Sensitivity Study of LOCA-Generated Debris on the Pressure Drop of Fuel Assembly

2017 ◽  
Author(s):  
Hammad Aslam Bhatti ◽  
Zhangpeng Guo ◽  
Weiqian Zhuo ◽  
Shahroze Ahmed ◽  
Da Wang ◽  
...  
Keyword(s):  
Pressure Drop ◽  
Fuel Assembly ◽  
Cooling System ◽  
Safety Issue ◽  
Cooling Water ◽  
Sensitivity Study ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Loss Of Coolant Accident ◽  
Core Cooling

The coolant of emergency core cooling system (ECCS), for long-term core cooling (LTCC), comes from the containment sump under the loss-of-coolant accident (LOCA). In the event of LOCA, within the containment of the pressurized water reactor (PWR), thermal insulation of piping and other materials in the vicinity of the break could be dislodged. A fraction of these dislodged insulation and other materials would be transported to the floor of the containment by coolant. Some of these debris might get through strainer and eventually accumulate over the suction sump screens of the emergency core cooling systems (ECCS). So, these debris like fibrous glass, fibrous wool, chemical precipitates and other particles cause pressure drop across the sump screen to increase, affecting the cooling water recirculation. As to address this safety issue, the downstream effect tests were performed over full-scale mock up fuel assembly. Sensitivity studies on pressure drop through LOCA-generated debris, deposited on fuel assembly, were performed to evaluate the effects of debris type and flowrate. Fibrous debris is the most crucial material in terms of causing pressure drop, with fibrous wool (FW) debris being more efficacious than fibrous glass (FG) debris.

Experimental Study of Debris-Loading Head Loss Test for PWR Fuel Assembly After LOCA

2017 ◽  
Author(s):  
Da Wang ◽  
Fenglei Niu ◽  
Weiqian Zhuo ◽  
Jingwen Ren ◽  
Zhangpeng Guo ◽  
...  
Keyword(s):  
Fuel Assembly ◽  
Safety Issue ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Pressure Drops ◽  
Loss Of Coolant Accident ◽  
Sensitivity Studies ◽  
Pass Through ◽  
Core Cooling

A loss of coolant accident (LOCA) in a PWR (pressurized water reactor) would generate debris from thermal insulation and other materials in the vicinity of the break. It is postulated that debris can transported to the containment sump strainer. Some of the debris may pass through the strainer and could challenge the long-term core cooling capability of the plant. To address this safety issue, the downstream effect tests for the PWR were performed. Sensitivity studies on pressure drops through LOCA-generated debris deposited on a fuel assembly were performed to evaluate the effects of debris type and flow rate. Fibrous debris is the most crucial material in terms of causing pressure drops.

IIST Passive Core Cooling on Pressurizer Top Break

2003 ◽  
Author(s):  
Tay-Jian Liu ◽  
Chien-Hsiung Lee
Keyword(s):  
Cooling System ◽  
Test Results ◽  
Pressurized Water Reactor ◽  
Relief Valve ◽  
Pressurized Water ◽  
Hydraulic Behavior ◽  
Loss Of Coolant Accident ◽  
Accident Management ◽  
Loss Of Coolant ◽  
Core Cooling

Two experiments for small-break loss-of-coolant accident on pressurizer top were conducted at the INER Integral System Test (IIST) facility to investigate thermal-hydraulic behavior of a passive core cooling system (PCCS) in a Westinghouse pressurized water reactor (PWR). The test results are compared with the previous IIST tests under the same initial and boundary conditions for a power-operated relief valve (PORV) stuck-open incident. The objectives of this study are to understand of the key thermal-hydraulic phenomena associated with PCCS and to compare the effectiveness of accident management with or without PCCS. The break sizes were scaled down based on one and all three fully-opened PORVs. This paper identified the key phenomena commonly observed and the phenomena unique to a PWR with PCCS.

scholarly journals Evaluation of passive autocatalytic recombiners operation efficiency by means of the lumped parameter approach

