scholarly journals PRELIMINARY ASSESSMENT OF ENGINEERED SAFETY FEATURES AGAINST STATION BLACKOUT IN SELECTED PWR MODELS

2021 ◽  
Vol 23 (2) ◽  
pp. 47
Author(s):  
Andi Sofrany Ekariansyah ◽  
Surip Widodo ◽  
Susyadi Susyadi ◽  
Hendro Tjahjono
Keyword(s):  
Heat Exchanger ◽  
Heat Removal ◽  
Air Cooling ◽  
Water Tank ◽  
Forced Circulation ◽  
Design Assessment ◽  
Decay Heat ◽  
Time Period ◽  
Station Blackout ◽  
Safety Features

The 2011 Fukushima accident did not prevent countries to construct new nuclear power plants (NPPs) as part of the electricity generation system. Based on the IAEA database, there are a total of 44 units of PWR type NPPs whose constructions are started after 2011. To assess the technology of engineered safety features (ESFs) of the newly constructed PWRs, a study has been conducted as described in this paper, especially in facing the station blackout (SBO) event. It is expected from this study that there are a number of PWR models that can be considered to be constructed in Indonesia from the year of 2020. The scope of the study is PWRs with a limited capacity from 900 to 1100 MWe constructed and operated after 2011 and small-modular type of reactors (SMRs) with the status of at least under licensing. Based on the ESFs design assessment, the passive core decay heat removal has been applied in the most PWR models, which is typically using steam condensing inside heat exchanger within a water tank or by air cooling. From the selected PWR models, the CPR-1000, HPR-1000, AP-1000, and VVER-1000, 1200, 1300 series have the capability to remove the core decay heat passively. The most innovative passive RHR of AP-1000 and the longest passive RHR time period using air cooling in several VVER models are preferred. From the selected SMR designs, the NuScale design and RITM-200 possess more advantages compared to the ACP-100, CAREM-25, and SMART. NuScale represents the model with full-power natural circulation and RITM-200 with forced circulation. NuScale has the longest time period for passive RHR as claimed by the vendor, however the design is still under licensing process. The RITM-200 reactor has a combination of passive air and water-cooling of the heat exchanger and is already under construction.  

AP1000® Passive Core Cooling System Performance Following an Extended Station Blackout Event

2014 ◽  
Author(s):  
Richard F. Wright ◽  
Stephen Swantner ◽  
Matthew M. Swartz ◽  
John Lojek ◽  
Yong Jae Song ◽  
...  
Keyword(s):  
Heat Exchanger ◽  
Cooling System ◽  
Heat Removal ◽  
Inside Surface ◽  
Decay Heat ◽  
Containment Vessel ◽  
Station Blackout ◽  
Fukushima Daiichi ◽  
Operator Action ◽  
Core Cooling

As an advanced Gen III+ plant with passive safety systems, the AP1000® plant is uniquely equipped to handle an extended station blackout (SBO) event similar to what occurred at the Fukushima-Daiichi plants in March of 2011. These passive systems have been designed to maintain core cooling for up to 72 hours following all design basis events without the need for AC power or operator action. These core and containment cooling systems self-actuate such that even DC power is not required for their actuation. The Fukushima-Daiichi event demonstrated the effectiveness and desirability of the AP1000 systems. The AP1000 plant, like other pressurized water reactors (PWRs), is provided with defense-in-depth active systems, such as auxiliary feed water pumps, to remove decay heat using the steam generators in the event that offsite power is lost. During an SBO the diesel generators powering this active equipment would not be available. In the event of an SBO the safety-grade heat removal function would be accomplished by the passive residual heat removal (PRHR) heat exchanger (HX) located in the in-containment refueling water storage tank (IRWST). The PRHR HX is designed to remove decay heat from the reactor coolant system (RCS) to the water in the IRWST, which increases in temperature and eventually boils. Steam from the IRWST is vented to the containment atmosphere and actuates the passive containment cooling system (PCS), which is used to apply water to the outside of the steel containment vessel and passively remove heat via evaporation to the environment. Steam that is condensed on the inside surface of the containment vessel forms a water film that flows down the containment wall and is returned to the IRWST using a system of water collection gutters and piping. The PCS is sized to remove reactor decay heat for 72 hours without the need for operator action. Effective operation of the PRHR heat exchanger and PCS to remove decay heat from the reactor core to the environment depends on the ability to maintain water in the IRWST. Condensate that is not collected and returned to the IRWST is lost into the containment sump. There are several possible sources of loss. At the start of IRWST boiling, all containment structures will condense steam until their surface temperature approaches the steam temperature. This process is dependent on the heat capacity of these structures, and all condensation formed on these structures is considered lost. Since the containment wall is cooled by the PCS operation, condensation continues on the inside surface of the containment throughout the event. There are areas on the containment wall where condensate could be lost including the region at the top of the dome where the surface is nearly horizontal, and areas where weld seams and other obstructions could strip off some condensate film. To determine the coping time limits following an extended SBO, it is necessary to characterize these condensate losses. A Phenomena Identification and Ranking Table (PIRT) process was conducted to determine the important phenomena associated with the return of condensate to the IRWST. This PIRT process identified the need for further experimentation to quantify the losses. This paper describes the PIRT and the experimental facility design used to determine the condensate return losses arising from phenomena identified by the PIRT.

