Preliminary Large Break LOCA and SGTR Accident Analysis of the Compact Small Reactor

2017 ◽  
Author(s):  
Mian Xing ◽  
Linsen Li ◽  
Feng Shen ◽  
Xiao Hu ◽  
Zhan Liu ◽  
...  
Keyword(s):  
Cooling System ◽  
Thermal Power ◽  
Reactor Core ◽  
Safety System ◽  
Passive Safety ◽  
Primary System ◽  
The Core ◽  
Passive Safety System ◽  
Safety Margins ◽  
Guideline Values

This paper gives a brief introduction of the Compact Small Reactor (CSR). It is a simplified two-loop reactor with thermal power of 660MW and with compact primary system and passive safety feature. Preliminary safety analysis of the CSR is conducted to evaluate and further optimize the design of passive safety system, especially the passive core cooling system. Large Break Loss Of Coolant Accident (LBLOCA) and Steam Generator Tube Rupture (SGTR) are selected as two reference accidental scenarios. Each scenario is modeled and computed by RELAP5/MOD3.4. For the LBLOCA analysis, a guillotine break happens in the cold leg of the loop containing the core makeup tanks balance lines. The results show certain safety margins from the guideline values, and the passive safety system could supply enough cooling of the core. For the SGTR analysis, the results show the robustness of the design from the safety perspective. It is concluded that the safety systems are capable of mitigating the accidents and protecting the reactor core from severe damage.

Investigation of effects of nonavailability of passive safety systems on the reactor behaviour during LOCA Scenario in AP600

Kerntechnik ◽  
2021 ◽  
Vol 86 (3) ◽  
pp. 244-255
Author(s):  
S. H. Abdel-Latif ◽  
A. M. Refaey
Keyword(s):  
Passive Safety ◽  
Primary System ◽  
Water Storage Tank ◽  
The Core ◽  
Passive Safety Systems ◽  
Loss Of Coolant Accident ◽  
Passive Safety System ◽  
Stage 4 ◽  
Core Cooling ◽  
Safety Systems

Abstract The AP600 is a Westinghouse Advanced Passive PWR with a two–loop 1 940 MWt. This reactor is equipped with advanced passive safety systems which are designed to operate automatically at desired set-points. On the other hand, the failure or nonavailability to operate of any of the passive safety systems may affect reactor safety. In this study, modeling and nodalization of primary and secondary loops, and all passive reactor cooling systems are conducted and a 10-inch cold leg break LOCA is analyzed using ATHLET 3.1A Code. During loss of coolant accident in which the passive safety system failure or nonavailability are considered, four different scenarios are assumed. Scenario 1 with the availability of all passive systems, scenario 2 is failure of one of the accumulators to activate, scenario 3 is without actuation of the automatic depressurization system (ADS) stages 1–3, and scenario 4 is without actuation of ADS stage 4. Results indicated that the actuation of passive safety systems provide sufficient core cooling and thus could mitigate the accidental consequence of LOCAs. Failure of one accumulator during LOCA causes early actuation of ADS and In-Containment Refueling Water Storage Tank (IRWST). In scenario 3 where the LOCA without ADS stages 1–3 actuations, the depressurization of the primary system is relatively slow and the level of the core coolant drops much earlier than IRWST actuation. In scenario 4 where the accident without ADS stage-4 activation, results in slow depressurization and the level of the core coolant drops earlier than IRWST injection. During the accident process, the core uncovery and fuel heat up did not happen and as a result the safety of AP600 during a 10-in. cold leg MBLOCA was established. The relation between the cladding surface temperature and the primary pressure with the actuation signals of the passive safety systems are compared with that of RELAP5/Mode 3.4 code and a tolerable agreement was obtained.

The Development of the Evolutionary BWR (AB1600)

2008 ◽  
Author(s):  
Akira Murase ◽  
Mikihide Nakamaru ◽  
Ryoichi Hamazaki ◽  
Masahiko Kuroki ◽  
Munetaka Takahashi
Keyword(s):  
Fuel Cycle ◽  
Safety System ◽  
Safety Performance ◽  
Passive Safety ◽  
Main Role ◽  
Core Diameter ◽  
The Core ◽  
Passive Safety System ◽  
Bundle Size ◽  
Near Term

Considering the delay of the first breeding reactor (FBR), it is expected that the light water reactor will still play the main role of the electric power generation in the 2030’s. Accordingly, Toshiba has been developing a new conceptual ABWR as the near-term BWR. We tentatively call it AB1600. The AB1600 has introduced the hybrid active/passive safety system in order to have independent countermeasure for severe accidents and better probability of core damage frequency (CDF) considered external events such as earthquake. On the other hand, we have another goal of the AB1600, which is to retain the safety performance superior or equivalent to the current ABWR without deterioration of economy. In order to achieve both economy and safety performance, we have optimized the safety system configuration of the AB1600 by partly introducing passive safety system to design basis event (DBEs). At the same time, we have adopted the simplification of the overall plant systems in order to improve economy. In order to reduce capital cost, to shorten refueling period and to reduce maintenance effort, the AB1600 introduces the large fuel bundle size. The bundle size is 1.2 times as large as that of the ABWR and the fuel rod array is 12 by 12. And then by progressing the core design, we can reduce the number of reactor internal pumps (RIPs) to eight from the current ABWR of ten. The core power density, the number of fuel bundles, and the core diameter of AB1600 are decided in order to achieve 24 months fuel cycle length on the condition with below 5wt% enrichment of fuel and with eight RIPs.