Nukleonika ◽  
2015 ◽  
Vol 60 (2) ◽  
pp. 339-345 ◽  
Author(s):  
Tomasz Bury
Keyword(s):  
Hydrogen Generation ◽  
Cooling System ◽  
Component Mixture ◽  
Pressurized Water ◽  
Water Reactor ◽  
Safety Issues ◽  
Loss Of Coolant Accident ◽  
Lumped Parameter ◽  
Core Cooling ◽  
Parameter Approach

Abstract The problem of hydrogen behavior in containment buildings of nuclear reactors belongs to thermal-hydraulic area. Taking into account the size of systems under consideration and, first of all, safety issues, such type of analyses cannot be done by means of full-scale experiments. Therefore, mathematical modeling and numerical simulations are widely used for these purposes. A lumped parameter approach based code HEPCAL has been elaborated in the Institute of Thermal Technology of the Silesian University of Technology for simulations of pressurized water reactor containment transient response. The VVER-440/213 and European pressurised water reactor (EPR) reactors containments are the subjects of analysis within the framework of this paper. Simulations have been realized for the loss-of-coolant accident scenarios with emergency core cooling system failure. These scenarios include core overheating and hydrogen generation. Passive autocatalytic recombiners installed for removal of hydrogen has been taken into account. The operational efficiency of the hydrogen removal system has been evaluated by comparing with an actual hydrogen concentration and flammability limit. This limit has been determined for the three-component mixture of air, steam and hydrogen. Some problems related to the lumped parameter approach application have been also identified.

scholarly journals Thermal Hydraulic Analysis of a Nuclear Reactor due to Loss of Coolant Accident with and without Emergency Core Cooling System

2020 ◽  
Vol 01 (02) ◽  
pp. 53-60
Author(s):  
Pronob Deb Nath ◽  
Kazi Mostafijur Rahman ◽  
Md. Abdullah Al Bari
Keyword(s):  
Nuclear Reactor ◽  
Cooling System ◽  
Computational Study ◽  
Reactor Core ◽  
Primary Circuit ◽  
Feed Water ◽  
Pressurized Water ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Core Cooling

This paper evaluates the thermal hydraulic behavior of a pressurized water reactor (PWR) when subjected to the event of Loss of Coolant Accident (LOCA) in any channel surrounding the core. The accidental break in a nuclear reactor may occur to circulation pipe in the main coolant system in a form of small fracture or equivalent double-ended rupture of largest pipe connected to primary circuit line resulting potential threat to other systems, causing pressure difference between internal parts, unwanted core shut down, explosion and radioactivity release into environment. In this computational study, LOCA for generation III+ VVER-1200 reactor has been carried out for arbitrary break at cold leg section with and without Emergency Core Cooling System (ECCS). PCTRAN, a thermal hydraulic model-based software developed using real data and computational approach incorporating reactor physics and control system was employed in this study. The software enables to test the consequences related to reactor core operations by monitoring different operating variables in the system control bar. Two types of analysis were performed -500% area break at cold leg pipe due to small break LOCA caused by malfunction of the system with and without availability of ECCS. Thermal hydraulic parameters like, coolant dynamics, heat transfer, reactor pressure, critical heat flux, temperature distribution in different sections of reactor core have also been investigated in the simulation. The flow in the reactor cooling system, steam generators steam with feed-water flow, coolant steam flow through leak level of water in different section, power distribution in core and turbine were plotted to analyze their behavior during the operations. The simulation showed that, LOCA with unavailability of Emergency Core Cooling System (ECCS) resulted in core meltdown and release of radioactivity after a specific time.

