Improvement of heat-removal capability using heat conduction on a novel reactor cavity cooling system (RCCS) design with passive safety features through radiation and natural convection

2018 ◽  
Vol 122 ◽  
pp. 201-206 ◽  
Author(s):  
Kuniyoshi Takamatsu ◽  
Tatsuya Matsumoto ◽  
Wei Liu ◽  
Koji Morita
Keyword(s):  
Heat Conduction ◽  
Natural Convection ◽  
Cooling System ◽  
Heat Removal ◽  
Passive Safety ◽  
Safety Features ◽  
Cavity Cooling

scholarly journals REACTOR CAVITY COOLING SYSTEM WITH PASSIVE SAFETY FEATURES ON RDE: THERMAL ANALYSIS DURING ACCIDENT

2019 ◽  
Vol 21 (2) ◽  
pp. 87
Author(s):  
Rahayu Kusumastuti ◽  
Sriyono Sriyono ◽  
Mulya Juarsa ◽  
Hendro Tjahjono ◽  
I. D. Irianto ◽  
...  
Keyword(s):  
Thermal Analysis ◽  
Natural Convection ◽  
Cooling System ◽  
Power Level ◽  
Heat Removal ◽  
Reactor Vessel ◽  
Normal Operation ◽  
Passive Safety ◽  
Diesel Generator ◽  
Cavity Cooling

Reaktor Daya Eksperimental (RDE) is an experimental power reactor based on HTGR technology that implements inherent safety system. Its safety systems are in compliance with “defense in depth” philosophy. RDE is also equipped with reactor cavity cooling system (RCCS) used to remove the heat transferred from the reactor vessel to the containment structure. The RCCS is designed to fulfil this role by maintain the reactor vessel under the maximum allowable temperature during normal operation and protecting the containment structure in the event of failure of all passive cooling systems. The performance and reliability of the RCCS, therefore, are considered as critical factors in determining maximum design power level related to heat removal. RCCS for RDE will use a novel shape to efficiently remove the heat released from the RPV through thermal radiation and natural convection. This paper discusses the calculation of RCCS thermal analysis during accident. The RPV temperature must be maintained below 65ºC. The accident is assumed that there is no electricity from diesel generator supplied to the blower. The methodology used is based on the calculation of mathematical model of the RCCS in the passive mode. The heat is released through cavity by natural convection, in which the RCCS is capable to withdraw the heat at the rate of 50.54 kW per hour.Keywords: Passive safety, RCCS, RDE, Thermal analysis

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 in Sodium Cooled Super-Safe, Small and Simple Reactor

2002 ◽  
Author(s):  
N. Ueda ◽  
I. Kinoshita ◽  
Y. Nishi ◽  
A. Minato ◽  
H. Matsumiya ◽  
...  
Keyword(s):  
Cooling System ◽  
Heat Removal ◽  
Reactor Vessel ◽  
Passive Safety ◽  
Primary Coolant ◽  
Decay Heat ◽  
The Core ◽  
Design Requirements ◽  
Safety Features ◽  
Dynamics Analyses

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.

Passive Safety Systems of Advanced Nuclear Power Plant: AP1000

2010 ◽  
Author(s):  
Guohua Yan ◽  
Chen Ye
Keyword(s):  
Nuclear Power ◽  
Human Error ◽  
Cooling System ◽  
Heat Removal ◽  
Injection System ◽  
Passive Safety ◽  
Human Errors ◽  
Power Sources ◽  
Passive Safety Systems ◽  
Safety Systems

In the entire history of commercial nuclear power so far, only two major accidents leading to damage of reactor core have taken place. One is Three Mile Island (TMT) accident (1979), which is caused by a series of human error, and the other is Chernobyl accident (1986), which is due to the combined reason of design defects and human errors. After TMI and Chernobyl accidents, in order to reduce manpower in operation and maintenance and influence of human errors on reactor safety, consideration is given to utilization of passive safety systems. According to the IAEA definition, passive safety systems are based on natural forces, such as convection and gravity, and stored energy, making safety functions less dependent on active systems and operators’ action. Recently, the technology of passive safety has been adopted in many reactor designs, such as AP1000, developed by Westinghouse and EP1000 developed by European vendor, and so on. AP1000 as the first so-called Generation III+ has received the final design approval from US NRC in September 2004, and now being under construction in Sanmen, China. In this paper, the major passive safety systems of AP1000, including passive safety injection system, automatic depressurization system passive residual heat removal system and passive containment cooling system, are described and their responses to a break loss-of-coolant accident (LOCA) are given. Just due to these passive systems’ adoption, the nuclear plant can be able to require no operator action and offsite or onsite AC power sources for at least 72h when one accident occurs, and the core melt and large release frequencies are significantly below the requirement of operating plants and the NRC safety goals.

Simulation and Analysis of Helium Circulator Trip ATWS Test at Full Power on the HTR-10

2008 ◽  
Author(s):  
Yujie Dong ◽  
Fubing Chen ◽  
Zuoyi Zhang ◽  
Shouyin Hu ◽  
Lei Shi ◽  
...  
Keyword(s):  
High Temperature ◽  
Cooling System ◽  
Power Level ◽  
Reactor Core ◽  
Heat Removal ◽  
Reactor Power ◽  
Code System ◽  
Safety Features ◽  
Forced Cooling ◽  
High Temperature Gas

Safety demonstration tests on the 10 MW High Temperature Gas-cooled Reactor-Test Module (HTR-10) were conducted to verify the inherent safety characteristics of modular High Temperature Gas-cooled Reactors (HTGRs) as well as to obtain the reactor core and primary cooling system transient data for validation of HTGR safety analysis models and codes. As one of these safety demonstration tests, a simulated anticipated transient without scram (ATWS) test called loss of forced cooling by tripping the helium circulator without reactor scram was carried out at 100% rated power level in July, 2005. This paper simulates the reactor transient behaviour during the test by using the THERMIX code system. The reactor power transition and a comparison with the test result are presented. Owing to the negative temperature coefficient of reactivity, the reactor undergoes a self-shutdown after the stop of the helium circulator and keeps subcritical till the end of the test. Due to the loss of forced cooling, the residual heat is slowly transferred from the core to the Reactor Cavity Cooling System (RCCS) by conduction, radiation and natural convection. The thermal response of this heat removal process is investigated. The calculated and test temperature transients of the measuring points in the reactor internals are given and the differences are preliminarily discussed. With respect to the safety features of the HTR-10, it is of most importance that the maximum fuel center temperature is always lower than 1230 °C which is the limited value at the first phase of the HTR-10 project. The simulation and test results show that the HTR-10 has the built-in passive safety features, and the THERMIX code system is applicable and reasonable for simulating and analyzing the helium circulator trip ATWS test.

0617 A Fundamental Study on heat removal by natural convection for a waste storage cooling system

2007 ◽  
Vol 2007.3 (0) ◽  
pp. 155-156
Author(s):  
Hiroshi IKEDA ◽  
Daichi SUZUKI ◽  
Kotaro NAKADA ◽  
Chikako IWAKI ◽  
Yoshikazu MASAKI ◽  
...  
Keyword(s):  
Natural Convection ◽  
Cooling System ◽  
Heat Removal ◽  
Fundamental Study ◽  
Waste Storage
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