cooling and plant pressure control. It is expected that these operator actions should be incorporated into the ERG to reduce the fission product release. To support the development of the AP1000 ERGs, several transient and accident analyses were performed. These include analyses for LOCA, post-LOCA cooldown and depressurization, passive safety system termination, SGTR and faulted SG isolation. These analyses results were incorporated into the ERG background information documents. In the event of SGTR, several cases were analyzed, including consideration of operator recovery actions. These cases were modeled using the best-estimate state-of-art RELAP5 code. The analyses results show that operator recovery actions are effective for SGTR to be placed under operator control.

1986 ◽  
Vol 108 (3) ◽  
pp. 346-351
Author(s):  
W. T. Kaiser ◽  
B. S. Monty

The operational concern of pressurized thermal shock (PTS) can be minimized by proper operator guidance. This paper presents a method for calculating a pressure temperature limit curve for reactor vessel integrity which can be used to identify an ongoing potential PTS event. This method has been developed for use and is applicable to all pressurized water reactors. The curve is used in emergency operating procedures developed to prioritize various plant safety concerns including PTS and core cooling to ensure proper operator action during accident conditions. This paper emphasizes the development of the pressure-temperature limit and how it is used within the emergency operating procedures.


Author(s):  
Terry L. Schulz ◽  
Julie Gorgemans

The AP1000 plant is an 1100-MWe pressurized water reactor (PWR) with passive safety features and extensive plant simplifications that enhance construction, operation, maintenance and safety. One of the key design approaches in the AP1000 plant design is to use passive features to mitigate design basis accidents. Active defense-in-depth (DiD) features provide investment protection, reduce the demands on the passive features and support the Probabilistic Risk Assessment (PRA). The passive features are classified as safety-related in the United States. The active DiD features are classified as nonsafety-related (with supplemental requirements) in the United States. The AP1000 plant design has also incorporated a standardization approach, which together with the level of safety achieved by the passive safety features, results in a plant design that can be applied to different geographical regions with varying regulatory standards and utility expectations without major changes. This paper will discuss the approach taken to defining DiD in the AP1000 plant and the effectiveness of that approach. It will also address the capability of the AP1000 plant to meet deterministic DiD guidelines such as the ones in application in the UK or described in the Western European Nuclear Regulators’ Association (WENRA) safety objectives for new plants.


Author(s):  
Zhanfei Qi ◽  
Sheng Zhu

CAP1400 Pressurized Water Reactor is developed by China’s State Nuclear Power Technology Corporation (SNPTC) based on the passive safety concept and advanced system design. The Advanced Core-cooling Mechanism Experiment (ACME) integral effect test facility, which was constructed at Tsinghua University, represents a 1/3-scale height of CAP1400 RCS and passive safety features. It is designed to simulate the performance of CAP1400 passive core cooling system in the small break loss of coolant accidents (SBLOCA) for design certification, safety review and safety analysis code development. The Long Term Core Cooling (LTCC) post-LOCA could be simulated by ACME as well. A series of test cases with various break sizes and locations with post-LOCA LTCC period were conducted in ACME facility. This paper describes the post-LOCA LTCC test conducted in ACME test facility. The LTCC phenomena in different cases are very similar. It’s found that the interval that switching from IRWST injection to sump recirculation has the least safety margin. However, it’s shown that the post-LOCA LTCC in ACME could be well maintained by passive core cooling system according to the test results even though the recirculation water level in ACME IRWST-2 is lower than the containment recircualtion level in CAP1400 conservatively.


Author(s):  
Roberta Ferri ◽  
Andrea Achilli ◽  
Cinzia Congiu ◽  
Gustavo Cattadori ◽  
Fosco Bianchi ◽  
...  

The SPES3 facility is being built at the SIET laboratories, in the frame of an R&D program on Nuclear Fission, led by ENEA and funded by the Italian Ministry of Economic Development. The facility is based on the IRIS reactor design, an advanced medium size, integral layout, pressurized water reactor, based on the proven technology of PWR with an innovative configuration and safety features suitable to cope with Loss of Coolant Accidents through a dynamic coupling of the primary and containment systems. SPES3 is suitable to test the plant response to postulated Design and Beyond Design Basis Events, providing experimental data for code validation and plant safety analysis. It reproduces the primary, secondary and containment systems of the reactor with 1:100 volume scale, full elevation, prototypical fluid and thermal-hydraulic conditions. A design-calculation feedback process, based on the comparison between IRIS and SPES3 simulations, performed respectively by FER, with GOTHIC and RELAP5 coupled codes, and by SIET, with RELAP5 code, led to reduce the differences in the two plants behaviour, versus a 2-inch equivalent DVI line DEG break, considered the most challenging LOCA for the IRIS plant. Once available the final design of SPES3, further calculations were performed to investigate Beyond Design Basis Events, where the intervention of the Passive Containment Condenser is fundamental for the accident recovery. Sensitivity analyses showed the importance of the PCC actuation time, to limit the containment pressure, to reach an early pressure equalization between the primary and containment systems and to allow passive water transfer from the containment to the RPV, enhanced by the ADS Stage-II opening.


