Containment Sump Active Strainers

2005 ◽  
Author(s):  
A. R. Mehta ◽  
A. J. Bilanin ◽  
J. Hamel ◽  
A. Kaufman
Keyword(s):  
Nuclear Power ◽  
Cooling System ◽  
Safety Issue ◽  
Water Source ◽  
Nuclear Regulatory Commission ◽  
Test Facility ◽  
Plant Design ◽  
Test Environment ◽  
Pressurized Water

The containment sump, also known as emergency or recirculation sump, is part of the Emergency Core Cooling System (ECCS). Every nuclear power plant is required by regulations to have an ECCS to mitigate a design basis accident. The containment sump of a Pressurized Water Reactor (PWR) collects reactor coolant and chemically reactive spray solutions following a Loss of Coolant Accident (LOCA). The containment sump serves as the water source to support long-term recirculation. This water source, the related pump inlets and the piping between the source and inlets are all important safety components. Suppression pools in Boiling Water Reactors (BWRs) serve the same purpose as PWR containment sumps. Historically, a passive debris screen has been used to prevent debris from entering the ECCS suction lines surrounding the containment sump. Previous incidents demonstrated that the potential for excessive head loss across the containment sump screens exists because of the accumulation of debris on the containment sump. Because of this, the US Nuclear Regulatory Commission (NRC) has concluded that containment sump blockage is a potential concern for PWRs. US BWRs were required to conduct plant-specific evaluations of their suction strainer performance and, as required, modify their plant design. While all US PWRs are required to resolve this Generic Safety Issue (GSI-191), containment sump blockage continues to be a major concern for both BWRs and PWRs internationally. This paper describes the GE Active Strainer design, one of several strainers developed to resolve this generic safety issue. The Active Strainer presents an innovative and novel method of addressing containment sump blockage. This strainer employs a rotating, or “active”, plow and brush that sweep over a perforated surface. By keeping the perforated surface free of debris, fluid is allowed to pass through, providing sufficient coolant to the ECCS pumps to support long-term recirculation. Due to the unique method by which the Active Strainer filters coolant, a test program was developed to demonstrate its functionality and viability. Intrinsic differences between passive and active solutions make previous methods of testing obsolete for the GE Active Strainer. Moreover, the complex and varying geometries and conditions of actual plant containment sumps are difficult to replicate. Therefore, a methodology was developed to ensure prototypical test environment and strainer debris loads in a scaled test facility. This paper will discuss the GE Active Strainer design, the testing conducted and subsequent conclusions.

Post Small-Break LOCA Long Term Core Cooling Performance in ACME Test Facility

2016 ◽  
Author(s):  
Zhanfei Qi ◽  
Sheng Zhu
Keyword(s):  
Nuclear Power ◽  
Cooling System ◽  
Safety Margin ◽  
Test Facility ◽  
Passive Safety ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Safety Features ◽  
Core Cooling

CAP1400 Pressurized Water Reactor is developed by China’s State Nuclear Power Technology Corporation (SNPTC) based on the passive safety concept and advanced system design. The Advanced Core-cooling Mechanism Experiment (ACME) integral effect test facility, which was constructed at Tsinghua University, represents a 1/3-scale height of CAP1400 RCS and passive safety features. It is designed to simulate the performance of CAP1400 passive core cooling system in the small break loss of coolant accidents (SBLOCA) for design certification, safety review and safety analysis code development. The Long Term Core Cooling (LTCC) post-LOCA could be simulated by ACME as well. A series of test cases with various break sizes and locations with post-LOCA LTCC period were conducted in ACME facility. This paper describes the post-LOCA LTCC test conducted in ACME test facility. The LTCC phenomena in different cases are very similar. It’s found that the interval that switching from IRWST injection to sump recirculation has the least safety margin. However, it’s shown that the post-LOCA LTCC in ACME could be well maintained by passive core cooling system according to the test results even though the recirculation water level in ACME IRWST-2 is lower than the containment recircualtion level in CAP1400 conservatively.

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.

