Safety analysis of increase in heat removal from reactor coolant system with inadvertent operation of passive residual heat removal at no-load conditions

The inherent system safety of the 100 MW integral pressurized water reactor (IPWR) can be improved by placing the core, the efficient once-through steam generators and the main coolant pumps in the reactor pressure vessel, and omitting some large pipes and valves in the primary coolant system which can prevent the occurrence of large break loss of coolant accident and reduce the possibility of core melt accident. The application of the passive safety systems simplifies the structures of IPWR and improves the economy of the reactor. In case of accidents, the primary coolant system establishes natural circulation to take the core decay heat away by passive safety systems using gravity and other natural driving forces, thereby enhancing the safety and reliability of the system IPWR. It’s of great significance to establish reasonable and correctable models, including the primary coolant system model, the second loop model and passive core cooling system model, to study thermal-hydraulic phenomena under steady state, transient state and accident conditions. Based on transient safety analysis program RELAP5/MOD3.4, 100 MW IPWR system was simulated. A series of models of reactor coolant system and passive safety systems were established. The main system models are composed of primary coolant system model, part of second loop model, passive safety injection system model and passive residual heat removal system model. The primary coolant system model includes core, lower plenum, downcomer, region of steam generators, upper plenum, riser, pressurizer, and surge line; the second loop model includes the main feed water line, the steam line, and steam generator tubes; passive safety injection system model includes core makeup tank, accumulator, automatic depressurization system, direct vessel injection line; and passive residual heat removal system model includes passive residual heat removal heat exchanger in containment refueling water storage tank. Based on the established models, the steady state was debugged with the RELAP5 input card. Steady state calculation was performed and the results agree well with designed values, which verifies the validity of the model and the input card. Using the steady state results as initial conditions, transient calculation was performed. Typical accidents (loss of main water accident) were calculated, which were relieved by auxiliary feedwater system (AFWS) and passive residual heat removal system (PRHR SYSTEM). The results, obtained from AFWS and PRHR SYSTEM, were contrasted and process of accident and thermal-hydraulic phenomena were analyzed according to transient calculation results. The transient calculation results showed that the integral PWR system and the passive safety system model can provide a reference for IPWR transient safety analysis.

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Towards the Development of Shutdown Emergency Operating Procedures for Power Nuclear Reactors

Volume 3: Nuclear Safety and Security; Codes, Standards, Licensing and Regulatory Issues; Computational Fluid Dynamics and Coupled Codes ◽

10.1115/icone21-16788 ◽

2013 ◽

Author(s):

Isaac Malgas

Keyword(s):

Heat Removal ◽

Optimal Recovery ◽

Reactor Coolant ◽

Design Basis ◽

Power Operation ◽

Restoration Strategies ◽

Procedure Guidance ◽

Heat Removal System ◽

Coolant System ◽

Emergency Operating

Virtually all operating nuclear plants have a well-developed suite of accident procedures which can be used in the event of an accident while the unit is at power. The procedures guide the operators through optimal recovery strategies to place the plant in a safe stable shutdown state. These procedures cater for design basis accidents as well as certain accidents which are not considered design basis accidents, but may be included in the plant’s licensing basis. Generally procedures used for accidents during power operation can also be used for initiating accidents which occur while the unit is shutdown while the core is cooled with steam generators. When the unit is in lower shutdown states and the residual heat removal system is in service, however, the procedure scope becomes somewhat constrained. With the reactor coolant system in an open or closed configuration, the state/configuration specific procedure guidance provided by Emergency Operating Procedures for power operation is often deficient. There is ample literature on developing emergency operating procedures for plant states where the energy (as a function of pressure and temperature) in reactor coolant system is high. However, much fewer sources of literature exist for shutdown plant states when the energy in the reactor coolant system is lower. This paper attempts to bridge this gap by proposing a framework for the development of shutdown emergency operating procedures using the plant specific Probabilistic Safety Assessment (PSA) model. The paper provides an initial framework for the development of Shutdown Emergency Operating Procedures (SD-EOPs) using the PSA. It describes the process of identifying event sequences for which restoration strategies are required from the PSA and briefly explains the analysis that should be performed to demonstrate that restoration strategies are successful in mitigating event consequences.

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Natural Safety Analysis of the Spent Fuel Residual Heat Removal in Loading and Storage Process of HTR-10

Energy Procedia ◽

10.1016/j.egypro.2013.07.209 ◽

2013 ◽

Vol 39 ◽

pp. 227-239 ◽

Cited By ~ 2

Author(s):

Jin-hua Wang ◽

Yi-fan Huang ◽

Yong Tang ◽

Bin Wu

Keyword(s):

Heat Removal ◽

Safety Analysis ◽

Spent Fuel ◽

Storage Process ◽

And Storage ◽

Residual Heat

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Risk Analysis of the Residual Heat Removal System at South Texas Project With a Special Focus on Breach of Containment

Volume 2: Plant Systems, Structures, and Components; Safety and Security; Next Generation Systems; Heat Exchangers and Cooling Systems ◽

10.1115/icone20-power2012-54200 ◽

2012 ◽

Author(s):

Matt Solom ◽

Christopher Chance ◽

Christopher Pannier ◽

Robert Seager ◽

Alan Lee ◽

...

