Comparisons between passive RCCSS on degree of passive safety features against accidental conditions and methodology to determine structural thickness of scaled-down heat removal test facilities

2021 ◽  
Vol 162 ◽  
pp. 108512
Author(s):  
Kuniyoshi Takamatsu ◽  
Tatsuya Matsumoto ◽  
Wei Liu ◽  
Koji Morita
Author(s):  
Richard F. Wright ◽  
James R. Schwall ◽  
Creed Taylor ◽  
Naeem U. Karim ◽  
Jivan G. Thakkar ◽  
...  

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.


Author(s):  
N. Ueda ◽  
I. Kinoshita ◽  
Y. Nishi ◽  
A. Minato ◽  
H. Matsumiya ◽  
...  

This paper describes the passive safety features utilized in the updated sodium cooled Super-Safe, Small and Simple fast reactor, which is the improved 4S reactor. This reactor can operate up to ten years without refueling and neutron reflector regulates the reactivity. One of the design requirements is to secure the core against all anticipated transients without reactor scram. Therefore, the reactor concept is to design to enhance the passive safety features. All temperature reactivity feedback coefficients including whole core sodium void worth are negative. Also, introducing of RVACS (Reactor Vessel Auxiliary Cooling System) can enhance the passive decay heat removal capability. Safety analyses are carried out to simulate various transient sequences, which are loss of flow events, transient overpower events and loss of heat sink events, in order to evaluate the passive safety capabilities. A calculation tool for plant dynamics analyses for fast reactors has been modified to model the 4S including the unique plant system, which are reflector control system, circulation pumps and RVACS. The analytical results predict that the designed passive features improve the safety in which temperature variation in transients are satisfied with the safety criteria for the fuel element and the structure of the primary coolant boundary.


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

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.


1992 ◽  
Vol 99 (3) ◽  
pp. 318-329 ◽  
Author(s):  
Hiroshi Endo ◽  
Yoshio Kumaoka ◽  
Simcha Golan ◽  
Hiroshi Nakagawa

Author(s):  
Chul-Hwa Song ◽  
Tae-Soon Kwon ◽  
Byong-Jo Yun ◽  
Ki-Yong Choi ◽  
Hwan-Yeol Kim ◽  
...  

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.


Author(s):  
Anwar Hussain ◽  
Ammar Khan

The aim of this paper is to summarize the research that has already been conducted in the field of PWR technology which utilizes TRISO fuel (Tri-Structural Isotropic) instead of standard PWR fuel. TRISO fuel which in mainly used in HTGRs has been used in SMRs. This type of reactor can be a potential source of energy for a limited power production, heating and desalination facilities at remote location. Combination of TRISO fuel and PWR technology results in a reliable and safe NPP design. Due to enhanced and outstanding passive safety features advanced SMRs are being studied around the world and substantial amount of conceptual research has already been conducted in this field. However, further research is required to be undertaken to optimize these design concepts. Finally, few areas have been pointed out for further research to overcome the challenges and gaps in this research field.


Author(s):  
Fatih Aydogan ◽  
Geoffrey Black ◽  
Meredith A. Taylor Black ◽  
David Solan

In recent years, several small modular reactor (SMR) designs have been developed. These nuclear power plants (NPPs) not only offer a small power size (less than 300 MWe), a reduced spatial footprint, and modularized compact designs fabricated in factories and transported to the intended sites, but also passive safety features. Some light water (LW)-SMRs have already been granted by Department of Energy: NuScale and mPower. New LW-SMRs are mainly inspired by the early LW-SMRs (such as process-inherent ultimate safety (PIUS), international reactor innovative and secure (IRIS), and safe integral reactor (SIR)). LW-SMRs employ significantly fewer components to decrease costs and increase simplicity of design. However, new physical challenges have appeared with these changes. At the same time, advanced SMR (ADV-SMR) designs (such as PBMR, MHR Antares, Prism, 4S, and Hyperion) are being developed that have improved passive safety and other features. This paper quantitatively and qualitatively compares most of the LW- and ADV-SMRs with respect to reactors, nuclear fuel, containment, reactor coolant systems, refueling, and emergency coolant systems. Economic and financing evaluations are also included in the paper. The detailed comparisons in this paper elucidate that one reactor is not superior to the others analyzed in this study, as each reactor is designed to meet different needs.


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