Development of Best-Estimate ECCS Evaluation Methodology for APR1400

2009 ◽  
Author(s):  
Seok-Ho Lee ◽  
Mun-Soo Kim ◽  
Han-Gon Kim
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Cooling System ◽  
Evaluation Methodology ◽  
Preliminary Calculation ◽  
Pressurized Water ◽  
Water Storage Tank ◽  
Best Estimate ◽  
Design Characteristics ◽  
Core Cooling

Advanced Power Reactor 1400 (APR1400) is an evolutionary Pressurized Water Reactor (PWR) equipped with such advanced features as the Direct Vessel Injection (DVI), the Fluidic Device (FD) in the Safety Injection Tank (SIT), and the In-containment Refueling Water Storage Tank (IRWST) in the Emergency Core Cooling System (ECCS). To verify the performance of these advanced features, more realistic performance evaluation methodology is desired since existing methodologies use too conservative assumptions which cause negative biases to these features. In this study, therefore, a best estimate evaluation methodology for the APR1400 ECCS under large break loss of cooling accident (LBLOCA) is developed targeting operating license of the Shin Kori 3&4 nuclear power plants (SKN 3&4), the first commercial APR1400 plants. On this purpose, a variety of existing best estimate evaluation methodologies previously used are reviewed. As a result of this review, a methodology named KREM is selected for this study. The KREM is based on RELAP5/MOD3.1K and has been used for Korean operating plants since 2002 when it was first approved by Korean regulation. For this study, RELAP5/MOD3.3 (Patch 3), the latest version of RELAP series is selected since it could appropriately simulate the multi-dimensional phenomena for the APR1400 design characteristics. To quantify the code accuracy, analyses covering experimental data have been performed for 36 kinds of separated effect tests (SETs) and integral effect tests (IETs). The uncertainty in the peak cladding temperature (PCT) of the APR1400 is evaluated preliminarily. Based on the preliminary calculation, final uncertainty quantification and bias evaluation are performed to obtain the licensing PCT for Shin-Kori 3&4 plants and the result shows that the LBLOCA licensing acceptance criteria are well satisfied.

Emergency Core Cooling System Sump Strainer Fibre Bypass Testing and Measurement

2016 ◽  
Author(s):  
Qingwu Cheng ◽  
Harry Adams ◽  
Metin Yetisir
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Cooling System ◽  
Reactor Core ◽  
Regulatory Requirements ◽  
Pressurized Water ◽  
Load Pressure ◽  
Analysis Techniques ◽  
Core Cooling

The potential of losing post-Loss Of Coolant Accident (LOCA) recirculation capability due to debris blockage of Emergency Core Cooling (ECC) strainers resulted in early replacements of ECC strainers in most nuclear power plants. To validate the performance of ECC strainers, extensive testing representing plant conditions is required. Such testing programs include thin-bed and full debris load pressure drop tests, fibre bypass tests and chemical effects tests. Multiple testing loops and state-of-the-art analysis techniques have provided in-depth understanding of sump strainer performance and the effect of chemical precipitation on debris bed head loss. ECC strainers typically use perforated plates as filtering surfaces with 1.6 to 2.5 mm holes and 35 to 40% open area, allowing some particulates and fibres to pass through the strainer filtering surfaces. Recently, the bypassed fibrous debris has been identified as a potential safety concern due to its possible deposition in the reactor core and blocking of flow into fuel assemblies. In some cases, the amount of fibre that is specified as allowed to enter a reactor core is only 15 g per fuel assembly for pressurized water reactors. Characterization and quantification of bypassed fibre debris for nuclear power plants are needed to address regulatory requirements. Testing methodology and analysis techniques to address regulatory requirements and concerns are presented in this paper. In particular, a newly developed technique is presented to address debris bypass quantification.