Keyword(s):

Risk Analysis ◽

Failure Modes ◽

Heat Removal ◽

South Texas ◽

Special Focus ◽

Event Tree ◽

Reactor Coolant ◽

Heat Removal System ◽

Removal System ◽

Residual Heat

A unique feature in the design of the reactors at South Texas Project (STP) is that each unit’s Residual Heat Removal System (RHRS) is located within containment. The aim of this work is to identify the potential failure modes of the Residual Heat Removal System that could lead to a breach of containment during reactor operation and thereby may increase Core Damage Frequency (CDF). The analysis began with a Failure Modes and Effects Analysis (FMEA) of the RHRS based on a Piping and Instrumentation Diagram. The two motor operated valves that isolate the RHRS from the Reactor Coolant System (RCS) were assumed to fail with an internal leak, exposing downstream components to reactor coolant. Pathways for coolant to exit containment were identified and analyzed for severity, occurrence, and detectability of the failure modes. The analyses of these factors lead to the determination of a criticality rating, which assisted in the ultimate findings. The results of the FMEA were used to construct an event tree of the failure modes of interest and the composite probability of each failure. The highest probability failure mode of interest was a breach of containment by a tube of the heat exchanger leaking into the Component Cooling Water (CCW) with a failure probability of 2.5E−10 per reactor year. The insights gained in this analysis will be used by the South Texas Project for future risk analysis and decision-making regarding the RHRS.

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The ACC Configuration Sensitivity Study for AP1000 Large Break LOCA

Volume 4: Thermal Hydraulics ◽

10.1115/icone21-15931 ◽

2013 ◽

Author(s):

Jianhui Yu

Keyword(s):

Nuclear Power ◽

Initial Conditions ◽

Heat Removal ◽

Initial Pressure ◽

Sensitivity Study ◽

Water Volume ◽

Initial Water ◽

The Core ◽

Coolant System ◽

Residual Heat

Similar to the traditional nuclear power plant (NPP), the Accumulator (ACC) of AP1000 is one of the most important facility against Large-Break LOCA (LBLOCA). Following a LBLOCA, the Reactor Coolant System (RCS) pressure will be decreased rapidly. And the Core Makeup Tank (CMT) and Passive Residual Heat Removal (PRHR) will be actuated following “S” signal. However, the transient is so rapid that the CMT and PRHR could not be actuated timely, because the ACC will inject water into the reactor vessel downcomer through Direct Vessel Injection (DVI) line and it will stop the CMT injection immediately when RCS has depressurized to the ACC pressure. Therefore, the ACC configuration is very important to LBLOCA mitigation for AP1000. And the Peak Cladding Temperature (PCT) highly relies on ACC configuration. Several sets of different configuration of both ACC, including initial pressure and initial water volume, are discussed. Different initial conditions (e.g. ACC initial pressure) are considered in the sensitivity study on ACC depressurization phase by phase. WCOBRA/TRAC code was used to perform the LBLOCA sensitivity study. The results of each sensitivity case are presented and analyzed. And the suggestion of how to make the optimal ACC configuration is provided in this paper.

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Analysis of the anticipated transient without scram (ATWS) initiated by emergency power mode through the full scope simulator

Kerntechnik ◽

10.1515/kern-2021-1022 ◽

2022 ◽

Vol 0 (0) ◽

Author(s):

Alexandre de Souza Soares ◽

Antonio C. M. Alvim

Keyword(s):

Reactor Core ◽

Heat Removal ◽

Reactor Power ◽

Operating Personnel ◽

Reactor Coolant ◽

Operational Variables ◽

Emergency Power ◽

Power Mode ◽

Reactor Shutdown ◽

Coolant System

Abstract The integrity of the reactor coolant system is severely challenged as a result of an Emergency Power Mode – ATWS event. The purpose of this paper is to simulate the Anticipated Transient without Scram (ATWS) using the full scope simulator of Angra 2 Nuclear Power Plant with the Emergency Power Case as a precursor event. The results are discussed and will be used to examine the integrity of the reactor coolant system. In addition, the results were compared with the data presented in Final Safety Analysis Report (FSAR – Angra 2) in order to guarantee the validation of the methodology and from there analyze other precursor events of ATWS which presented only plausibility studies in FSAR – Angra 2. In this way, the aim is to provide and develop the knowledge and skill necessaries for control room operating personnel to ensure safe and reliable plant operation and stimulate information in the nuclear area through the academic training of new engineers. In the presented paper the most severe scenario is analyzed in which the Reactor Coolant System reaches its highest level of coolant pressure. This scenario is initiated by the turbine trip jointly with the loss of electric power systems (Emergency Power Mode). In addition, the failure of the reactor shutdown system occurs, i.e., control rods fail to drop into the reactor core. The reactor power is safely reduced through the inherent reactivity feedback of the moderator and fuel, together with an automatic boron injection. Several operational variables were analyzed and their profiles over time are shown in order to provide data and benchmarking references. At the end of the event, it was noted that Reactor shutdown is assured, as is the maintenance of subcriticality. Residual heat removal is ensured.

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