scholarly journals Effect of Coolant Inventories and Parallel Loop Interconnections on the Natural Circulation in Various Heat Transport Systems of a Nuclear Power Plant during Station Blackout

2008 ◽  
Vol 2008 ◽  
pp. 1-11 ◽  
Author(s):  
Avinash J. Gaikwad ◽  
P. K. Vijayan ◽  
Sharad Bhartya ◽  
Kannan Iyer ◽  
Rajesh Kumar ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Heat Transport ◽  
Nuclear Power ◽  
Natural Circulation ◽  
Heat Removal ◽  
Transport Systems ◽  
Critical Examination ◽  
Decay Heat ◽  
Station Blackout

Provision of passive means to reactor core decay heat removal enhances the nuclear power plant (NPP) safety and availability. In the earlier Indian pressurised heavy water reactors (IPHWRs), like the 220 MWe and the 540 MWe, crash cooldown from the steam generators (SGs) is resorted to mitigate consequences of station blackout (SBO). In the 700 MWe PHWR currently being designed an additional passive decay heat removal (PDHR) system is also incorporated to condense the steam generated in the boilers during a SBO. The sustainability of natural circulation in the various heat transport systems (i.e., primary heat transport (PHT), SGs, and PDHRs) under station blackout depends on the corresponding system's coolant inventories and the coolant circuit configurations (i.e., parallel paths and interconnections). On the primary side, the interconnection between the two primary loops plays an important role to sustain the natural circulation heat removal. On the secondary side, the steam lines interconnections and the initial inventory in the SGs prior to cooldown, that is, hooking up of the PDHRs are very important. This paper attempts to open up discussions on the concept and the core issues associated with passive systems which can provide continued heat sink during such accident scenarios. The discussions would include the criteria for design, and performance of such concepts already implemented and proposes schemes to be implemented in the proposed 700 MWe IPHWR. The designer feedbacks generated, and critical examination of performance analysis results for the added passive system to the existing generation II & III reactors will help ascertaining that these safety systems/inventories in fact perform in sustaining decay heat removal and augmenting safety.

scholarly journals Heat Transfer Analysis of Passive Residual Heat Removal Heat Exchanger under Natural Convection Condition in Tank

2014 ◽  
Vol 2014 ◽  
pp. 1-8 ◽  
Author(s):  
Qiming Men ◽  
Xuesheng Wang ◽  
Xiang Zhou ◽  
Xiangyu Meng
Keyword(s):  
Heat Transfer ◽  
Heat Exchanger ◽  
Thermal Stratification ◽  
Heat Removal ◽  
Heat Transfer Analysis ◽  
Average Error ◽  
Water Tank ◽  
Shaped Tube ◽  
Heat Transfer Calculation ◽  
Residual Heat

Aiming at the heat transfer calculation of the Passive Residual Heat Removal Heat Exchanger (PRHR HX), experiments on the heat transfer of C-shaped tube immerged in a water tank were performed. Comparisons of different correlation in literatures with the experimental data were carried out. It can be concluded that the Dittus-Boelter correlation provides a best-estimate fit with the experimental results. The average error is about 0.35%. For the tube outside, the McAdams correlations for both horizontal and vertical regions are best-estimated. The average errors are about 0.55% for horizontal region and about 3.28% for vertical region. The tank mixing characteristics were also investigated in present work. It can be concluded that the tank fluid rose gradually which leads to a thermal stratification phenomenon.

