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.

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.

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.

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.

Inspection Optimization Justification for PWR Main Steam and Feedwater Nozzles Using Probabilistic Fracture Mechanics

2019 ◽  
Author(s):  
C. Lohse ◽  
D. J. Shim ◽  
D. Somasundaram ◽  
R. Grizzi ◽  
G. L. Stevens ◽  
...  
Keyword(s):  
Fracture Mechanics ◽  
Nuclear Regulatory Commission ◽  
Pressure Vessels ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Asme Code ◽  
Sensitivity Studies ◽  
Main Steam ◽  
Regulatory Commission ◽  
Technical Basis

Abstract Pressurized water reactor (PWR) steam generator (SG) main steam and feedwater nozzles are classified as ASME Code, Section XI, Class 2, Category C-B, pressure retaining welds in pressure vessels. Current ASME Code requirements specify that the nozzle-to-shell welds (Item No. C2.21 & C2.32) and nozzle inner radius sections (Item C2.22) are to be examined very 10 years. An evaluation was performed to establish a technical basis for optimized inspection frequencies for these items. The work included a review of inspection history and results, a survey of components in the PWR fleet (which included both U.S. and overseas plants), selection of representative main steam and feedwater nozzle configurations and operating transients for stress analysis, evaluation of potential degradation mechanisms, and flaw tolerance evaluations consisting of probabilistic and deterministic fracture mechanics analyses. The results of multiple inspection scenarios and sensitivity studies were compared to the U.S. Nuclear Regulatory Commission (NRC) safety goal of 10−6 failures per year.

SAMA Analysis in Support of License Renewal Application for the Fort Calhoun Station

2004 ◽  
Author(s):  
Raymond E. Schneider ◽  
Harold A. Hackerott ◽  
Mathew C. Jacob
Keyword(s):  
Power Level ◽  
Cost Effective ◽  
Nuclear Regulatory Commission ◽  
Severe Accident ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Public Power ◽  
Regulatory Commission ◽  
License Renewal ◽  
Station Unit

The Omaha Public Power District (OPPD) owns and operates the Fort Calhoun Station Unit 1 (FCS), a single unit Pressurized Water Reactor (PWR) plant. This plant was licensed by the U. S. Nuclear Regulatory Commission (NRC) to operate at a core power level of 1500 MWt for a period of 40 years. In 2002 OPPD submitted a license renewal application to the NRC and recently, this application for a license renewal for a period of 20 more years was approved by the NRC. In order to secure the license renewal, the utility evaluated Severe Accident Mitigation Alternatives (SAMAs) and identified potential risk significant and cost-beneficial plant changes. This process consisted of a review of past industry changes obtained via past SAMA analyses and advanced reactor design efforts as well as a review of FCS PSA insights. While most of the industry-based SAMAs were screened out, the internally focused changes were mostly found to be risk significant and cost effective. This paper presents the SAMA analysis performed in support of the FCS license renewal application and discusses the insights from this analysis.

The Health of Risk-Informed Regulation

2008 ◽  
Author(s):  
Robert J. Lutz ◽  
Raymond Schneider
Keyword(s):  
Fire Risk ◽  
Nuclear Regulatory Commission ◽  
Bright Spot ◽  
Regulatory Change ◽  
Pressurized Water Reactor ◽  
Regulatory Oversight ◽  
Pressurized Water ◽  
Technical Specifications ◽  
Regulatory Commission

Ten years ago, risk-informed regulation was proclaimed by many to be the future of regulatory activity. The three options for using risk in the regulatory arena had been defined in SECY-98-300 and efforts were initiated to develop the first regulatory changes. However, regulatory change has been slow. The revised 50.44 rule, which was supposed to be non-controversial, was not finalized until September of 2003, five years after it was started. The new 50.69 rule, which embodied the STP graded quality assurance principles, was not finalized until November of 2004 and the Pressurized Water Reactor Owners Group (PWROG) pilot plant application is still under review by the Nuclear Regulatory Commission (NRC). The 50.46a rule change, admittedly a very difficult and controversial undertaking, is still not finalized. On the other hand, the NRC has increased the use of plant risk insights and results in the Regulatory Oversight Process, particularly through the Significance Determination Process (SDP) and the Mitigating System Performance Index (MSPI). While the industry has been working on the major regulatory initiatives described above, they have also successfully initiated a significant number of risk-informed programs within the existing regulations. Significant among these industry efforts are the risk-informed changes to the Technical Specifications and the risk-informed changes to in-service inspection and in-service testing. One bright spot on the horizon is the use of NFPA-805, which uses fire risk insights, as an alternative to the Appendix R requirements for fire protection. The industry and the NRC are working together to develop an acceptable methodology for implementing this alternative. The increasing requirements for scope and quality of the probabilistic risk assessments (PRAs) that are used as the basis for many risk-informed activities has resulted in a major slow-down in licensee implementation of new risk-informed activities. The publication of industry consensus standards is resulting in significant resource expenditures to upgrade the scope and quality of the PRAs. Once the scope and quality of the plant-specific PRAs is completed, there should again be significant implementation of risk-informed applications. In summary, risk-informed regulation has already produced significant increases in safety and has potentially reduced licensee regulatory burdens. However, the promise of risk-informed changes to regulations is still an unfulfilled expectation and will likely remain so in the near future. Once the scope and quality of PRAs is upgraded to satisfy the industry consensus standards, there should be renewed activity in implementing risk-informed programs.

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.

scholarly journals Analysis of the Technological Evolution of Materials Requirements Included in Reactor Pressure Vessel Manufacturing Codes

2021 ◽  
Vol 13 (10) ◽  
pp. 5498
Author(s):  
Alvaro Rodríguez-Prieto ◽  
Mariaenrica Frigione ◽  
John Kickhofel ◽  
Ana M. Camacho
Keyword(s):  
Sustainable Development ◽  
Pressure Vessel ◽  
Nuclear Regulatory Commission ◽  
Green Energy ◽  
Reactor Pressure Vessel ◽  
Technical Literature ◽  
Pressurized Water ◽  
Water Reactors ◽  
Regulatory Commission ◽  
Reactor Pressure

The growth of green energy technologies within the frame of the 7th Sustainable Development Goal (SDG) along with the concern about climatic changes make nuclear energy an attractive choice for many countries to ensure energy security and sustainable development as well as to actively address environmental issues. Unlike nuclear equipment (immovable goods), which are often well-catalogued and analyzed, the design and manufacturing codes and their standardized materials specifications can be considered movable and intangible goods that have not been thoroughly studied based on a detailed evaluation of the scientific and technical literature on the reactor pressure vessel (RPV) materials behavior. The aim of this work is the analysis of historical advances in materials properties research and associated standardized design codes requirements. The analysis, based on the consolidated U.S. Nuclear Regulatory Commission (NRC) Regulatory Guide (RG) 1.99 Rev.2 model, enables determination of the best materials options, corresponding to some of the most widely used material specifications such as WWER 15Kh2MFAA (used from the 1970s and 1980s; already in operation), ASME SA-533 Grade B Cl.1 (used in pressurized water reactor-PWR 2nd–4th; already in operation), DIN 20MnMoNi55 and DIN 22NiMoCr37 (used in PWR 2nd–4th) as well as ASTM A-336 Grade F22V (current designs). Consequently, in view of the results obtained, it can be concluded that the best options correspond to recently developed or well-established specifications used in the design of pressurized water reactors. These assessments endorse the fact that nuclear technology is continually improving, with safety being its fundamental pillar. In the future, further research related to the technical heritage from the evolution of materials requirements for other clean and sustainable power generation technologies will be performed.

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