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.

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.

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.

On the Thermal-Hydraulic Essentials of the Holtec Inherently Safe Modular Underground Reactor (HI-SMUR) System

2011 ◽  
Author(s):  
Kris Singh ◽  
Indresh Rampall ◽  
Joseph Rajkumar
Keyword(s):  
Reactor Core ◽  
Reactor Vessel ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Control Rod ◽  
Hydraulic Conditions ◽  
Deep Underground ◽  
The Stability ◽  
Core Cooling

HI-SMUR 140 is a small (145 MWe) modular pressurized water reactor designed to harness fission energy without the use of a recirculation pump. HI-SMUR’s core resides deep underground in a thick-walled reactor vessel (RV) enclosed by a stainless steel-lined reinforced concrete “Reactor Well”. The HI-SMUR Nuclear Steam Supply System (NSSS) is a conjugated pressure vessel assemblage wherein the steam generator(s) are integrally joined to the RV, i.e., without any interconnecting piping, and the pressurizer is integral to the reactor vessel. There are no penetrations in the RV for over a height of 120 feet above the reactor core, which precludes the scenario of loss of coolant to the core from a postulated pipe break event. The rejection of decay heat from the reactor in the wake of a scram is engineered to occur without the aid of on-site or off-site power, making the HI-SMUR NSSS demonstrably capable of withstanding a cataclysmic environmental phenomenon of Fukushima’s intensity without loss of core cooling or without precipitating any damage to the plant or the surrounding environment. The system design places a premium on accessibility and maintainability of vital equipment such as the steam generators, RV internals, and the control rod drive assemblies. This paper is the first in a series of papers planned to explain and quantify the performance and safety aspects of HI-SMUR 140. In this paper, the thermal-hydraulic characteristics of the HI-SMUR NSSS are explored using classical hydraulic correlations which have served as the tool for the scoping parametric study of the system. In particular, the stability of the system under varying power output conditions and the long-term reliability of the fuel under the most adverse thermal/hydraulic conditions are presented.

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.

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.

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.

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.

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.

Fracture mechanics in design and service: ‘living with defects’ - Fracture mechanics as an aid to design and operation of nuclear plant

1981 ◽  
Vol 299 (1446) ◽  
pp. 227-236 ◽  
Keyword(s):  
Fracture Mechanics ◽  
Material Properties ◽  
Pressure Vessels ◽  
Nuclear Plant ◽  
Plant Design ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Plant Safety ◽  
Usual Criterion ◽  
Design Protection

Fracture mechanics analyses are an important part of nuclear plant design, supplementing the conventional design protection against failure to cover the possibility of the presence of crack-like defects. The degree of detail and accuracy required for a particular application depends on the possible consequences of a failure and whether the assessment is concerned with plant safety or with aspects of reliability. In the former case, a conservative approach is necessary and the prevention of initiation is the usual criterion. This approach is typified by the safety assessment applied to pressurized water reactor pressure vessels, which is outlined and discussed in relation to elastic plastic approaches and the importance of plant transient conditions, material properties (especially in weldments) and possible defect distributions. Fracture mechanics can help in defining quality control and quality assurance procedures, including both requirements for mechanical property appraisal and nondestructive testing. The latter aspects extend into operation, in respect of monitoring of plant conditions, surveillance of changes in material properties and the use of periodic inspection and plant condition monitoring techniques. A number of examples are quoted and recommendations made to permit improved fracture mechanics assessments.

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