AP1000® Passive Cooling Containment Analysis of a Double-Ended LBLOCA With a 3D Gothic Model

2018 ◽  
Author(s):  
Samanta Estevez-Albuja ◽  
Gonzalo Jimenez ◽  
Kevin Fernández-Cosials ◽  
César Queral ◽  
Zuriñe Goñi
Keyword(s):  
Cooling System ◽  
Heat Removal ◽  
Water Tank ◽  
Local Pressure ◽  
Water Storage Tank ◽  
Passive Safety Systems ◽  
Containment Model ◽  
Loss Of Coolant Accident ◽  
Ac Power ◽  
Safety Systems

In order to enhance Generation II reactors safety, Generation III+ reactors have adopted passive mechanisms for their safety systems. In particular, the AP1000® reactor uses these mechanisms to evacuate heat from the containment by means of the Passive Containment Cooling System (PCS). The PCS uses the environment atmosphere as the ultimate heat sink without the need of AC power to work properly during normal or accidental conditions. To evaluate its performance, the AP1000 PCS has been usually modeled with a Lumped Parameters (LP) approach, coupled with another LP model of the steel containment vessel to simulate the accidental sequences within the containment building. However, a 3D simulation, feasible and motivated by the current computational capabilities, may be able to produce more detailed and accurate results. In this paper, the development and verification of an integral AP1000® 3D GOTHIC containment model, taking into account the shield building, is briefly presented. The model includes all compartments inside the metallic containment liner and the external shield building. Passive safety systems, such as the In-containment Refueling Water Storage Tank (IRWST) with the Passive Residual Heat Removal (PRHR) heat exchanger and the Automatic Depressurization System (ADS), as well as the PCS, are included in the model. The model is tested against a cold leg Double Ended Guillotine Break Large Break Loss of Coolant Accident (DEGB LBLOCA) sequence, taking as a conservative assumption that the PCS water tank is not available during the sequence. The results show a pressure and temperature increase in the containment in consonance with the current literature, but providing a greater detail of the local pressure and temperature in all compartments.

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.

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.

Safety Assessment and Accident Analysis of VVER-1000/466B With Active and Passive Safety Systems for Belene NPP

2014 ◽  
Author(s):  
Luben Sabotinov ◽  
Borislav Dimitrov ◽  
Giovanni B. Bruna
Keyword(s):  
Cooling System ◽  
Safety Evaluation ◽  
Safety Issues ◽  
Passive Safety Systems ◽  
Loss Of Coolant Accident ◽  
Depth Analysis ◽  
Active Safety Systems ◽  
Accident Scenarios ◽  
Core Cooling ◽  
Safety Systems

The paper presents the methodology adopted to assess the Interim Safety Analysis Report (ISAR) of the Belene NPP in the framework of the contract between the Bulgarian Nuclear Regulatory Authority (BNRA) and RISKAUDIT (IRSN&GRS). It stresses the in-depth analysis carried-out for several relevant-to-safety issues and illustrates in some detail the investigation of the Large Break Loss of Coolant Accident (LB LOCA) with loss of power and failure of the active part of the Emergency Core Cooling System (High Pressure and Low Pressure Safety Injection pumps), performed with the French best estimate thermal-hydraulic code CATHARE. The role, problems and efficiency of the passive and active safety systems during the accident scenarios are discussed. Finally, the main conclusions of the safety evaluation of the Belene NPP project are summarized.

Study of Decay Heat Removal Approach for Advanced Passive PWR During Station Blackout

2014 ◽  
Author(s):  
Jie Zou ◽  
Lili Tong ◽  
Xuewu Cao
Keyword(s):  
Cooling System ◽  
Heat Removal ◽  
Safety System ◽  
Injection System ◽  
Passive Safety ◽  
System Failure ◽  
Decay Heat ◽  
Passive Safety Systems ◽  
Station Blackout ◽  
Safety Systems

After Fukushima accident, decay heat removal in station blackout (SBO) accident is concerned for different NPP design. Advanced passive PWR relies on passive systems to cool reactor core and containment, such as the passive residual heat removal system (PRHR), passive injection system and passive containment cooling system (PCCS). Passive safety systems are considered more reliable than traditional active safety system under accident condition. However, in long-term SBO situation, possible failure of passive safety systems is noticed as active valves are needed in system actuation. Moreover, probability safety analysis results of advanced passive PWR show that system failure is possible without external event. Given different passive safety system failure assumptions, response of reactor coolant system and containment of advanced passive PWR is calculated in SBO accident, the integrity of core, reactor pressure vessel and containment is assessed, and decay heat removal approach is studied. The results show that containment failure is predicted with the failure of PCCS and PRHR, reactor vessel failure together with containment failure is predicted with the failure of PCCS, passive injection system and PRHR. Advices to deal with the risk of advanced passive PWR in SBO are given based on the study.

