LOCA Break Spectrum Analysis of the Westinghouse Small Modular Reactor

2013 ◽  
Author(s):  
Vefa N. Kucukboyaci ◽  
Jun Liao
Keyword(s):  
Spectrum Analysis ◽  
Evaluation Model ◽  
Reactor Core ◽  
Reactor Vessel ◽  
Passive Safety ◽  
Plant Design ◽  
Pressurized Water ◽  
Passive Safety Systems ◽  
Small Modular Reactor ◽  
Safety Systems

The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (> 225 MWe) integral pressurized water reactor with all primary components, including the steam generator and the pressurizer located inside the reactor vessel. The reactor core is based on a partial-height 17×17 fuel assembly design used in the AP1000® reactor core. The Westinghouse SMR utilizes passive safety systems and proven components from the AP1000 plant design with a compact containment that houses the integral reactor vessel and the passive safety systems. A break spectrum analysis on the Westinghouse SMR LOCA has been performed to investigate the performance of the SMR passive cooling. The break type includes both the double-ended guillotine (DEG) break and the split break with the break size ranging from 0.5 inch to the diameter of direct vessel injection (DVI) line. The break spectrum analysis was performed using the WCOBRA/TRAC-TF2 code, which is designed to simulate PWR LOCA events from the smallest break size to the largest break size. The break spectrum analysis demonstrates that excellent performance of the passive safety system of the Westinghouse SMR in variable LOCA conditions. The study is also a necessary step to develop an evaluation model for the analysis of design basis LOCA accident.

An Overview of the Westinghouse Small Modular Reactor

2011 ◽  
Author(s):  
Robert J. Fetterman ◽  
Alexander W. Harkness ◽  
Matthew C. Smith ◽  
Creed Taylor
Keyword(s):  
Steam Generator ◽  
Pressure Vessel ◽  
Straight Tube ◽  
Plant Design ◽  
Operating Cycle ◽  
Pressurized Water ◽  
Power Sources ◽  
Containment Vessel ◽  
Small Modular Reactor ◽  
Safety Systems

The Westinghouse Small Modular Reactor (SMR) incorporates an integral pressurized water reactor (iPWR) design in which all components associated with the nuclear steam supply system are housed within one pressure vessel. The Westinghouse SMR design also utilizes many of the key features from the AP1000® plant, including passive safety systems. The Westinghouse SMR will be fueled by a derivative of the successful 17×17 Robust Fuel Assembly (RFA) product. An 89 assembly core with an active height of 8 feet will provide a 24 month operating cycle with a power output of 800 MWt. Derived from the AP1000 plant and adapted to operate inside the reactor pressure vessel, 37 control rod drive mechanisms provide reactor shutdown and reactivity control capabilities. Eight seal less pumps provide a nominal reactor coolant flow of 100,000 gallons per minute. An innovative evolution of a straight tube steam generator produces a saturated mixture that is delivered to a steam separating drum located outside of the containment vessel. The steam generator along with the integral pressurizer is attached to the reactor vessel with a single closure flange located near the center of gravity of the reactor assembly and is designed to be removed during refueling operations. Like the AP1000 plant, the Westinghouse SMR relies on the natural forces of gravity and natural circulation to provide core and containment cooling during accident conditions. The passive cooling systems provide sufficient heat removal for seven days without the need for offsite AC power sources. The Westinghouse SMR also includes traditional active components such as diesel generators and pumps; however these components are not required for the safe shutdown of the plant. At a diameter of 32 feet, approximately 25 of the Westinghouse SMR containment vessels can fit within the envelope of the AP1000 containment building. This compact containment will be completely submerged in water during power operation providing a heat sink for postulated accidents. For protection against external threats, the containment vessel and plant safety systems are located below ground level. At approximately one fifth the net electrical output of the AP1000 plant, the Westinghouse SMR is designed to address infrastructure challenges associated with replacing America’s aging fossil fuel plants by providing a safe, clean and reliable energy source. The challenges associated with economies of scale are offset with a compact and simplified plant design, rail shippable components and modular construction.

scholarly journals Role of Passive Safety Systems in Chinese Nuclear Power Development

2009 ◽  
Vol 2009 ◽  
pp. 1-7 ◽  
Author(s):  
X. Cheng ◽  
Y. H. Yang ◽  
Y. Ouyang ◽  
H. X. Miao
Keyword(s):  
Nuclear Power ◽  
Passive Safety ◽  
Pressurized Water Reactors ◽  
Chinese Government ◽  
Power Development ◽  
Pressurized Water ◽  
Passive Safety Systems ◽  
Nuclear Power Development ◽  
Water Reactors ◽  
Safety Systems

Passive safety systems have been widely applied to advanced water-cooled reactors, to enhance the safety of nuclear power plants. The ambitious program of the nuclear power development in China requires reactor concepts with high safety level. For the near-term and medium-term, the Chinese government decided for advanced pressurized water reactors with an extensive usage of passive safety systems. This paper describes some important criteria and the development program of the Chinese large-scale pressurized water reactors. An overview on representative research activities and results achieved so far on passive safety systems in various institutions is presented.

