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.

Prediction of Weld Residual Stress in a Pressurized Water Reactor Pressurizer Surge Nozzle

2015 ◽  
Vol 138 (2) ◽  
Author(s):  
Akira Maekawa ◽  
Atsushi Kawahara ◽  
Hisashi Serizawa ◽  
Hidekazu Murakawa
Keyword(s):  
Residual Stress ◽  
Heat Source ◽  
High Speed ◽  
Nuclear Regulatory Commission ◽  
Round Robin ◽  
Welding Residual Stress ◽  
3D Analysis ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor

Primary water stress corrosion cracking (PWSCC) phenomenon in dissimilar metal welds is one of the safety issues in ageing pressurized water reactor (PWR) piping systems. It is well known that analysis accuracy of cracking propagation due to PWSCC depends on welding residual stress conditions. The U.S. Nuclear Regulatory Commission (NRC) and the Electric Power Research Institute (EPRI) carried out an international round robin validation program to evaluate and quantify welding residual stress analysis accuracy and uncertainty. In this paper, participation results of the authors in the round robin program were reported. The three-dimensional (3D) analysis based on a fast weld simulation using an iterative substructure method (ISM), was shown to provide accurate results in a high-speed computation. Furthermore, the influence of different heat source models on analysis results was investigated. It was demonstrated that the residual stress and distortion calculated using the moving heat source model were more accurate.

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.

Development of Steam Generator Model Using RELAP5 to Simulate Two-Phase Natural Circulation for Pressurized Water Reactor

2014 ◽  
Author(s):  
Shinya Miyata ◽  
Satoru Kamohara ◽  
Wataru Sakuma ◽  
Hiroaki Nishi
Keyword(s):  
Steam Generator ◽  
Natural Circulation ◽  
Injection System ◽  
Pressurized Water Reactor ◽  
Two Phase ◽  
Pressurized Water ◽  
Water Reactor ◽  
Loss Of Coolant ◽  
Mass Inventory ◽  
Core Cooling

In typical pressurized water reactor (PWR), to cope with beyond design basis events such as station black out (SBO) or small break loss of coolant accident with safety injection system failure, injection from accumulator sustains core cooling by compensating for loss of coolant. Core cooling is sustained by single- or two-phase natural circulation or reflux condensation depending on primary coolant mass inventory. Behavior of the natural circulation in PWR has been investigated in the facilities such as Large Scale Test Facility (LSTF) which is a full-height and full-pressure and thermal-hydraulic simulator of typical four-loop PWR. Two steady-state natural circulation tests were conducted in LSTF at both high and low pressure. These two tests were conducted changing the primary mass inventory as a test parameter, while keeping the other parameters such as core power, steam generator (SG) pressure, and steam generator water level as they are. Mitsubishi Heavy Industries (MHI) plans new natural circulation tests to cover wider range of core power and pressure as test-matrix (including the previous LSTF tests) to validate applicability of the model in wider range of core power and pressure conditions including the SBO conditions. In this paper, the previous LSTF natural circulation tests are reviewed and the new test plan will be described. Additionally, MHI also started a feasibility study to improve the steam generator tube and inlet/outlet plenum model using the M-RELAP5 code [4]. Newly developed model gives reasonable agreement with the previous LSTF tests and applies to the new test conditions. The feasibility findings will also be described in this paper.

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.

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.

Results of Overpressurization Test of a 1:4-Scale Prestressed Concrete Containment Vessel Model

2002 ◽  
Author(s):  
Michael F. Hessheimer ◽  
Satoru Shibata ◽  
James F. Costello
Keyword(s):  
Prestressed Concrete ◽  
Nuclear Power ◽  
Limit State ◽  
Nuclear Regulatory Commission ◽  
Scale Model ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Containment Vessel ◽  
Engineering Corporation ◽  
Regulatory Commission

The Nuclear Power Engineering Corporation (NUPEC) of Japan and the U.S. Nuclear Regulatory Commission (NRC) have been co-sponsoring and jointly funding a Cooperative Containment Research Program at Sandia National Laboratories. The purpose of the program is to investigate the response of representative models of nuclear containment structures to pressure loading beyond the design basis accident and to compare analytical predictions with measured behavior. This is accomplished by conducting static, pneumatic overpressurization tests of scale models at ambient temperature. The first project in this program was a test of a mixed scale steel containment vessel (SCV). Next, a 1:4-scale model of a prestressed concrete containment vessel (PCCV), representative of a pressurized water reactor (PWR) plant in Japan, was constructed by NUPEC at Sandia National Laboratories from January 1997 through June, 2000. Concurrently, Sandia instrumented the model with over 1500 transducers to measure strain, displacement and forces in the model from prestressing through the pressure testing. The limit state test of the PCCV model was conducted in September, 2000 at Sandia National Laboratories. This paper describes the conduct and some of the results of this test.

Risk-Informed Pressurized Water Reactor Vessel Inspection Interval Extension

2004 ◽  
Author(s):  
Cheryl L. Boggess ◽  
Bruce A. Bishop ◽  
Nathan A. Palm ◽  
Owen F. Hedden
Keyword(s):  
Pressure Vessel ◽  
Nuclear Regulatory Commission ◽  
Reactor Vessel ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Pilot Studies ◽  
Water Reactor ◽  
Insignificant Change ◽  
Inspection Interval ◽  
Interval Extension

The methodology discussed in this paper provides a risk informed basis for decreasing the frequency of inspection for the Pressurized Water Reactor (PWR) reactor pressure vessel (RPV). The decrease in frequency is based on extending the interval between inspections from the current interval of 10 years to 20 years. Results of pilot studies on typical designs of PWR vessels show that the change in risk associated with extending the inspection interval by more than 10 years is within the guidelines specified in U.S. Regulatory Guide 1.174 for insignificant change in risk. The current requirements for inspection of reactor vessel pressure-containing welds have been in effect since the 1989 Edition of American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section XI, supplemented by U.S. Nuclear Regulatory Commission (NRC) Regulatory Guide 1.150, June 1981. The manner in which these examinations are conducted has recently been augmented by Appendix VIII of Section XI, 1996 Addenda, as implemented by the NRC in amendment to 10CFR50.55a effective November 22, 1999. This paper summarizes the insignificant change in risk results for the PWR pilot-plant studies, including the effects of fatigue crack growth and in-service inspection of postulated surface-breaking flaws. These results demonstrate that the proposed RPV inspection interval extension is a viable option for the industry.

Evaluation of a SGTR Accident in the Multi-Application Integrated Pressurized Water Reactor

2010 ◽  
Author(s):  
Liguo Jiang ◽  
Minjun Peng ◽  
Jiange Liu
Keyword(s):  
Steam Generator ◽  
Flow Model ◽  
Heat Removal ◽  
Injection System ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Heat Removal System ◽  
Removal System ◽  
Residual Heat

One of more frequent events in the Pressurized Water Reactor (PWR) is Steam Generator Tube Rupture (SGTR) accident, which is among the main accidents in the field of nuclear safety. This paper studies the SGTR event in the Multi-application Integrated Pressurized Water Reactor (IPWR) using the best-estimate thermal-hydraulic code RELAP5/MOD3.4. In the reactor of IPWR, several Once-Through Steam Generator (OTSG) cassettes are used and located between the core support and the pressure vessel. The tube rupture location is on the top of the tube sheet of a steam generator. Three different tube rupture modeling methods and several different subcooled discharge coefficients in the critical flow model are considered and compared. In the safety analysis, high pressure safety injection system, core makeup system and Passive Residual Heat Removal System (PRHRS) that would affect the accident consequences are considered.

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