Development of Passive Cooling System Models for the Assessment of the Long Term Cooling Capability in the Liquid Metal Reactor

2004 ◽  
Author(s):  
W. P. Chang ◽  
K. S. Ha ◽  
H. Y. Jeong ◽  
S. Heo ◽  
Y. B. Lee
Keyword(s):  
Cooling System ◽  
Heat Removal ◽  
Computer Code ◽  
Passive Safety ◽  
Primary Concern ◽  
Metallic Fuel ◽  
Passive Safety Systems ◽  
Operator Action ◽  
Safety Systems

This study has been carried out to assess the decay heat removal capability of the passive safety systems adopted in a conceptual design of the 600 MW(e), sodium cooled, metallic fuel loaded KALIMER. The applicability of the PVCS, which used to be the only passive safety system for KALIMER-150, is limited to a reactor capacity of 1,000 MW(t) or less. Another passive loop, PDRC, is conceptualized in order to overcome the limit as the KALIMER capacity scales up from the current 150 MW(e) to 600 MW(e). The safety analysis computer code, SSC-K, currently used for KALIMER is not capable of simulating such passive systems. With this concern, the PVCS and PDRC models are developed and they are coupled with the SSC-K for a long-term cooling assessment. The present paper thus presents the analysis results of the ULOHS using these models along with their brief introductions. The primary concern of the analyses is focused on the inherent safety as well as the system’s integrity for 72 hours without any operator action during the event.

iB1350: A Generation III.7 Reactor After the Fukushima Daiichi Accident

2016 ◽  
Author(s):  
Takashi Sato ◽  
Keiji Matsumoto ◽  
Kenji Hosomi ◽  
Keisuke Taguchi
Keyword(s):  
Cooling System ◽  
Boiling Water ◽  
Passive Safety ◽  
Active Safety ◽  
Water Reactor ◽  
Airplane Crash ◽  
Passive Safety Systems ◽  
Active Safety Systems ◽  
Fukushima Daiichi ◽  
Safety Systems

iB1350 stands for an innovative, intelligent and inexpensive boiling water reactor 1350. It is the first Generation III.7 reactor after the Fukushima Daiichi accident. It has incorporated lessons learned from the Fukushima Daiichi accident and Western European Nuclear Regulation Association safety objectives. It has innovative safety to cope with devastating natural disasters including a giant earthquake, a large tsunami and a monster hurricane. The iB1350 can survive passively such devastation and a very prolonged station blackout without any support from the outside of a site up to 7 days even preventing core melt. It, however, is based on the well-established proven Advance Boiling Water Reactor (ABWR) design. The nuclear steam supply system is exactly the same as that of the current ABWR. As for safety design it has a double cylinder reinforced concrete containment vessel (Mark W containment) and an in-depth hybrid safety system (IDHS). The Mark W containment has double fission product confinement barriers and the in-containment filtered venting system (IFVS) that enable passively no emergency evacuation outside the immediate vicinity of the plant for a severe accident (SA). It has a large volume to hold hydrogen, a core catcher, a passive flooding system and an innovative passive containment cooling system (iPCCS) establishing passively practical elimination of containment failure even in a long term. The IDHS consists of 4 division active safety systems for a design basis accident, 2 division active safety systems for a SA and built-in passive safety systems (BiPSS) consisting of an isolation condenser (IC) and the iPCCS for a SA. The IC/PCCS pools have enough capacity for 7-day grace period. The IC/PCCS heat exchangers, core and spent fuel pool are enclosed inside the containment vessel (CV) building and protected against a large airplane crash. The iB1350 can survive a large airplane crash only by the CV building and the built-in passive safety systems therein. The dome of the CV building consists of a single wall made of steel and concrete composite. This single dome structure facilitates a short-term construction period and cost saving. The CV diameter is smaller than that of most PWR resulting in a smaller R/B. Each active safety division includes only one emergency core cooling system (ECCS) pump and one emergency diesel generator (EDG). Therefore, a single failure of the EDG never causes multiple failures of ECCS pumps in a safety division. The iB1350 is based on the proven ABWR technology and ready for construction. No new technology is incorporated but design concept and philosophy are initiative and innovative.

Ex-Vessel Loss of Coolant Accident Analysis of ITER Divertor Cooling System Using Modified RELAP/SCADAPSIM/Mod 4.0

2017 ◽  
Vol 3 (4) ◽  
Author(s):  
S. P. Saraswat ◽  
P. Munshi ◽  
A. Khanna ◽  
C. Allison
Keyword(s):  
Heat Flux ◽  
Cooling System ◽  
Heat Removal ◽  
Hydraulic Modeling ◽  
Carbon Fiber Composites ◽  
High Heat Flux ◽  
Vacuum Vessel ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Accident Scenarios