Preliminary Safety Analysis of a Compact Small Reactor: Feedwater Line Break and Small-Break LOCA Accident Analysis

2017 ◽  
Author(s):  
Linsen Li ◽  
Feng Shen ◽  
Mian Xing ◽  
Zhan Liu ◽  
Zhanfei Qi
Keyword(s):  
Conceptual Design ◽  
Cooling System ◽  
Passive Safety ◽  
Accident Analysis ◽  
Primary System ◽  
Primary Coolant ◽  
Pressurized Water ◽  
Passive Safety Systems ◽  
Core Cooling ◽  
Safety Systems

A small Pressurized Water Reactor (PWR) with compact primary system and passive safety feature, which is called Compact Small Reactor (CSR), is under pre-conceptual design and development. For the purpose of preliminary assessment of the primary coolant system and capability evaluation of the passive safety system, a detailed thermal-hydraulic (T-H) system model of the CSR was developed. Several design-basis accidents, including feedwater line break, double ended direct vessel injection line break (one of the small-break Loss Of Coolant Accidents, LOCA) and etc, are selected and simulated so as to evaluate and further optimize the design of passive safety systems, especially the passive core cooling system. The results of preliminary accident analysis show that the passive safety systems are basically capable of mitigating the accidents and protecting the reactor core from severe damage. Further research will be focused on the optimization of pre-conceptual design of the thermal-hydraulic system and the passive core cooling system.

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.

Emergency Strategy Research of Loss of Coolant Accident for Small Modular Reactor

2017 ◽  
Author(s):  
Wang Yuqi ◽  
Yu Aimin ◽  
Yang Qingming
Keyword(s):  
Simulation System ◽  
Passive Safety ◽  
Strategy Research ◽  
Passive Safety Systems ◽  
Loss Of Coolant Accident ◽  
Small Modular Reactor ◽  
Loss Of Coolant ◽  
Full Power ◽  
Safety Systems ◽  
Emergency Operating

This paper researched the behavior of 20mm break Loss of Coolant Accident (LOCA) which is located in the Direct Vessel Injection (DVI) line of the integrated small modular reactor (SMR) in case of full power with RELAP5-3KEYMASTER simulation system. The response of passive safety systems is analyzed and compared with the Primary Safety Analysis Report (PSAR) post-accident. Tendency for the variation of main parameters after the accident agree well with the PSAR, which validates the accuracy and rationality of the model, and solves the new problems in the process of modeling and provides an important tool for the research and development of SMR. Cooling and depressurization are calculated post-accident. The variation of main parameters post-accident and the accident advancement and results have been analyzed. Operation intervention is given and the effects with it are discussed. And the emergency strategy for development and verification of Emergency Operating Procedures (EOP) is given.

Preliminary Study of Large-Power PWR With Safety Balanced Design

2010 ◽  
Author(s):  
Jue Yang ◽  
Xuenong Zhu ◽  
Xiangang Fu ◽  
Wei Cai ◽  
Jie Ye ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Safety System ◽  
Passive Safety ◽  
Active Safety ◽  
Plant Design ◽  
Large Power ◽  
Passive Safety Systems ◽  
Active Safety Systems ◽  
Environmental Goals ◽  
Safety Systems

Developing the advanced nuclear power plant design to meet the demanding safety, efficiency and environmental goals of electric utilities requires great efforts. In this paper, a design of the safety systems for the large-power PWR units is introduced, which is deemed a optimal combination of the passive safety systems with the active safety systems. The typical design basis accidents are analyzed for this safety system design, such as the Small Break LOCA, SGTR, SLB and Loss of Flow Accidents (LOFA). The results show that the safety systems of the passives combined the actives can mitigate effectively these typical accidents in large-power PWRs. PSA results also show that the passive safety systems contributes to the reduction of the CDF. It is preliminarily concluded that the passive combined active safety system is designed in balance.

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.

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.

Improved Occupant Protection through Cooperation of Active and Passive Safety Systems – Combined Active and Passive Safety CAPS

2006 ◽  
Author(s):  
Alfred Kuttenberger ◽  
Sybille Eisele ◽  
Thomas Lich ◽  
Thorsten Sohnke ◽  
Jorge Sans Sangorrin ◽  
...  
Keyword(s):  
Passive Safety ◽  
Occupant Protection ◽  
Passive Safety Systems ◽  
Safety Systems
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