Modeling of Thermal Hydraulics Features of Top Water Reflood Experiment PARAMETER-SF3 Using SOCRAT/V2 Code

2009 ◽  
Author(s):  
Arcadii E. Kisselev ◽  
Valerii F. Strizhov ◽  
Alexander D. Vasiliev ◽  
Vladimir I. Nalivayev ◽  
Nikolay Ya. Parshin
Keyword(s):  
Experimental Data ◽  
Nuclear Power ◽  
Cooling System ◽  
Severe Accident ◽  
Numerical Code ◽  
Thermal Hydraulics ◽  
Vver Fuel ◽  
Design Basis ◽  
Best Estimate ◽  
Core Cooling

The PARAMETER-SF3 test conditions simulated a severe LOCA (Loss of Coolant Accident) nuclear power plant sequence in which the overheated up to 1700÷2300K core would be reflooded from the top and the bottom in occasion of ECCS (Emergency Core Cooling System) recovery. The test was successfully conducted at the NPO “LUTCH”, Podolsk, Russia, in October 31, 2008, and was the third of four experiments of series PARAMETER-SF. PARAMETER facility of NPO “LUTCH”, Podolsk, is designed for studies of the VVER fuel assemblies behavior under conditions simulating design basis, beyond design basis and severe accidents. The test bundle was made up of 19 fuel rod simulators with a length of approximately 3.12 m (heated rod simulators) and 2.92 m (unheated rod simulator). Heating was carried out electrically using 4-mm-diameter tantalum heating elements installed in the center of the rods and surrounded by annular UO2 pellets. The rod cladding was identical to that used in VVERs: Zr1%Nb, 9.13 mm outside diameter, 0.7 mm wall thickness. After the maximum cladding temperature of about 1900K was reached in the bundle during PARAMETER-SF3 test, the top flooding was initiated. The thermal hydraulic and SFD (Severe Fuel Damage) best estimate numerical complex SOCRAT/V2 was used for the calculation of PARAMETER-SF3 experiment. The counter-current flow limitation (CCFL) model was implemented to best estimate numerical code SOCRAT/V2 developed for modeling thermal hydraulics and severe accident phenomena in a reactor. Thermal hydraulics in PARAMETER-SF3 experiment played very important role and its adequate modeling is important for the thermal analysis. The results obtained by the complex SOCRAT/V2 were compared with experimental data concerning different aspects of thermal hydraulics behavior including the CCFL phenomenon during the reflood. The temperature experimental data were found to be in a good agreement with calculated results. It is indicative of the adequacy of modeling the complicated thermo-hydraulic behavior in the PARAMETER-SF3 test.

scholarly journals ANALYTICAL SIMULATION OF A PWR USING MELCOR

2021 ◽  
Vol 8 (3A) ◽  
Author(s):  
Maritza Rodríguez Gual ◽  
Nathalia N. Araújo ◽  
Marcos C. Maturana
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Cooling System ◽  
Nuclear Regulatory Commission ◽  
Severe Accident ◽  
Normal Operation ◽  
Detailed Knowledge ◽  
Pressurized Water ◽  
Severe Accidents ◽  
Reference Plant

After the two most significant nuclear accidents in history – the Chernobyl Reactor Four explosion in Ukraine(1986) and the Fukushima Daiichi accident in Japan (2011) –, the Final Safety Analysis Report (FSAR) included a new chapter (19) dedicated to the Probabilistic Safety Assessment (PSA) and Severe Accident Analysis (SAA), covering accidents with core melting. FSAR is the most important document for licensing of siting, construction, commissioning and operation of a nuclear power plant. In the USA, the elaboration of the FSAR chapter 19 is according to the review and acceptance criteria described in the NUREG-0800 and U.S. Nuclear Regulatory Commission (NRC) Regulatory Guide (RG) 1.200. The same approach is being adopted in Brazil by National Nuclear Energy Commission (CNEN). Therefore, the FSAR elaboration requires a detailed knowledge of severe accident phenomena and an analysis of the design vulnerabilities to the severe accidents, as provided in a PSA – e.g., the identification of the initiating events involving significant Core Damage Frequency (CDF) are made in the PSA Level 1. As part of the design and certification activities of a plant of reference, the Laboratory of Risk Analysis, Evaluating and Management (LabRisco), located in the University of São Paulo (USP), Brazil, has been preparing a group of specialists to model the progression of severe accidents in Pressurized Water Reactors (PWR), to support the CNEN regulatory expectation – since Brazilian Nuclear Power Plants (NPP), i.e., Angra 1, 2 and 3, have PWR type, the efforts of the CNEN are concentrated on accidents at this type of reactor. The initial investigation objectives were on completing the detailed input data for a PWR cooling system model using the U.S. NRC MELCOR 2.2 code, and on the study of the reference plant equipment behavior – by comparing this model results and the reference plant normal operation main parameters, as modeled with RELAP5/MOD2 code.