Analysis of ESFR Decay Heat Removal Systems in Protected Station Blackout

2021 ◽  
Author(s):  
Janos Bodi ◽  
Alexander Ponomarev ◽  
Evaldas Bubelis ◽  
Konstantin Mikityuk
Keyword(s):  
Power Plant ◽  
Fast Reactor ◽  
Heat Removal ◽  
Current Paper ◽  
Decay Heat ◽  
Sodium Fast Reactor ◽  
Station Blackout ◽  
Safety Assessments ◽  
The Individual

Abstract As part of the ESFR-SMART project, safety assessments are being conducted on the updated European Sodium Fast Reactor (ESFR) design. An important part of the study is the evaluation of the reactor's behavior within hypothetical accidental conditions to assess and ensure that the accident would not lead to unexpected and disastrous events. In the current paper, the analyzed accidental scenario is the so called Protected Station Blackout (PSBO), where the offsite power is lost for the power plant, simulated by using the TRACE and SIM-SFR system codes. The assessment started from the simulation of the reactor behavior without the decay heat removal systems (DHRS). Following this, calculations of multiple DHRS arrangements have been performed to evaluate the individual and combined efficiency of the systems. Where it was possible, the results from the two system codes have been compared to show the consistency of the separate calculations. Through this study, the design of the DHRSs proposed at the beginning of the project have been investigated, and certain recommendations have been made for further improvement of the DHRS systems performance.

Validation Experiment for the Cooling Capability of Passive Auxiliary Feedwater System (PAFS) With PASCAL Facility

2012 ◽  
Author(s):  
Byoung-Uhn Bae ◽  
Seok Kim ◽  
Yu-Sun Park ◽  
Bok-Deuk Kim ◽  
Kyoung-Ho Kang ◽  
...  
Keyword(s):  
Heat Transfer ◽  
Natural Convection ◽  
Heat Exchanger ◽  
System Analysis ◽  
Thermal Power ◽  
Heat Removal ◽  
Steady State Condition ◽  
Single Tube ◽  
Decay Heat ◽  
Validation Experiment

The Passive Auxiliary Feedwater System (PAFS) is one of the advanced safety features adopted in the APR+ (Advanced Power Reactor Plus) which is intended to completely replace the conventional active auxiliary feedwater system. It removes the decay heat by cooling down the secondary system of the SG using condensation heat exchanger installed in the Passive Condensation Cooling Tank (PCCT). With an aim of validating the cooling and operational performance of the PAFS, PASCAL (PAFS Condensing Heat Removal Assessment Loop), was constructed to experimentally investigate the condensation heat transfer and natural convection phenomena in the PAFS. It simulates a single tube of the passive condensation heat exchangers, a steam-supply line, a return-water line, and a PCCT with a reduced area, which is equivalent to 1/240 of the prototype according to a volumetric scaling methodology with a full height. The objective of the experiment is to investigate the cooling performance and natural circulation characteristics of the PAFS by simulating a steady state condition of the thermal power. From the experiment, two-phase flow phenomena in the horizontal heat exchanger and PCCT were investigated and the cooling capability of the condensation heat exchanger was validated. Test results showed that the design of the condensation heat exchanger in PAFS could satisfy the requirement for heat removal rate of 540 kW per a single tube and the prevention of water hammer phenomenon inside the tube. It also proved that the operation of PAFS played an important role in cooling down the decay heat by natural convection without any active system. The present experimental results will contribute to improve the model of the condensation and boiling heat transfer, and also to provide the benchmark data for validating the calculation performance of a thermal hydraulic system analysis code with respect to the PAFS.

Analysis of Protected Accidental Transients in the EFIT Reactor With the RELAP5 Thermal-Hydraulic Code

2008 ◽  
Author(s):  
Giacomino Bandini ◽  
Maddalena Casamirra ◽  
Francesco Castiglia ◽  
Mariarosa Giardina ◽  
Paride Meloni ◽  
...  
Keyword(s):  
Proton Beam ◽  
Thermal Power ◽  
Heat Removal ◽  
Reactor Power ◽  
Lead Target ◽  
Primary System ◽  
Safety Behavior ◽  
Forced Circulation ◽  
Detailed Model ◽  
Decay Heat