Test Analysis of CAP1400 5cm Break Accident Based on the ACME Test Facility

2016 ◽  
Author(s):  
Sheng Zhu
Keyword(s):  
System Design ◽  
Nuclear Power ◽  
Reactor Core ◽  
Safety System ◽  
Test Facility ◽  
Passive Safety ◽  
Pressurized Water ◽  
Power Simulation ◽  
Passive Safety System ◽  
Core Cooling

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.

The Comparative Advantages of CAP1400 Nuclear Power Plant Passive Safety System

2017 ◽  
Author(s):  
Ye Cheng ◽  
Wang Minglu ◽  
Qiu Zhongming ◽  
Wang Yong
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Cooling System ◽  
Assessment Method ◽  
Safety System ◽  
Passive Safety ◽  
Passive Safety System ◽  
First Choice ◽  
Comparative Advantages

With the demand for nuclear power increasing, the first choice of almost all countries who want to build a new nuclear power plant is to use generation III technology, primarily because the safety of generation III technology is greatly improved compared with that of generation II and II + technology. The passive safety technology was introduced by the AP1000 and is one of the best applications of generation III technologies. In this study, the representative passive containment cooling system of the CAP1400 (developed based on AP1000) and the containment spray system of a generation II nuclear power plant are compared and analyzed using the Probabilistic Safety Assessment method. The reasons why a passive safety system has comparative advantages are determined by concrete calculations.

Assessment of passive safety system performance under gravity driven cooling system drain line break accident

2014 ◽  
Vol 74 ◽  
pp. 136-142 ◽  
Author(s):  
J. Lim ◽  
J. Yang ◽  
S.W. Choi ◽  
D.Y. Lee ◽  
S. Rassame ◽  
...  
Keyword(s):  
System Performance ◽  
Cooling System ◽  
Safety System ◽  
Passive Safety ◽  
Passive Safety System ◽  
Gravity Driven

scholarly journals The Investigation of Nonavailability of Passive Safety Systems Effects on Small Break LOCA Sequence in AP1000 Using RELAP5 MOD 4.0

2016 ◽  
Vol 2016 ◽  
pp. 1-11 ◽  
Author(s):  
Anwar Hussain ◽  
Amjad Nawaz
Keyword(s):  
Active Region ◽  
Reactor Core ◽  
Passive Safety ◽  
Primary System ◽  
Passive Safety Systems ◽  
Passive Safety System ◽  
Stage 4 ◽  
Loss Of Coolant ◽  
Core Cooling ◽  
Safety Systems

The passive safety systems of AP1000 are designed to operate automatically at desired set-points. However, the unavailability or failure to operate of any of the passive safety systems will change the accident sequence and may affect reactor safety. The analysis in this study is based on some hypothetical scenarios, in which the passive safety system failure is considered during the loss of coolant accidents. Four different cases are assumed, that is, with all passive systems, without actuation of one of the accumulators, without actuation of ADS stages 1–3, and without actuation of ADS stage 4. The actuation of all safety systems at their actuation set-points provides adequate core cooling by injecting sufficient water inventory into reactor core. The LOCA with actuation of one of the accumulators cause early actuation of ADS and IRWST. In case of LOCA without ADS stages 1–3, the primary system depressurization is relatively slow and mixture level above core active region drops much earlier than IRWST actuation. The accident without ADS stage 4 actuation results in slow depressurization and mixture level above core active region drops earlier than IRWST injection. Moreover, the comparison of cladding surface temperature is performed in all cases considered in this work.

Survey of Research Activities for Passive Safety System of AC600

1999 ◽  
Author(s):  
Bingde Chen ◽  
Zhumao Yang ◽  
Fuyun Ji
Keyword(s):  
Cooling System ◽  
Program Planning ◽  
Experimental Studies ◽  
Heat Removal ◽  
Safety System ◽  
Injection System ◽  
Passive Safety ◽  
Design Modification ◽  
Research Activities ◽  
Passive Safety System

Abstract The use of passive safety system in AC600, the Chinese advanced 600 MWe PWR proposed by NPIC, together with other improvements, such as simplification and advanced I&C etc., makes the plant more safe, economic and reliable. The core damage frequency (CDF) decreases from less than 10−4 of conventional PWR to less than 10−5 to 10−6 and the plant available factor increases to ∼90%. The passive safety system of AC600 consists of three complete independent systems. They are passive containment cooling system (passive CC system), passive core residual heat removal system (passive CRHR system) and passive safety injection system (CMT). To verify and demonstrate the AC600’s innovative passive safety features and to obtain an experimental database for system design modification and optimizing, and for computer code development and assessment, the experimental studies on these systems were finished in NPIC during the eighth national Five Year period under the national support. In this paper, the experimental research activities on passive containment cooling system, passive CRHR system and CMT injection system, including test rigs and main results are summarized. These experiments proved the design of all these passive systems are feasible and reliable and can meet basically the required safety functions. Some undesired thermal hydraulic phenomena, for example, “water hammer”, which may have bad impacts on its safety functions and to which high attention should be given, was found and identified in these studies. All data obtained have already been used in the design improvement and next R&D program planning.

Sign in / Sign up

Export Citation Format

Share Document