Thermal-Hydraulic R&Ds for the APR+ Developments in Korea

2010 ◽  
Author(s):  
Chul-Hwa Song ◽  
Tae-Soon Kwon ◽  
Byong-Jo Yun ◽  
Ki-Yong Choi ◽  
Hwan-Yeol Kim ◽  
...  
Keyword(s):  
Power Reactor ◽  
Severe Accident ◽  
Injection System ◽  
Passive Safety ◽  
Paper Briefly ◽  
Standard Design ◽  
Fluidic Device ◽  
System 2 ◽  
The Republic ◽  
Safety Features

This paper briefly introduces recent progress in thermal-hydraulic R&Ds, which is mainly being performed at KAERI, for the APR+ (Advanced Power Reactor plus) development. The main R&D items for the APR+ reactor are associated directly with recent efforts to introduce new safety concepts in the APR+ standard design developments, which is currently in progress in the Republic of Korea. The R&D activities reported here mainly cover the thermal-hydraulic and severe accident areas and are being performed in experimental and/or analytical ways. They include: (1) advancement and optimization of safety injection system, (2) incorporation of passive safety features, such as advanced Fluidic Device (FD+) and passive auxiliary feedwater system (PAFS), and (3) incorporation of severe accident mitigation features.

NSSS Design Features of Advanced Power Reactor Plus (APR+)

2010 ◽  
Author(s):  
Hae Yun Choi ◽  
Kwang Won Lee ◽  
Jong Tae Seo
Keyword(s):  
Nuclear Power ◽  
Power Reactor ◽  
Regulatory Body ◽  
Pressurized Water Reactor ◽  
Design Features ◽  
Pressurized Water ◽  
Standard Design ◽  
Fluidic Device ◽  
New Construction ◽  
And Performance

A feasibility study for developing Advanced Power Reactor Plus (APR+), an improved nuclear power unit to succeed to the Korea’s current Advanced Power Reactor 1400 MWe (APR1400), has been carried out for 2 years from August 2007 to July 2009. The major goals of this study are to identify top-tier requirements for the candidate APR+, to develop a preliminary design concept, and to evaluate conceptual design in terms of safety, economics, and performance characteristics. From this study, it is presumed that the APR+ can be developed as a two loop evolutionary pressurized water reactor with a number of advanced design features to enhance safety and economics based on the APR1400 technology. For economic enhancement, the APR+ core power has increased up to 4,290 MWth which corresponds to a 1500MWe class nuclear power plant. Several new construction technologies are introduced so as to shorten construction period. For safety enhancement, several advanced design features have been proposed in APR+ design such as an improved direct vessel injection (DVI+), an advanced fluidic device (FD+), a passive auxiliary feedwater system (PAFS), and a mechanical and electrical four train safety concept based on N+2 design philosophy. As accident mitigation features are improved, the in-vessel retention through external reactor vessel cooling (IVR-ERVC) system will be incorporated. The standard design of APR+ is in progress and expected to acquire the standard design approval from the Korean nuclear regulatory body by the end of 2012.

scholarly journals Simulation of the Westinghouse AP1000 Response to SBLOCA Using RELAP/SCDAPSIM

2014 ◽  
Vol 2014 ◽  
pp. 1-9 ◽  
Author(s):  
Ayah Elshahat ◽  
Timothy Abram ◽  
Judith Hohorst ◽  
Chris Allison
Keyword(s):  
Driving Forces ◽  
Natural Circulation ◽  
Safety System ◽  
Operating Conditions ◽  
Severe Accident ◽  
Passive Safety ◽  
System Component ◽  
Pressurized Water ◽  
Loss Of Coolant Accident ◽  
Accident Conditions

Great interest is given now to advanced nuclear reactors especially those using passive safety components. The Westinghouse AP1000 Advanced Passive pressurized water reactor (PWR) is an 1117 MWe PWR designed to achieve a high safety and performance record. The AP1000 safety system uses natural driving forces, such as pressurized gas, gravity flow, natural circulation flow, and convection. In this paper, the safety performance of the AP1000 during a small break loss of coolant accident (SBLOCA) is investigated. This was done by modelling the AP1000 and the passive safety systems employed using RELAP/SCDAPSIM code. RELAP/SCDAPSIM is designed to describe the overall reactor coolant system (RCS) thermal hydraulic response and core behaviour under normal operating conditions or under design basis or severe accident conditions. Passive safety components in the AP1000 showed a clear improvement in accident mitigation. It was found that RELAP/SCDAPSIM is capable of modelling a LOCA in an AP1000 and it enables the investigation of each safety system component response separately during the accident. The model is also capable of simulating natural circulation and other relevant phenomena. The results of the model were compared to that of the NOTRUMP code and found to be in a good agreement.

Thermal hydraulic challenges of Gas Cooled Fast Reactors with passive safety features

2009 ◽  
Vol 239 (5) ◽  
pp. 840-854 ◽  
Author(s):  
Michael A. Pope ◽  
Jeong Ik Lee ◽  
Pavel Hejzlar ◽  
Michael J. Driscoll
Keyword(s):  
Passive Safety ◽  
Fast Reactors ◽  
Safety Features

Level 2 PSA Overview of HPR1000 Nuclear Power Plant

2020 ◽  
Author(s):  
Wang Ziguan ◽  
Lu Fang ◽  
Yang Benlin ◽  
Chen Shi ◽  
Hu Lingsheng
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Heat Removal ◽  
Accident Prevention ◽  
Severe Accident ◽  
Mitigation Measures ◽  
Injection System ◽  
Risk Level ◽  
Level 2

Abstract Risk-informed design approaches are comprehensively implemented in the design and verification process of HPR1000 nuclear power plant. Particularly, Level 2 PSA is applied in the optimization of severe accident prevention and mitigation measures to avoid the extravagant redundancy of system configurations. HPR1000 preliminary level 2 PSA practices consider internal events of the reactor in the context of at-power condition. Severe accidents mitigation and prevention system and its impact on the overall large release frequency (LRF) level are evaluated. The results showed that severe accident prevention and mitigation systems, such as fast depressurization system, the cavity injection system and the passive containment heat removal system perform well in reducing LRF and overall risk level of HPR1000 NPP. Bypass events, reactor rapture events, and the containment bottom melt-through induced by MCCI are among the dominant factors of the LRF. The level 2 PSA analysis results indicate that HPR1000 design is reliable with no major weaknesses.