Application of Thermal Hydraulic Code SOCRAT/V2 to Top Water Reflood Experiment PARAMETER-SF3

2009 ◽  
Author(s):  
Alexander D. Vasiliev
Keyword(s):  
Experimental Data ◽  
Nuclear Power ◽  
Cooling System ◽  
Severe Accident ◽  
Numerical Code ◽  
Thermal Hydraulics ◽  
Vver Fuel ◽  
Design Basis ◽  
Best Estimate ◽  
Core Cooling

The PARAMETER-SF3 test conditions simulated a severe LOCA (Loss of Coolant Accident) nuclear power plant sequence in which the overheated up to 1700–2300K core would be reflooded from the top and the bottom in occasion of ECCS (Emergency Core Cooling System) recovery. The test was successfully conducted at the NPO “LUTCH”, Podolsk, Russia, in October 31, 2008, and was the third of four experiments of series PARAMETER-SF. PARAMETER facility of NPO “LUTCH”, Podolsk, is designed for studies of the VVER fuel assemblies behavior under conditions simulating design basis, beyond design basis and severe accidents. The test bundle was made up of 19 fuel rod simulators with a length of approximately 3.12 m (heated rod simulators) and 2.92 m (unheated rod simulator). Heating was carried out electrically using 4-mm-diameter tantalum heating elements installed in the center of the rods and surrounded by annular UO2 pellets. The rod cladding was identical to that used in VVERs: Zr1%Nb, 9.13 mm outside diameter, 0.7 mm wall thickness. After the maximum cladding temperature of about 1900K was reached in the bundle during PARAMETER-SF3 test, the top flooding was initiated. The thermal hydraulic and SFD (Severe Fuel Damage) best estimate numerical complex SOCRAT/V2 was used for the calculation of PARAMETER-SF3 experiment. The counter-current flow limitation (CCFL) model was implemented to best estimate numerical code SOCRAT/V2 developed for modeling thermal hydraulics and severe accident phenomena in a reactor. Thermal hydraulics in PARAMETER-SF3 experiment played very important role and its adequate modeling is important for the thermal analysis. The results obtained by the complex SOCRAT/V2 were compared with experimental data concerning different aspects of thermal hydraulics behavior including the CCFL phenomenon during the reflood. The temperature experimental data were found to be in a good agreement with calculated results. It is indicative of the adequacy of modeling the complicated thermo-hydraulic behavior in the PARAMETER-SF3 test.

scholarly journals Early quantification of reliability for a safety critical and control system: a case study of reactor core cooling system

2018 ◽  
Vol 7 (2.12) ◽  
pp. 248
Author(s):  
Vinay Kumar ◽  
Suraj Gupta ◽  
Anil Kumar Tripathi
Keyword(s):  
Control System ◽  
Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Cooling System ◽  
Reactor Core ◽  
Safety Critical ◽  
Core Cooling ◽  
And Control

Using Probabilistic Reliability analysis for Quantifying reliability of a system is already a common practice in Reliability Engineering community. This method plays an important role in analyzing reliability of nuclear plants and its various components. In Nuclear Power Plants Reactor Core Cooling System is a component of prime importance as its breakdown can disrupt Cooling System of power plant. In this paper, we present a framework for early quantification of Reliability and illustrated with a Safety Critical and Control System as case study which runs in a Nuclear Power Plant.  

