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.

Technology Lecture

1983 ◽  
Vol 385 (1789) ◽  
pp. 241-251 ◽  
Keyword(s):  
Nuclear Power ◽  
Task Force ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Nuclear Accidents ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Turbine Generators ◽  
Set Up ◽  
Regulatory Commission

The Central Electricity Generating Board propose to build a pressurized water reactor at Sizewell in Suffolk. The PWR Task Force was set up in June 1981 to provide a communications centre for developing firm design proposals for this reactor. These were to follow the Standardized Nuclear Unit Power Plant System designed by Bechtel for the Westinghouse nuclear steam supply system for reactors built in the United States. Changes were required to the design to accommodate, for example, the use of two turbine generators and to satisfy British safety requirements. Differences exist between the British and American licensing procedures. In the U.K. the statutory responsibility for the safety of a nuclear power station rests unambiguously with the Generating Boards. In the U.S.A. the Nuclear Regulatory Commission issues detailed written instructions, which must be followed precisely. Much of the debate on the safety of nuclear powrer focuses on the risks of big nuclear accidents. It is necessary to explain to the public what, in a balanced perspective, the risks of accidents actually are. The vocabulary used in the nuclear power industry contributes to the misunderstanding and fear felt by the general public. The long-term consequences of big nuclear accidents can be presented in terms of reduction in life expectancy, increased chance of cancer or the equivalent pattern of compulsory cigarette smoking.

The AP1000® PWR Project Moving Toward Completion in China

2013 ◽  
Author(s):  
Grenville Harrop
Keyword(s):  
Nuclear Power ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Fuel Load ◽  
Modular Construction ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Supply Chain Strategy ◽  
Power Company ◽  
Regulatory Commission

The AP1000® pressurized water reactor (PWR) is the first and only Generation III+ nuclear power plant to be granted design certification by the United States Nuclear Regulatory Commission. The initial deployment of this technology has been the construction of dual AP1000 units in each of two coastal sites in the People’s Republic of China (PRC), at Sanmen (Zhejiang Province) and Haiyang (Shandong Province). The contracts for these units were framed to support the PRC’s intention to achieve self reliance in its nuclear supply infrastructure. Westinghouse is implementing its innovative supply chain strategy, “We Buy Where We Build”™, to promote the technology transfer and increasing levels of localization needed as each unit is constructed. Since the initial contract award in 2007, the Westinghouse Consortium and the purchasers, State Nuclear Power Technology Corporation of China (SNPTC), the Shandong Nuclear Power Company (SDNPC), and the Sanmen Nuclear Power Company (SMNPC) have worked in harmony to build the units using advanced modular construction techniques that reduce construction timescales and associated risks. First-of-a-kind (FOAK) plant components have been manufactured and delivered, including reactor vessels, steam generators, and other safety equipment. With construction and equipment installation in the final stages, the planning and implementation of the pre-operational testing, system turnover, and commissioning are now underway to prepare for fuel load and future commercial operation.

A Risk-Informed Methodology for ASME Section XI, Appendix G

2012 ◽  
Vol 134 (3) ◽  
Author(s):  
Ronald Gamble ◽  
William Server ◽  
Bruce Bishop ◽  
Nathan Palm ◽  
Carol Heinecke
Keyword(s):  
Nuclear Regulatory Commission ◽  
The United States ◽  
Service Level ◽  
Pressure Vessels ◽  
Operational Flexibility ◽  
Pressurized Water ◽  
Leak Testing ◽  
Alternative Methodology ◽  
Water Reactors ◽  
Regulatory Commission