Head Loss Through Fibrous Beds Generated on Different Types of Containment Sump Strainers

2014 ◽  
Author(s):  
Saya Lee ◽  
Suhaeb Abdulsattar ◽  
Yassin A. Hassan
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Cooling System ◽  
Safety Issue ◽  
Theoretical Models ◽  
Nuclear Regulatory Commission ◽  
High Energy ◽  
Head Loss ◽  
Debris Accumulation ◽  
Approach Velocity

During a Loss of Coolant Accident (LOCA), the high energy jet from the break may impinge on surrounding surfaces and materials, producing a relatively large amount of fibrous debris (mostly insulation materials). The debris may be transported through the reactor containment and reach the sump strainers. Accumulation of such debris on the strainers’ surface can cause a loss of Net Positive Suction Head (NPSH) and negatively affect the Emergency Core Cooling System (ECCS) capabilities. The U.S. Nuclear Regulatory Commission (U.S.NRC) initiated the Generic Safety Issue (GSI) 191 to understand the physical phenomena involved in this type of event, and help develop the tools to prove the safety and reliability of the existing Light Water Reactors (LWR) under these conditions. Some nuclear power plants have already adopted countermeasures in an attempt to limit the effect of the debris accumulation on the ECCS performance, by replacing or modifying the existing strainer configurations. In this paper, two different strainer designs have been considered and sensitivity analysis was conducted to study the effect of the approach velocity on the pressure drop at the strainer caused by the debris accumulation. The development of the fibrous beds was visually recorded in order to correlate the head loss, the approach velocity, and the thickness of the fibrous bed. The experimental results were compared to semi-empirical models and theoretical models proposed by previous researchers.

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.

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.

Alternate Requirements for Protection Against Pressurized Thermal Shock (PTS)

2013 ◽  
Author(s):  
Stephen M. Parker ◽  
Nathan A. Palm ◽  
Xavier Pitoiset
Keyword(s):  
Thermal Shock ◽  
Nuclear Power ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Plant Design ◽  
Pressurized Water ◽  
Screening Criteria ◽  
Pressurized Thermal Shock ◽  
Prediction Techniques ◽  
The U.S

Plants in the United States (U.S.) and many plants outside of the U.S. are required to meet the regulations of the Pressurized Thermal Shock (PTS) Rule, 10 CFR 50.61. The Alternate Pressurized Thermal Shock (PTS) Rule (10 CFR 50.61a) was approved by the U.S. Nuclear Regulatory Commission (NRC) and included in the Federal Register, with an effective date of February 3, 2010. This Alternate Rule provides a new metric and screening criteria for PTS. This metric, RTMAX-X, and the corresponding screening criteria are far less restrictive than the RTPTS metrics and screening criteria in the original PTS Rule (10 CFR 50.61). The Alternate PTS Rule was developed through probabilistic fracture mechanics (PFM) evaluations performed for selected U.S. pilot plants. A Generalization Study was also performed which determined that the plants used for these evaluations were representative of and applicable to the U.S. Pressurized Water Reactor (PWR) nuclear power plant fleet. Plants outside of the U.S. may be interested in implementing the Alternate PTS Rule. However, direct implementation of the Alternate PTS Rule may not be possible due to differences in plant design, embrittlement prediction techniques, inservice inspection requirements, etc. The objective of this paper is to explore the use the Alternate PTS Rule by PWR plants outside of the U.S. by proposing methods to account for the potential differences mentioned above.

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.

Westinghouse AP1000 Solution to Long-Term Cooling Debris Concerns

2009 ◽  
Author(s):  
Terry L. Schulz ◽  
Timothy S. Andreychek ◽  
Yong J. Song ◽  
Kevin F. McNamee
Keyword(s):  
Nuclear Power ◽  
Safety Evaluation ◽  
Nuclear Regulatory Commission ◽  
Head Loss ◽  
Plant Design ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Geographical Regions ◽  
Core Cooling