Analysis of Liquid Sloshing in an In-Containment Refueling Water Storage Tank for AP1000

2013 ◽  
Author(s):  
Yu Liu ◽  
Daogang Lu ◽  
Junjie Dang ◽  
Licun Wu ◽  
Wenhui Ma
Keyword(s):  
Storage Tank ◽  
Nuclear Power ◽  
Cooling System ◽  
Water Storage ◽  
Sine Wave ◽  
Liquid Sloshing ◽  
Normal Operation ◽  
Water Storage Tank ◽  
Core Cooling ◽  
Residual Heat

Although much work has been performed on the liquid sloshing inside simple structures like rectangular and cylindrical vessels, this paper deals with the analysis of the liquid sloshing in a more complex structures, the in-containment refueling water storage tank (IRWST). The IRWST is an important component of AP1000 passive core cooling system to ensure the safe operation of the AP1000 nuclear power plant. In postulated non-LOCA events, the water in the IRWST absorbs the residual heat then transfers the heat into the containment atmosphere. However, in the case of earthquake, the sloshing fluid may influence the safety of the appropriate semi-cylindrical IRWST. In this paper, the liquid transient response in the IRWST was formulated based on finite element modal analysis when the three resonance sine wave was applied as excitation. The result shows that the maximum wave excited by excitation from different directions always emerges from the corner or the edge of tank. Another finding is that water will not overflow at the normal operational water level when exited by the selected excitations in any directions. The safety of the IRWST’s roof is achieved which guarantees that the water vapor and radioactive gases within the tank during normal operation will not release to atmosphere in the containment. The influence of the excitation direction and the water depth are also analyzed.

Use of LBB Methodology to Support the Transition Break Size Concept

2011 ◽  
Author(s):  
James E. Nestell ◽  
David W. Rackiewicz
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Cooling System ◽  
Nuclear Regulatory Commission ◽  
Reactor Coolant ◽  
Design Basis ◽  
Piping Systems ◽  
Pipe Break ◽  
Core Cooling ◽  
Pipe Size

The design basis for a loss of coolant accident in nuclear power plants has previously been based on the assumption that the largest size coolant pipe instantaneously undergoes a double ended “guillotine” break (DEGB) and the resulting loss of water must be mitigated by an emergency core cooling system (ECCS) to maintain core cooling after shutdown. The U.S. Nuclear Regulatory Commission (NRC) is close to allowing a risk-informed design basis break size, called the Transition Break Size (TBS), to be used for LOCA break size assumptions for ECCS design. The TBS approach will require full safety redundancy for an ECCS system sized to handle a break of the next largest reactor coolant pipe size (rather than the largest reactor coolant pipe size), and it will allow relaxed system redundancy requirements for handling the largest pipe break size. The TBS will thereby reduce the cost of the safety-grade ECCS system in new plant designs and will increase operational flexibility in existing plants. The TBS approach is based on the results of NRC elicitation studies with piping experts regarding historical pipe performance and risk of sudden failure. The approach is non-deterministic and is a conceptual change from the largest-pipe-size break assumption. The conceptual discontinuity between deterministic and elicitation-based break size assumptions could be uncomfortable for those schooled in strictly deterministic accident analyses. In this paper we explore the “leak-before-break” (LBB) methodology as it applies to large pipe break analyses in nuclear piping systems, and show through examples that the elicitation-based TBS approach is indeed conservative when TBS results are compared with deterministic LBB evaluations of similar piping systems. Thus, LBB provides a deterministic means for showing defense in depth against LOCAs greater than the TBS break size.

Experiments and Analytical Simulation Work on an Innovative Steam-Injector-Driven Passive Core Injection Cooling System

2007 ◽  
Author(s):  
Michitsugu Mori ◽  
Tadashi Narabayashi ◽  
Shuichi Ohmori ◽  
Fumitoshi Watanabe
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
High Pressures ◽  
Cooling System ◽  
Steam Pressure ◽  
Injection System ◽  
The Core ◽  
Coolant Injection ◽  
Core Cooling