The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code [1], Section XI, Appendix G provides a deterministic procedure for defining Service Level A and B pressure–temperature limits for ferritic components in the reactor coolant pressure boundary. An alternative risk-informed methodology has been developed for ASME Section XI, Appendix G. This alternative methodology provides easy to use procedures to define risk-informed pressure–temperature limits for Service Level A and B events, including leak testing and reactor start-up and shut-down. Risk-informed pressure–temperature limits provide more operational flexibility, particularly for reactor pressure vessels with relatively high irradiation levels and radiation sensitive materials. This work evaluated selected plants spanning the population of pressurized water reactors (PWRs) and boiling water reactors (BWRs). The evaluation included determining appropriate material properties, reviewing operating history and system operational constraints, and performing probabilistic fracture mechanics (PFM) analyses. The analysis results were used to define risk-informed pressure–temperature relationships that comply with safety goals defined by the United States (U.S.) Nuclear Regulatory Commission (NRC). This alternative methodology will provide greater operational flexibility, especially for Service Level A and B events that may adversely affect efficient and safe plant operation, such as low-temperature-over-pressurization for PWRs and system leak testing for BWRs. Overall, application of this methodology can result in increased plant efficiency and increased plant and personnel safety.

Status of the United States Nuclear Regulatory Commission Pressurized Thermal Shock Rule Re-Evaluation Project

2002 ◽  
Author(s):  
Terry L. Dickson ◽  
Shah N. Malik ◽  
Mark T. Kirk ◽  
Deborah A. Jackson
Keyword(s):  
United States ◽  
Thermal Shock ◽  
Computational Models ◽  
Structural Integrity ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Pressurized Water ◽  
Pressurized Thermal Shock ◽  
Computational Methodology ◽  
Regulatory Commission

The current federal regulations to ensure that nuclear reactor pressure vessels (RPVs) maintain their structural integrity when subjected to transients such as pressurized thermal shock (PTS) events were derived from computational models that were developed in the early to mid 1980s. Since that time, there have been advancements in relevant technologies associated with the physics of PTS events that impact RPV integrity assessment. Preliminary studies performed in 1999 suggested that application of the improved technology could reduce the conservatism in the current regulations while continuing to provide reasonable assurance of adequate protection to public health and safety. A relaxation of PTS regulations could have profound implications for plant license extension considerations. Based on the above, in 1999, the United States Nuclear Regulatory Commission (USNRC) initiated a comprehensive project, with the nuclear power industry as a participant, to re-evaluate the current PTS regulations within the framework established by modern probabilistic risk assessment (PRA) techniques. During the last three years, improved computational models have evolved through interactions between experts in the relevant disciplines of thermal hydraulics, PRA, human reliability analysis (HRA), materials embrittlement effects on fracture toughness (crack initiation and arrest), fracture mechanics methodology, and fabrication-induced flaw characterization. These experts were from the NRC staff, their contractors, and representatives from the nuclear industry. These improved models have now been implemented into the FAVOR (Fracture Analysis of Vessels: Oak Ridge) computer code, which is an applications tool for performing risk-informed structural integrity evaluations of embrittled RPVs subjected to transient thermal-hydraulic loading conditions. The baseline version of FAVOR (version 1.0) was released in October 2001. The updated risk-informed computational methodology in the FAVOR code is currently being applied to selected domestic commercial pressurized water reactors to evaluate the adequacy of the current regulations and to determine whether a technical basis can be established to support a relaxation of the current regulations. This paper provides a status report on the application of the updated computational methodology to a commercial pressurized water reactor (PWR) and discusses the results and interpretation of those results. It is anticipated that this re-evaluation effort will be completed in 2002.

Large Break LOCA Safety Injection Sensitivity for a CE/ABB System 80+ PWR

2002 ◽  
Author(s):  
J. Pottorf ◽  
S. M. Bajorek
Keyword(s):  
Technical Specification ◽  
Nuclear Regulatory Commission ◽  
Injection System ◽  
Engineering System ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Core Recovery ◽  
Regulatory Commission ◽  
Saturated Boiling