An Experimental Study on Head Loss of Prototypical Fibrous Debris Beds During Loss-of-Coolant Accident Conditions

2016 ◽  
Vol 2 (3) ◽  
Author(s):  
Amir Ali ◽  
Edward D. Blandford
Keyword(s):  
Safety Issue ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Head Loss ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Debris Accumulation ◽  
Loss Of Coolant Accident ◽  
Accident Conditions ◽  
Regulatory Commission

The United States Nuclear Regulatory Commission (NRC) initiated a generic safety issue (GSI-191) assessing debris accumulation and resultant chemical effects on pressurized water reactor (PWR) sump performance. GSI-191 has been investigated using reduced-scale separate-effects testing and integral-effects testing facilities. These experiments focused on developing a procedure to generate prototypical debris beds that provide stable and reproducible conventional head loss (CHL). These beds also have the ability to filter out chemical precipitates resulting in chemical head loss. The newly developed procedure presented in this paper is used to generate debris beds with different particulate to fiber ratios (η). Results from this experimental investigation show that the prepared beds can provide reproducible CHL for different η in a single and multivertical loops facility within ±7% under the same flow conditions. The measured CHL values are consistent with the predicted values using the NUREG-6224 correlation. Also, the results showed that the prepared debris beds following the proposed procedure are capable of detecting standard aluminum and calcium precipitates, and the head loss increase (chemical head loss) was measured and reported in this paper.

Recapturing Net Positive Suction Head Margins in Boiling Water Reactor Emergency Core Cooling Systems

2017 ◽  
Author(s):  
Alan J. Bilanin ◽  
Andrew E. Kaufman ◽  
Warren J. Bilanin
Keyword(s):  
Cooling System ◽  
Cooling Water ◽  
Boiling Water ◽  
Boiling Water Reactor ◽  
Cooling Systems ◽  
Water Reactor ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Core Cooling ◽  
Reactor Pressure

Boiling Water Reactor pressure suppression pools have stringent housekeeping requirements, as well as restrictions on amounts and types of insulation and debris that can be present in the containment, to guarantee that suction strainers that allow cooling water to be supplied to the reactor during a Loss of Coolant Accident remain operational. By introducing “good debris” into the cooling water, many of these requirements/restrictions can be relaxed without sacrificing operational readiness of the cooling system.

Probabilistic Model for Debris-Induced Loss of Long Term Core Cooling

2008 ◽  
Author(s):  
Heather L. Detar ◽  
Daniel T. McLaughlin ◽  
Robert J. Lutz
Keyword(s):  
Safety Issue ◽  
Management Strategies ◽  
Plant Design ◽  
Pressurized Water Reactor ◽  
Original Design ◽  
Pressurized Water ◽  
Design Basis ◽  
Design Basis Accident ◽  
Core Cooling

Generic Safety Issue (GSI) 191 deals with the potential for generation and transport of debris following a design basis accident that is in excess of quantities assumed in the original design basis and licensing of Pressurized Water Reactor (PWR) plants. In addition to physical modifications to the sump screens to comply with the Generic Letter requirements, some plants have also changed Emergency Operating Procedures (EOPs) to include contingency actions to prevent debris-induced loss of long term core cooling. ASME Probabilistic Risk Assessment (PRA) standard RA-Sb-2005 requires that the plant PRA be maintained and updated to reflect the current plant design and operation. Development of PRA models to quantify the potential for debris-induced loss of long term core cooling supports the PRA updated to reflect the as-built as-operated plant. The PWR Owners Group (PWROG) has undertaken a program to develop a generic PRA model for this issue. The generic PRA model was developed to address the overall plant risk, including new physical and procedural modifications. The model also addresses applications, such as Maintenance Rule screening and the assessment of the risk significance of deviations from the licensing basis analyses. The new PRA model probabilistically treats several facets of the potential for debris-induced challenges to long term core cooling; including debris generation and transport as a function of Reactor Coolant System (RCS) break size and location. The PRA model will permit plant operators to easily incorporate the potential for inadequate core cooling during emergency core cooling recirculation from the containment sump into their PRA Level 1 and Level 2 models. The methodology is based on realistically modeling the conditions that may lead to a debris-induced loss of long term core cooling. The PWROG model also includes consideration of water management strategies being implemented by several PWR plant operators.