The AP1000 is a pressurized water reactor with passive safety features and extensive plant simplifications that provides significant and measurable improvements in safety, construction, reliability, operation, maintenance and costs. The design of the AP1000 incorporates a standard approach, which results in a plant design that can be constructed in multiple geographical regions with varying regulatory standards and expectations. The AP1000 uses proven technology, which builds on more than 2,500 reactor years of highly successful Westinghouse PWR operation. The AP1000 received Final Design Approval by the Nuclear Regulatory Commission in September 2004. The AP1000 Nuclear Power Plant uses natural recirculation of coolant to cool the core following a postulated Loss Of Coolant Accident (LOCA). Recirculation screens are provided in strategic areas of the plant to remove debris that might migrate with the water in containment and adversely affect core cooling. The approach used to avoid the potential for debris to plug the AP1000 recirculation screens is consistent with the guidance identified in Regulatory Guide 1.82 Revision 3, the Pressurized Water Reactor (PWR) Industry Guidance of NEI 04–07, and the Nuclear Regulatory Commission’s Safety Evaluation on NEI 04–07. Various contributors to screen plugging were considered, including debris that could be produced by a LOCA, resident containment debris and post accident chemical products that might be generated in the coolant pool that forms on the containment floor post-accident. The solution developed for AP1000 includes three major aspects, including the elimination of debris sources by design, features that prevent transportation of debris to the screens and the use of large advanced screen designs. Measures were taken to design out debris sources including fibers, particles and chemicals. Available industry data from walkdowns in existing plants is used to determine the characteristics and amounts of the fibrous and particulate debris that could exist in containment prior to the LOCA. Materials used in the AP1000 containment are selected to eliminate post accident chemical debris generation. Large, advanced screen designs that can tolerate significant quantities of debris have been incorporated. Testing has been performed which demonstrates that the AP1000 screens will have essentially no head loss considering the debris that could be transported to them. Testing has also been performed on an AP1000 fuel assembly that demonstrates that it will also have essentially no head loss considering the debris that could be transported to it.

Use of HFE Guidance for the Review of Nuclear Power Plants

2020 ◽  
Vol 64 (1) ◽  
pp. 1-5
Author(s):  
John O’Hara ◽  
Stephen Fleger
Keyword(s):  
Interface Design ◽  
Nuclear Power ◽  
Power Plants ◽  
Health And Safety ◽  
Nuclear Regulatory Commission ◽  
Program Review ◽  
Human Factors Engineering ◽  
Plant Design ◽  
Review Model ◽  
Regulatory Commission

The U.S. Nuclear Regulatory Commission (NRC) evaluates the human factors engineering (HFE) of nuclear power plant design and operations to protect public health and safety. The HFE safety reviews encompass both the design process and its products. The NRC staff performs the reviews using the detailed guidance contained in two key documents: the HFE Program Review Model (NUREG-0711) and the Human-System Interface Design Review Guidelines (NUREG-0700). This paper will describe these two documents and the method used to develop them. As the NRC is committed to the periodic update and improvement of the guidance to ensure that they remain state-of-the-art design evaluation tools, we will discuss the topics being addressed in support of future updates as well.

scholarly journals Simulation of MASPn Experiments in MISTRA Test Facility with COCOSYS Code

2008 ◽  
Vol 2008 ◽  
pp. 1-7 ◽  
Author(s):  
Mantas Povilaitis ◽  
Egidijus Urbonavičius
Keyword(s):  
Boundary Conditions ◽  
Nuclear Power ◽  
Power Plants ◽  
Volume Fraction ◽  
Characteristic Length ◽  
Initial Conditions ◽  
Test Facility ◽  
Work Package ◽  
Acceptable Agreement

An issue of the stratified atmospheres in the containments of nuclear power plants is still unresolved; different experiments are performed in the test facilities like TOSQAN and MISTRA. MASPn experiments belong to the spray benchmark, initiated in the containment atmosphere mixing work package of the SARNET network. The benchmark consisted of MASP0, MASP1 and MASP2 experiments. Only the measured depressurisation rates during MASPn were available for the comparison with calculations. When the analysis was performed, the boundary conditions were not clearly defined therefore most of the attention was concentrated on MASP0 simulation in order to develop the nodalisation scheme and define the initial and boundary conditions. After achieving acceptable agreement with measured depressurisation rate, simulations of MASP1 and MASP2 experiments were performed to check the influence of sprays. The paper presents developed nodalisation scheme of MISTRA for the COCOSYS code and the results of analyses. In the performed analyses, several parameters were considered: initial conditions, loss coefficient of the junctions, initial gradients of temperature and steam volume fraction, and characteristic length of structures. Parametric analysis shows that in the simulation the heat losses through the external walls behind the lower condenser installed in the MISTRA facility determine the long-term depressurisation rate.

Sign in / Sign up

Export Citation Format

Share Document