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.

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.

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.

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.

scholarly journals Onshore Nuclear Power Plant Concept with Enhanced Passive Safety System

2019 ◽  
Vol 4 (6) ◽  
pp. 155-159
Author(s):  
A.H.M. Iftekharul Ferdous ◽  
T. H. M Sumon Rashid ◽  
Md Asaduzzaman Shobug ◽  
Tanveer Ahmed ◽  
Nitol Kumar Dutta
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Water Source ◽  
Passive Safety ◽  
Electric Power Supply ◽  
Passive Safety Systems ◽  
Passive Safety System ◽  
Safety Systems

Bangladesh is a developing country and it’s increasing economy can be maintained by providing sufficient amount of electric power supply. Therefore government is initiating Rooppur nuclear power project is one of them which is needed to be sited beside a vast amount of water source, lowest populated area and away from the locality to reduce the damage caused by any nuclear accidents. In this thesis paper we have shown that, the the dangers of residing errors of Rooppur nuclear power plant and give a proposal to go for onshore nuclear power plant in Bangladesh with two proposed designs of passive safety systems PSS-I & PSS-II. These systems will give safety to the power plants in the case of plant blackout during accidents.

scholarly journals Experiment for Justification the Reliability of Passive Safety System in NPP

2018 ◽  
Vol 3 (3) ◽  
pp. 1
Author(s):  
D.S. Samokhin ◽  
Mohammad Alslman ◽  
A. D. Vostrilova ◽  
O.Yu. Kochnov
Keyword(s):  
Power Systems ◽  
Nuclear Power ◽  
Critical Issue ◽  
Nuclear Industry ◽  
Passive Safety ◽  
Mathematical Methods ◽  
Mean Time Between Failures ◽  
Passive Safety Systems ◽  
Passive Safety System ◽  
Safety Systems

This article gives an overview of the formation of the global nuclear industry, highlighted a critical issue of ensuring safe operation of nuclear power systems in modern projects. Considering the use of passive safety systems in the design of a nuclear power plant, and discussed the different mathematical methods for assessing the reliability of passive systems. Also it considers the possibility of finding the mean time between failures, using these methods to assess the reliability of passive safety systems.

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.

scholarly journals Application of REPAS Methodology to Assess the Reliability of Passive Safety Systems

2009 ◽  
Vol 2009 ◽  
pp. 1-18 ◽  
Author(s):  
Franco Pierro ◽  
Dino Araneo ◽  
Giorgio Galassi ◽  
Francesco D'Auria
Keyword(s):  
Monte Carlo ◽  
Safety Assessment ◽  
Probability Distributions ◽  
Passive Safety ◽  
Passive System ◽  
Passive Systems ◽  
Passive Safety Systems ◽  
Passive Safety System ◽  
Uncertain Input ◽  
Safety Systems

The paper deals with the presentation of the Reliability Evaluation of Passive Safety System (REPAS) methodology developed by University of Pisa. The general objective of the REPAS is to characterize in an analytical way the performance of a passive system in order to increase the confidence toward its operation and to compare the performances of active and passive systems and the performances of different passive systems. The REPAS can be used in the design of the passive safety systems to assess their goodness and to optimize their costs. It may also provide numerical values that can be used in more complex safety assessment studies and it can be seen as a support to Probabilistic Safety Analysis studies. With regard to this, some examples in the application of the methodology are reported in the paper. A best-estimate thermal-hydraulic code, RELAP5, has been used to support the analyses and to model the selected systems. Probability distributions have been assigned to the uncertain input parameters through engineering judgment. Monte Carlo method has been used to propagate uncertainties and Wilks' formula has been taken into account to select sample size. Failure criterions are defined in terms of nonfulfillment of the defined design targets.

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.

SIRIO: An Experimental Facility for a New Heat Removal System Passively Controlled by Non-Condensable Gases

2018 ◽  
Author(s):  
R. Marinari ◽  
M. Tarantino ◽  
F. S. Nitti ◽  
A. Alemberti ◽  
M. Caramello ◽  
...  
Keyword(s):  
Heat Removal ◽  
Passive Safety ◽  
Primary System ◽  
Experimental Facility ◽  
The Past ◽  
Passive Safety Systems ◽  
Design Activities ◽  
Safety Systems ◽  
Heat Removal System ◽  
Removal System

Heat removal systems are of major importance for both present and future nuclear power plants as they belong to the set of systems devoted to ensure the integrity of the reactor core and to avoid core damage. The past experience and lessons learned on this topic suggest to adopt passive safety systems which can perform the safety function independently from operators’ actions and external energy sources, ensuring long term reactor cooling. Application of these systems to innovative reactor concepts such as (heavy) liquid metal reactors poses a problem related to the characteristic properties of the coolant: as the final heat sink of passive safety systems is often the external environment, the liquid metal will eventually undergo a phase change and solidify at the end of a complex dynamic process. The solidification of the coolant may have important effects on the transient behavior if it happens at an early stage of an accident, as the main flow path of the heat exchanger can be blocked by the coolant freezing while the decay heat in the core is still sufficiently high and need to be efficiently removed. An innovative passive safety system has been proposed for the decay heat removal system of ALFRED reactor (DEMO LFR, Gen.IV) where the issue of early coolant freezing is prevented. The innovation has been object of a patent and the system is potentially able to avoid solidification by reducing the amount of heat removed from the primary system by means of non-condensable gases passively injected into the water/steam mixture, which induce heat transfer degradation. Several numerical studies have been performed during the past years, but a complete validation of the operating principle requires an experimental assessment and characterization. To this aim the SIRIO experimental facility, scaled on the DHR of ALFRED, has been conceived. Several design activities have been performed so far for the development of the facility, such as scaling analysis on the basis of ALFRED DHR to determine the facility size, numerical simulations by means of RELAP5-3D to determine whether the facility is able to reproduce the expected physical phenomena and numerical simulations by means of Ansys CFX to investigate the performance of a heating system simulating the primary system of ALFRED based on a molten salt annulus. The present paper describes the design activities performed and provides insights on the methodologies adopted, as well as the current status of the design of the SIRIO facility.

Injury Criteria Based Optimisation of Automobile Passive Safety Systems to Mitigate the Effects of Full-Frontal Automobile Collisions

2014 ◽  
Author(s):  
Volkan Esat
Keyword(s):  
Interior Design ◽  
Simulation Software ◽  
Safety System ◽  
Passive Safety ◽  
Steering Wheel ◽  
System Parameters ◽  
Passive Safety Systems ◽  
Injury Criteria ◽  
Passive Safety System ◽  
Safety Systems

Passive safety systems such as airbags, seat belts, and interior structural design of the automobile play a significant role in injury prevention of the occupant during collisions. Important design and operation parameters of the passive safety systems such as airbag firing times and steering wheel position as an interior design attribute potentially affect the dynamics of the occupant during impact and determine the amount of mitigation of a possible injury. This research aims to contribute towards improving passive safety systems in automobile design for mitigation of injuries by optimising the features and parameters of various subsystems such as driver’s airbag and steering wheel. Two separate computational models, a 5th percentile female and a 50th percentile male, comprising of a typical automobile interior with passive safety systems are constructed in the specialised impact simulation software MADYMO. Two different crash pulses of 30 kph and 48 kph are applied to the computational human body models in full-frontal crashes. Passive safety system parameters; in particular, airbag firing times and steering wheel column angles, are varied to investigate their effects on the head, neck and upper torso through injury criteria. Injury criteria predictions are employed in optimisation algorithms to figure out the best combinations for passive safety system parameters in order to mitigate possible injuries for all drivers.

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