Sensitivity Analysis of a PWR Response During a Loss of Coolant Accident Under a Hypothetical Core Blockage Scenario Using RELAP5-3D

2014 ◽  
Author(s):  
Timothy Crook ◽  
Rodolfo Vaghetto ◽  
Alessandro Vanni ◽  
Yassin A. Hassan
Keyword(s):  
Cooling System ◽  
Safety Concern ◽  
System Response ◽  
Flow Paths ◽  
Pressurized Water ◽  
The Core ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Pass Through ◽  
Core Cooling

During a Loss of Coolant Accident (LOCA) a substantial amount of debris may be generated in containment during the blowdown phase. This debris can become a major safety concern since it can potentially impact the Emergency Core Cooling System (ECCS). Debris, produced by the LOCA break flow and transported to the sump, could pass through the filtering systems (debris bed and sump strainer) in the long term cooling phase. If the debris were to sufficiently accumulate at the core inlet region, the core flow could theoretically decrease, affecting the core coolability. Under such conditions, the removal of decay heat would only be possible by coolant flow reaching the core through alternative flow paths, such as the core bypass (baffle). There are certain plant specific features that can play a major role in core cooling from this bypass flow. One of these of key interest is the pressure relief holes. A typical 4-loop Pressurized Water Reactor (PWR) was modeled using RELAP5-3D to simulate the reactor system response during the phases of a large break LOCA and the effectiveness of core cooling under full core blockage was analyzed. The simulation results showed that the presence of alternative flow paths may significantly increase core coolability and prevent cladding temperatures from reaching safety limits, while the lack of LOCA holes may lead to a conservative over-prediction of the cladding temperature.

The Benefits of Using a Risk-Informed Approach to Resolving GSI-191

2012 ◽  
Author(s):  
Timothy D. Sande ◽  
Gilbert L. Zigler ◽  
Ernie J. Kee ◽  
Bruce C. Letellier ◽  
C. Rick Grantom ◽  
...  
Keyword(s):  
Failure Modes ◽  
Cooling System ◽  
Safety Issue ◽  
Acceptance Criteria ◽  
Pressurized Water ◽  
Water Reactor ◽  
Advantages And Disadvantages ◽  
Spray System ◽  
Core Cooling

The emergency core cooling system (ECCS) and containment spray system (CSS) in a pressurized water reactor (PWR) are designed to safely shutdown the plant following a loss of coolant accident (LOCA). The assurance of long term core cooling in PWRs following a LOCA has a long history dating back to the NRC studies of the mid 1980s associated with Unresolved Safety Issue (USI) A-43. Results of the NRC research on boiling water reactor (BWR) ECCS suction strainer blockage of the early 1990s identified new phenomena and failure modes that were not considered in the resolution of USI A-43. As a result of these concerns, Generic Safety Issue (GSI) 191 was identified in September 1996 related to debris clogging of the ECCS sump suction strainers at PWRs. Although plants have taken steps to prevent strainer clogging (by increasing the screen area, for example), satisfactory closure of this issue has proved elusive due to long term cooling issues and the effect of chemical precipitates on head loss. Previous investigators have identified bounding scenarios using conservative inputs, methods, and acceptance criteria. The acceptance criteria are applied in a “pass/fail” fashion that ignores risk. That is, if the results are acceptable, the issue has been resolved. Otherwise, it is necessary to either redo the analysis with partial relaxation of analytical conservatisms or perform additional plant modifications to ensure that the acceptance criteria are met. This article describes a new approach to close out the GSI-191 issue by evaluating the risk associated with ECCS performance on post-LOCA core cooling as a basis to change the plant license. The approach includes an assessment of LOCA frequencies as a function of break size at locations along the reactor coolant system, as well as a full quantification of the uncertainties associated with LOCA frequencies and the generation, transport, accumulation, and subsequent impact of debris on ECCS performance. The overall frameworks for the deterministic and risk-informed approaches are summarized with emphasis on the risk-informed method. The differences between the deterministic approach taken in the past and the new risk-informed approach are described. Advantages and disadvantages between the two methods are described and contrasted for the GSI-191 issue. The South Texas Project (STP) GSI-191 risk-informed closure efforts are presented.

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