The European Facility for Industrial Transmutation (EFIT) is aimed at demonstrating the feasibility of transmutation process through the Accelerator Driven System (ADS) route on an industrial scale. The conceptual design of this reactor of about 400 MW thermal power is under development in the frame of the European EUROTRANS Integrated Project of the EURATOM Sixth Framework Program (FP6). EFIT is a pool-type reactor cooled by forced circulation of lead in the primary system where the heat is removed by steam generators installed inside the reactor vessel. The reactor power is sustained by a spallation neutron source supplied by a proton beam impinging on a lead target at the core centre. A safety-related Decay Heat Removal (DHR) system provided with four independent inherently safe loops is installed in the primary vessel to remove the decay heat in case of loss of secondary circuits heat removal capability. A quite detailed model of the EFIT reactor has been developed for the RELAP5 thermal-hydraulic code to be used in preliminary accidental transient analyses aimed at verifying the validity of the adopted solutions for the current reactor design with respect to the safety requirements, and confirm the inherent safety behavior of the reactor, such as decay heat removal in accidental conditions relying on natural circulation in the primary system. The accident analyses for the EFIT reactor include both protected and unprotected transients, on whether the reactor automatic trip, consisting in proton beam switch off, is actuated or not by the protection system. In this paper, the main results of the analyses of some protected transients with RELAP5 are presented. The analyzed transients concern the Protected Loss of Heat Sink (PLOHS), in which the DHR system plays a key role in bringing the reactor in safe conditions, and the Protected Loss of Flow (PLOF) transients with partial or total loss of forced circulation in the primary system.

Performance Evaluation of DRACS System for FHTR and Time Assessment of Operation Procedure

2013 ◽  
Author(s):  
Junya Nakata ◽  
Mikihiro Wakui ◽  
Michitsugu Mori ◽  
Hiroto Sakashita ◽  
Charles Forsberg
Keyword(s):  
Performance Evaluation ◽  
High Temperature ◽  
Cooling System ◽  
Heat Removal ◽  
Severe Accident ◽  
Air Cooling ◽  
Fluoride Salt ◽  
Nuclear Power Reactor ◽  
Decay Heat ◽  
Operation Procedure

The Fluoride-salt-cooled High-temperature Reactor (FHR) is a new concept of nuclear power reactor being investigated mainly in U.S. and China. The coolant is a liquid salt with a melting point of about 460°C and a boiling point of over 1400°C. As the baseline decay heat removal system, a passive Direct Reactor Air Cooling System (DRACS) is utilized. Though DRACS system has been developed in Sodium Fast reactors (SFR), there are some differences between both. For example, the system in FHR must decrease heat removal when temperatures are low to avoid freezing of the salt and blocking the flow of liquid. Therefore, considering its characteristics, a numerical investigation of DRACS system is needed to evaluate whether FHR has proper ability to remove decay heat and to be robust for a long-time cooling operation after even a severe accident. Furthermore, in addition to its performance evaluation, it is required to make up the operation plan of FHR considering features of this system. It is highly important, with the view of avoiding severe accident, to determine by when the system should be started up. In both countries mentioned above, Fluoride-salt-cooled High-temperature Test Reactor (FHTR) is currently in progress to build. Reviewing its design and system is a crucial step needed to develop the FHR technology. In this research, a performance of DRACS system under some thermal-hydraulic basic events was evaluated by numerical simulation. This paper also suggested the adequate operation procedure suitable for FHTR to avoid a severe accident.

Analysis of Heat Transfer and Flow Characteristics of AP1000 Passive Residual Heat Removal Heat Exchanger

2014 ◽  
Author(s):  
Xu Xie ◽  
Changhua Nie ◽  
Li Zhan ◽  
Hua Zheng ◽  
Pengzhou Li ◽  
...  
Keyword(s):  
Heat Transfer ◽  
Heat Exchanger ◽  
Temperature Difference ◽  
Tube Bundle ◽  
Natural Circulation ◽  
Heat Removal ◽  
Large Temperature ◽  
Water Tank ◽  
Secondary Side ◽  
Residual Heat

In this paper, the computational fluid dynamics (CFD) method is applied to the thermal-hydraulic analysis, while the porous media model is used to simplify AP1000 passive residual heat removal heat exchanger tube. The temperature as well as flow distribution in the secondary side of the heat exchanger are obtained, aiming at analysis of natural circulation ability. It can be noted that the fluid in the secondary side of heat exchanger moves driven by the effect of thermal buoyancy, forming the natural cycle, which takes away heat in tube bundle region. The heat transfer in water tank is mainly enhanced by vortex and turbulent flow, caused by the large resistance of tube bundle region as well as large temperature difference. This phenomenon is obvious especially for the recirculation of flow near the tube bundle. The enduring change of flow rate and direction enhance the heat transfer. Besides, the big temperature difference helps to increase the driving effect of natural circulation. Consequently, the heat transfer of the tank is enhanced by above mechanism. The results of this study contribute to the capacity analysis of passive residual heat removal of natural circulation system, providing valuable information for safe operation of AP1000.