Application of Level 2 PSA in the Design of Cavity Injection System for Nuclear Power Plant

2018 ◽  
Author(s):  
Wentao Zhu ◽  
Wenjing Li
Keyword(s):  
Decision Making ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Assessment Tool ◽  
Quantitative Risk Assessment ◽  
Severe Accident ◽  
Injection System ◽  
Retention Strategy ◽  
Level 2

After Fukushima nuclear power plant accident, severe accident is getting more and more concerns all over the world. In order to mitigate severe accident and improve the safety of nuclear power plant, two different strategies are applied in different plants. One is in-vessel melt retention strategy, and the other is ex-vessel melt retention strategy. Tianwan nuclear power plant is an improved Gen II nuclear power plant and in-vessel melt retention strategy is adopted in the plant. In order to achieve this strategy, cavity injection system is designed for the plant. Probabilistic Safety Analysis is the most commonly used quantitative risk assessment tool for decision-making in selecting the optimal design among alternative options. For this plant, in order to optimize the design of cavity injection system, improve the safety level of nuclear power plant, and meanwhile, improve the engineering implementation and economization, Level 2 PSA was used for this decision-making process. In this paper, the Level 2 PSA for this plant and the application for the design of cavity injection system are introduced.

Simulation of Hydrogen Distribution due to In-Vessel Severe Accident in WWER-1000 NPP Containment: A Comparison of CONTAIN and MELCOR Codes Results

2018 ◽  
Author(s):  
Omid Noorikalkhoran ◽  
Massimiliano Gei
Keyword(s):  
Severe Accident ◽  
Second Step ◽  
Hydrogen Distribution ◽  
Short Term ◽  
Long Time ◽  
Design Basis Accident ◽  
Safety Features ◽  
Design Pressure ◽  
Good Agreement

During a severe accident or Beyond Design Basis Accident (BDBA), the reaction of water with zirconium alloy as fuel clad, radiolysis of water, molten corium-concrete interaction (MCCI) and post-accident corrosion can generate a source of hydrogen. In the present work, hydrogen distribution due to in-vessel reaction (between zircaloy and steam) has been simulated inside a WWER-1000 reactor containment. In the first step, the thermal hydraulic parameters of containment have been simulated for a DECL (Double Ended Cold Leg) accident (DBA phase) in both short and long time and the effects of spray as Engineering Safety Features (ESFs) on mitigating the parameters have been studied. In the second step, it has been assumed that the accident developed into an in-vessel core melting accident. While in pre-phase of core melting (severe accident phase), hydrogen will be produced as a result of zircaloy and steam reaction (BDBA phase), the hydrogen distribution has been simulated for 23 cells inside the reactor containment by using CONTAIN 2.0 (Best estimate code) and MELCOR 1.8.6 codes. Finally, the results have been compared to FSAR results. As it can be seen from the comparisons, both CONTAIN and MELCOR codes can predict the results in good agreement with FSAR (ANGAR code) results. CONTAIN shows peak pressure around 0.36 MPa in short-term and this amount is about 0.38 and 0.4 MPa for MELCOR and ANGAR (FSAR) results respectively. All these values are under design pressure that is around 0.46 MPa. Cell 20 has the maximum mole fraction of hydrogen in long-term about 9.5% while the maximum amount of hydrogen takes place in cell 22. The differences between the results of codes are because of different equations, Models, Numerical methods and assumptions that have been considered by the codes. The simulated Hydrogen Distribution Map (HDM) can be used for upgrading the location of HCAV systems and Hydrogen Mitigator features (like the recombiners and ignitors) inside the containment to reduce the risk of hydrogen explosion.

Preliminary LOCA Analysis of Heating-Reactor of Advanced Low-Pressurized and Passive Safety System (HAPPY)

2018 ◽  
Author(s):  
Mian Xing ◽  
Zhaocan Meng ◽  
Xiaotao Liao ◽  
Canhui Sun ◽  
Shuming Zhang ◽  
...  
Keyword(s):  
Conceptual Design ◽  
District Heating ◽  
Central Research Institute ◽  
Safety System ◽  
Passive Safety ◽  
Central Research ◽  
Pressurized Water ◽  
Design And Optimization ◽  
Passive Safety System ◽  
Safety Features

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.

AP1000 Passive Residual Heat Removal Heat Exchanger Confirmatory Analysis

2006 ◽  
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.

Passive Safety Features of a Bottom-Supported Fast Breeder Reactor Vessel

1992 ◽  
Vol 99 (3) ◽  
pp. 318-329 ◽  
Author(s):  
Hiroshi Endo ◽  
Yoshio Kumaoka ◽  
Simcha Golan ◽  
Hiroshi Nakagawa
Keyword(s):  
Reactor Vessel ◽  
Passive Safety ◽  
Fast Breeder Reactor ◽  
Safety Features
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