A Steam Injector (SI) is a simple, compact, passive pump which also functions as a high-performance direct-contact compact heater. We are developing this innovative concept by applying the SI system to core injection systems in Emergency Core Cooling Systems (ECCS) to further improve the safety of nuclear power plants. Passive ECCS in nuclear power plants would be inherently very safe and would prevent serious accidents by keeping the core covered with water (Severe Accident-Free Concept). The Passive Core Coolant Injection System driven by a high-efficiency SI is one that, in an accident such as a loss of coolant accident (LOCA), attains a higher discharge pressure than the supply steam pressure used to inject water into the reactor by operating the SI using water stored in the pool as the water supply source and steam contained in the reactor as the source of pressurization energy. The passive SI equipment would replace large, rotating machines such as pumps and motors, so eliminating the possibility of such equipment failing. In this Si-driven Passive Core Coolant Injection System (SI-PCIS), redundancy will be provided to ensure that the water and steam supply valves to the SI open reliably, and when the valves open, the SI will automatically start to inject water into the core to keep the core covered with water. The SI used in SI-PCIS works for a range of steam pressure conditions, from atmosphere pressure through to high pressures, as confirmed by analytical simulations which were done based on comprehensive experimental data obtained using reduced scale SI. We did further simulations and evaluations of the core cooling and coolant injection performance of SI-PCIS in BWR using RETRAN-3D code, developed using EPRI and other utilities, for the case of small LOCA. Reactors equipped with passive safety systems — the gravity-driven core cooling/injection system (GDCS) and depressurization valves (DPV) — would inevitably end up having large LOCA, even if they are initially small LOCA, as depressurization valves are forcibly opened in order to inject coolant from the GDCS pool to the GDCS water head at up to ∼0.2MPa. On the other hand, our simulation demonstrated that SI-PCIS could prevent large LOCA occurring in reactors by having by coolant discharged into the core in the event of small LOCA or when DPV unexpectedly open.

Experimental Study on the Performance of IIST Passive Core Cooling System

2002 ◽  
Author(s):  
Chin-Jang Chang ◽  
Chien-Hsiung Lee ◽  
Wen-Tan Hong ◽  
Lance L. C. Wang
Keyword(s):  
Nuclear Power ◽  
Cooling System ◽  
Heat Removal ◽  
General System ◽  
System Response ◽  
Water Storage Tank ◽  
The Core ◽  
Third Stage ◽  
Core Cooling

The purpose of this study is to conduct the experiments at the Institute of Nuclear Energy Research (INER) Integral System Test (IIST) facility for evaluation of the performance of the passive core cooling system (PCCS) during the cold-leg small break loss-of-coolant accidents (SBLOCAs). Five experiments were performed with (1) three different break sizes, 2%, 0.5%, and 0.2% (approximately corresponding to 1 1/4”, 2”, and 4” breaks for Maanshan nuclear power plant), and (2) 0.2% and 0.5% without actuation of the first-stage and third-stage automatic depressurization valve (ADS-1 and ADS-3) to initiate PCCS for assessing its capacity in accident management. The detailed descriptions of general system response and the interactions of core makeup tanks (CMTs), accumulators (ACCs), automatic depressurization system (ADS), passive residual heat Removal (PRHR), and in-containment refueling water storage tank (IRWST) on the core heat removal are included. The results show: (1) core long term cooling can be maintained for all cases following the PCCS procedures, (2) the core can be covered for the cases of the 0.2% and 0.5% breaks without actuation of ADS-1 and ADS-3.

Revised LOCA Frequency Estimates From an Expert Elicitation Process

2008 ◽  
Author(s):  
Paul M. Scott ◽  
Robert Lee Tregoning ◽  
Lee Richard Abramson
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Cooling System ◽  
Nuclear Regulatory Commission ◽  
Expert Elicitation ◽  
Sensitivity Analyses ◽  
Piping System ◽  
Plant Design ◽  
Elicitation Process ◽  
Core Cooling

The double-ended-guillotine break (DEGB) criterion of the largest primary piping system in the plant, which generally provides the limiting condition for the emergency core cooling system requirements, is widely recognized as an extremely unlikely event. As a result, the US Nuclear Regulatory Commission (NRC) is considering a risk-informed revision of the design-basis break size requirements for commercial nuclear power plants. In support of this effort, loss-of-coolant accident (LOCA) frequency estimates were developed using an expert elicitation process by consolidating service history data and insights from probabilistic fracture mechanics (PFM) studies with knowledge of plant design, operation, and material performance. This paper describes, and presents the results for, two of the sensitivity analyses conducted as part of this effort (overconfidence adjustment and aggregation method) to examine the assumptions, structure, and techniques used to process the elicitation responses to develop group estimates of the LOCA frequency estimates.

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