AP1000 Passive Residual Heat Removal Heat Exchanger Confirmatory Analysis

2006 ◽  
Author(s):  
Richard F. Wright ◽  
James R. Schwall ◽  
Creed Taylor ◽  
Naeem U. Karim ◽  
Jivan G. Thakkar ◽  
...  
Keyword(s):  
Heat Exchanger ◽  
Cooling System ◽  
Heat Removal ◽  
Passive Safety ◽  
Tube Heat Exchanger ◽  
Removal Capacity ◽  
Safety Margins ◽  
Final Design ◽  
Safety Features ◽  
Residual Heat

The AP1000 is an 1100 MWe advanced nuclear power plant that uses passive safety features to enhance plant safety and to provide significant and measurable improvements in plant simplification, reliability, investment protection and plant costs. The AP1000 received final design approval from the US-NRC in 2004. The AP1000 design is based on the AP600 design that received final design approval in 1999. Wherever possible, the AP1000 plant configuration and layout was kept the same as AP600 to take advantage of the maturity of the design and to minimize new design efforts. As a result, the two-loop configuration was maintained for AP1000, and the containment vessel diameter was kept the same. It was determined that this significant power uprate was well within the capability of the passive safety features, and that the safety margins for AP1000 were greater than those of operating PWRs. A key feature of the passive core cooling system is the passive residual heat removal heat exchanger (PRHR HX) that provides decay heat removal for postulated LOCA and non-LOCA events. The PRHR HX is a C-tube heat exchanger located in the in-containment refueling water storage tank (IRWST) above the core promoting natural circulation heat removal between the reactor cooling system and the tank. Component testing was performed for the AP600 PRHR HX to determine the heat transfer characteristics and to develop correlations to be used for the AP1000 safety analysis codes. The data from these tests were confirmed by subsequent integral tests at three separate facilities including the ROSA facility in Japan. Owing to the importance of this component, an independent analysis has been performed using the ATHOS-based computational fluid dynamics computer code PRHRCFD. Two separate models of the PRHR HX and IRWST have been developed representing the ROSA test geometry and the AP1000 plant geometry. Confirmation of the ROSA test results were used to validate PRHRCFD, and the AP1000 plant model was used to confirm the heat removal capacity for the full-sized heat exchanger. The results of these simulations show that the heat removal capacity of the PRHR HX is conservatively represented in the AP1000 safety analyses.

Setup of the Supercritical CO2 Test Facility “SCARLETT” for Basic Experimental Investigations of a Compact Heat Exchanger for an Innovative Decay Heat Removal System

2018 ◽  
Vol 4 (3) ◽  
Author(s):  
Wolfgang Flaig ◽  
Rainer Mertz ◽  
Joerg Starflinger
Keyword(s):  
Heat Exchanger ◽  
Supercritical Co2 ◽  
Nuclear Power ◽  
Power Plants ◽  
Heat Removal ◽  
Test Facility ◽  
Experimental Investigations ◽  
Decay Heat ◽  
Heat Removal System ◽  
Removal System

Supercritical fluids show great potential as future coolants for nuclear reactors, thermal power, and solar power plants. Compared to the subcritical condition, supercritical fluids show advantages in heat transfer due to thermodynamic properties near the critical point. A specific field of interest is an innovative decay heat removal system for nuclear power plants, which is based on a turbine-compressor system with supercritical CO2 as the working fluid. In case of a severe accident, this system converts the decay heat into excess electricity and low-temperature waste heat, which can be emitted to the ambient air. To guarantee the retrofitting of this decay heat removal system into existing nuclear power plants, the heat exchanger (HE) needs to be as compact and efficient as possible. Therefore, a diffusion-bonded plate heat exchanger (DBHE) with mini channels was developed and manufactured. This DBHE was tested to gain data of the transferable heat power and the pressure loss. A multipurpose facility has been built at Institut für Kernenergetik und Energiesysteme (IKE) for various experimental investigations on supercritical CO2, which is in operation now. It consists of a closed loop where the CO2 is compressed to supercritical state and delivered to a test section in which the experiments are run. The test facility is designed to carry out experimental investigations with CO2 mass flows up to 0.111 kg/s, pressures up to 12 MPa, and temperatures up to 150 °C. This paper describes the development and setup of the facility as well as the first experimental investigation.

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