The initial design of ITER incorporated the use of carbon fiber composites in high heat flux regions and tungsten was used for low heat flux regions. The current design includes tungsten for both these regions. The present work includes thermal hydraulic modeling and analysis of ex-vessel loss of coolant accident (LOCA) for the divertor (DIV) cooling system. The purpose of this study is to show that the new concept of full tungsten divertor is able to withstand in the accident scenarios. The code used in this study is RELAP/SCADAPSIM/MOD 4.0. A parametric study is also carried out with different in-vessel break sizes and ex-vessel break locations. The analysis discusses a number of safety concerns that may result from the accident scenarios. These concerns include vacuum vessel (VV) pressurization, divertor temperature profile, passive decay heat removal capability of structure, and pressurization of tokamak cooling water system. The results show that the pressures and temperatures are kept below design limits prescribed by ITER organization.

scholarly journals Modelling of Passive Heat Removal Systems: A Review with Reference to the Framatome BWR Reactor KERENA: Part II

Energies ◽  
2019 ◽  
Vol 13 (1) ◽  
pp. 109 ◽  
Author(s):  
René Manthey ◽  
Frances Viereckl ◽  
Amirhosein Moonesi Shabestary ◽  
Yu Zhang ◽  
Wei Ding ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Gravitational Force ◽  
Heat Removal ◽  
Dimensional System ◽  
Thermal Gradients ◽  
Two Phase ◽  
Passive Safety Systems ◽  
External Power ◽  
Safety Systems

Passive safety systems are an important feature of currently designed and constructed nuclear power plants. They operate independent of external power supply and manual interventions and are solely driven by thermal gradients and gravitational force. This brings up new needs for performance and reliably assessment. This paper provides a review on fundamental approaches to model and analyze the performance of passive heat removal systems exemplified for the passive heat removal chain of the KERENA boiling water reactor concept developed by Framatome. We discuss modeling concepts for one-dimensional system codes such as ATHLET, RELAP and TRACE and furthermore for computational fluid dynamics codes. Part I dealt with numerical and experimental methods for modeling of condensation inside the emergency condenser and on the containment cooling condenser. This second part deals with boiling and two-phase flow instabilities.

The Effect of IRWST Simulation Method on the Analysis of Small Break LOCA Using RELAP5

2017 ◽  
Author(s):  
Caihong Xu ◽  
Guobao Shi ◽  
Kemei Cao ◽  
Xiaoyu Cai ◽  
Zhanfei Qi
Keyword(s):  
Nuclear Power ◽  
Heat Removal ◽  
Simulation Method ◽  
Horizontal Direction ◽  
Water Storage Tank ◽  
Initiation Phase ◽  
Injection Flow ◽  
Loss Of Coolant Accident ◽  
Perfect Mixing ◽  
Core Cooling

The In-containment Refueling Water Storage Tank (IRWST) provides low-pressure safety injection flow for passive CAP1400 Nuclear Power Plant (NPP) during Loss-Of-Coolant-Accident (LOCA) and subsequent Long Term Core Cooling (LTCC). The Passive Residual Heat Removal Heat Exchanger (PRHR HX) and the spargers of Automatic Depressurization System (ADS) stage 1∼3 are submerged in the IRWST. During small break LOCA, heat and mass are delivered through PRHR HX and ADS spargers to IRWST, and IRWST is heated up before its safety injection. However, numerical and experimental investigation has shown that IRWST is not perfect mixing, and thermal stratification exists. During ADS-4/IRWST initiation phase, the temperature of IRWST injection flow is of great importance, and is affected greatly by IRWST simulation method when modeling with system code like RELAP5. In this paper, two different IRWST simulation methods where one use multi channels in horizontal direction while the other use only one, are analyzed for CAP1400 SBLOCA with RE-LAP5, and their effects are compared. Finally, the better method which uses only one channel in horizontal direction is recommended.

scholarly journals Numerical Study of the Impingement of Water Film on a Small Attached Bulging Plate on a Vertical Plane

2021 ◽  
Vol 9 ◽  
Author(s):  
Po Hu ◽  
Zhen Hu
Keyword(s):  
Cooling System ◽  
Numerical Study ◽  
Plane Surface ◽  
Vertical Plane ◽  
Heat Removal ◽  
Water Film ◽  
Loss Ratio ◽  
Water Storage Tank ◽  
Inner Surface ◽  
Condensed Water

In the passive containment cooling system of AP1000, the condensed water is expected to flow down on the inner surface of the steel wall of the containment, and recover to the in-containment refueling water storage tank (IRWST), therefore, to maintain the long-term coolability of the passive residual heat removal system. However, there are attached bulging plates on the inner surface for various engineering needs, such as supporting, and the impingement of condensed water film on these bulging plates can reduce the amount of the recovered water. In this article, a 3-D Eulerian wall film model in FLUENT was used to study a series of flow behaviors when a water film impinged on the bulging plate on a plane surface. The loss ratio of falling film impinging on attached plates of different sizes under different flow rates were calculated and in good agreement with the experiment results. Four stages of the film behavior during the impingement were identified and analyzed; in addition, the influences of the bulging height of attached plate and flow rate were studied. And a correlation between the loss ratio of impinging water film, the bulging height of the attached plate, and the Weber number was obtained.

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