Author(s):  
Mian Xing ◽  
Zhaocan Meng ◽  
Xiaotao Liao ◽  
Canhui Sun ◽  
Shuming Zhang ◽  
...  

SPICRI (State Power Investment Central Research Institute) is developing a new conceptual design of heating-reactor, named Heating-reactor of Advanced low-Pressurized and Passive safetY system (HAPPY), which is targeted for the district heating, desalination of seawater, and other heat applications. It is a 200MWth two-loop low-pressurized water reactor with low thermal parameters. The whole reactor vessel is deployed inside a shielding and cooling pool with thermal insulation measure. The conceptual design of HAPPY is described in this paper, including the design criteria, safety features, main parameters and main components. A preliminary safety analysis is carried out to provide a reference for the design and optimization of HAPPY. In this paper, four different LOCA analyses are described and compared. The results show that the current design can deal well with all the selected LOCA scenarios and the effectiveness of the safety systems is proved.


Author(s):  
Xiaochuan Zang ◽  
Tao Liu

The emergency action level (EAL) scheme for a modular high temperature gas-cooled reactor (HTR) plant refers to the generic EAL development guidance for pressurized water reactors (PWR) with HTR modification due to its design issues. Based on reactor’s accidents analysis and consequence assessment, EAL scheme of HTR is established through the steps of category and classification. Four emergency classes are set for HTR consisting of U (Emergency Standby), A (Facilities Emergency), S (Site Area Emergency) and G (General Emergency). The Recognition Category of Initiating Condition (IC) and EAL contains A - Abnormal Rad Levels / Radiological Effluent, F - Fission Product Barrier, H - Hazards and Other Conditions Affecting Plant Safety, S - System Malfunction. The methodology for development of EALs for HTR on Fission Product Barrier and System Malfunction has some differences from PWR’s due to differences on operating mode, inherent safety features and system characteristics.


Author(s):  
Sheng Zhu

CAP1400 is a large pressurized water reactor based on the passive safety conception. An ACME (Advanced Core-cooling Mechanism Experiment) facility has been designed and constructed in order to validate that the CAP1400 system design is acceptable to mitigate the loss of coolant accident (LOCA). The ACME test facility is an isotonic pressure, 1/3-scale height and 1/54.32-scale power simulation of the prototype CAP1400 nuclear power plant. It contains the main-loop system, passive safety system, secondary steam system and auxiliary system etc. The all of ACME test matrix including 5 kinds 21 cases .In this paper, the test results and the Realp5 prediction of the cold leg 5cm break accident of CAP1400 are compared and analyzed to briefly evaluate the ACME capability. Furthermore, 3 different types of 5cm cold leg break test cases are presented, and the transient process, system responses and key parameters tendency are analyzed based on the test. The results indicate that the passive safety system design can successfully combine to provide a continuous removal of core decay heat and the reactor core remains to be covered with considerable margin for the 3 different 5cm cold leg break accidents.


Author(s):  
Linsen Li ◽  
Feng Shen ◽  
Mian Xing ◽  
Zhan Liu ◽  
Zhanfei Qi

A small Pressurized Water Reactor (PWR) with compact primary system and passive safety feature, which is called Compact Small Reactor (CSR), is under pre-conceptual design and development. For the purpose of preliminary assessment of the primary coolant system and capability evaluation of the passive safety system, a detailed thermal-hydraulic (T-H) system model of the CSR was developed. Several design-basis accidents, including feedwater line break, double ended direct vessel injection line break (one of the small-break Loss Of Coolant Accidents, LOCA) and etc, are selected and simulated so as to evaluate and further optimize the design of passive safety systems, especially the passive core cooling system. The results of preliminary accident analysis show that the passive safety systems are basically capable of mitigating the accidents and protecting the reactor core from severe damage. Further research will be focused on the optimization of pre-conceptual design of the thermal-hydraulic system and the passive core cooling system.


Author(s):  
Nathan R. Murray ◽  
Mitchell E. Sailsbery ◽  
Samuel E. Bischoff ◽  
Paul R. Wilding ◽  
Matthew J. Memmott

A passive endothermic reaction cooling system (PERCS) is proposed to provide reactor core cooling during a station blackout (SBO). During a SBO, a PWR in which PERCS has been installed has a peak reactor core outlet temperature remains below 640 K (692.3°F) for 30 days, which is well below the nominal accident core outlet temperature during a SBO. During a LOCA, LOFA, and LOHSA, installation of a PERCS has no significant impact on safety performance. It should be noted that the PERCS will represent a minimal heat source (unless the PERCS is very large) during DBAs as emergency systems lower the coolant temperature below the PERCS temperature. A typical PWR with an installed PERCS is modeled using RELAP5-3D. The results of the model demonstrate the high level of passive safety afforded by the PERCS which contributes to the mitigation of SBO consequences without adversely affecting nuclear plant safety during a LOCA, LOHSA, or LOFA. Future work in validating the PERCS as a method of passive safety for existing light water reactors is underway, including the refining the physical design, determining better kinetic and thermodynamic properties for MgCO3, updating the PERCS model, and using a more robust PWR plant model.


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