A WCOBRA/TRAC model of an evolutionary pressurized water reactor with direct vessel injection was constructed using publicly available information and a hypothetical double-ended guillotine break of a cold leg pipe was simulated. The model is an approximation of a ABB/Combustion Engineering System 80+ pressurized water reactor (PWR). WCOBRA/TRAC is the thermal-hydraulics code approved by the U.S. Nuclear Regulatory Commission for use in realistic large break LOCA analyses of Westinghouse 3- and 4-loop PWRs and the AP600 passive design. The AP600 design uses direct vessel injection, and the applicability of WCOBRA/TRAC to such designs is supported by comparisons to appropriate test data. This study extends the application of WCOBRA/TRAC to the investigation of the predicted behavior of direct vessel injection in an evolutionary design. A series of large break LOCA simulations were performed assuming a core power of 3914 MWt, and Technical Specification limits of 2.5 on total peaking factor and 1.7 on hot channel enthalpy rise factor. Two cladding temperature peaks were predicted during reflood, one following bottom of core recovery and a second caused by saturated boiling of water in the downcomer. This behavior is consistent with prior WCOBRA/TRAC calculations for some Westinghouse PWRs. The simulation results are described, and the sensitivity to failure assumptions for the safety injection system is presented.

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.

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.

Use of Risk Insights to Enhance Safety Focus of Small Modular Reactor Reviews

2011 ◽  
Author(s):  
Stewart L. Magruder
Keyword(s):  
Review Process ◽  
Nuclear Regulatory Commission ◽  
Pressurized Water Reactor ◽  
Design Features ◽  
Pressurized Water ◽  
Analysis And Evaluation ◽  
Small Modular Reactor ◽  
Efficiency And Effectiveness ◽  
Regulatory Commission ◽  
Regulatory Controls

The U.S. Nuclear Regulatory Commission staff plans to apply a more integrated, graded approach to the review of small modular reactor (SMR) pre-application activities and design applications. The concept is to improve the efficiency and effectiveness of the reviews by focusing on safety significant structures, systems, and components (SSCs). The unique design features associated with SMRs and knowledge gained reviewing other passive reactor designs present opportunities to risk-inform the SMR design certification process to a greater extent than previously employed. The review process can be modified for SMR applications by considering the aggregate of regulatory controls pertaining to SSCs as part of the review and determining those regulatory controls which may supplement or replace, as appropriate, part of the technical or engineering analysis and evaluation. Risk insights acquired from staff reviews of passive LWR designs (i.e. AP1000, ESBWR) can also be incorporated into the review process. Further, risk insights associated with integral pressurized water reactor (iPWR) design features (i.e. Underground facilities impact on turbine missiles review) can be incorporated into the review process.

Modeling Improvements for System Code Evaluation of Inlet Plenum Mixing Under Severe Accident Conditions Using CFD Predictions

2010 ◽  
Author(s):  
Christopher Boyd ◽  
Kenneth Armstrong
Keyword(s):  
Natural Circulation ◽  
Nuclear Regulatory Commission ◽  
Mixing Model ◽  
Reactor Vessel ◽  
Severe Accident ◽  
Dimensional System ◽  
Pressurized Water ◽  
Surge Line ◽  
Accident Conditions ◽  
Regulatory Commission

An updated mixing model is developed for application to system codes used for predicting severe accident-induced failures of steam generator (SG) U-tubes in a pressurized-water reactor. Computational fluid dynamics is used to predict the natural circulation flows between a simplified reactor vessel and the primary side of an SG during a hypothesized severe accident scenario. The results from this analysis are used to extend earlier experimental results and predictions. These new predictions benefit from the inclusion of the entire natural circulation loop between the reactor vessel upper plenum and the SG. Tube leakage and mass flow into the pressurizer surge line also are considered. The predictions are utilized as a numerical experiment to improve the basis for simplified models applied in one-dimensional system codes that are used during the prediction of severe accident natural circulation flows. An updated inlet plenum mixing model is proposed that accounts for mixing in the hot leg as well as the inlet plenum region. The new model is consistent with the predicted behavior and can account for flow into a side-mounted pressurizer surge line if present. Sensitivity studies demonstrate the applicability of the approach over a range of conditions. The predictions are most sensitive to changes in the SG secondary side temperatures or heat-transfer rates at the SG tubes. Grid independence is demonstrated through comparisons with previous models and by increasing the number of cells in the model. This work supports the U.S. Nuclear Regulatory Commission (NRC) studies of SG tube integrity under severe accident conditions.

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