Application of the performance-goal based approach for establishing the SSE site specific response spectrum for new nuclear power plants in South Africa

The U.S. Nuclear Regulatory Commission (NRC) established a new process for licensing nuclear power plants under Title 10 of the Code of Federal Regulations (10 CFR) Part 52, “Licenses, Certifications, and Approvals for Nuclear Power Plants,” which provides requirements for early site permits (ESPs), standard design certifications (DCs), and combined license (COL) applications. In this process, an application for a COL may incorporate by reference a DC, an ESP, both, or neither. This approach allows for early resolution of safety and environmental issues. The COL review will not reconsider the safety issues resolved by the DC and ESP processes. However, a COL application that incorporates a DC by reference needs to demonstrate that pertinent site-specific parameters are confined within the safety envelopes established by the DC. This paper provides an overview of site parameters related to seismic designs and associated seismic issues encountered in DC and COL application reviews using the 10 CFR Part 52 process. Since DCs treat the seismic design and analysis of nuclear power plant (NPP) structures, systems, and components (SSC) as bounding to future potential sites, the design ground motions and associated site parameters are often conservatively specified, representing envelopes of site-specific seismic hazards and parameters. For a COL applicant to incorporate a DC by reference, it needs to demonstrate that the site-specific hazard in terms of ground motion response spectra (GMRS) is enveloped by the certified design response spectra of the DC. It also needs to demonstrate that the site-specific seismic parameters, such as foundation-bearing capacities, soil profiles, and the like, are confined within the site parameter envelopes established by the DC. For the noncertified portion of the plant SSCs, the COL applicant should perform the seismic design and analysis with respect to the site-specific GMRS and associated site parameters. This paper discusses the seismic issues encountered in the safety reviews of DC and COL applications. Practical issues dealing with comparing site-specific features to the standard designs and lessons learned are also discussed.

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Overview of Implementation of Probabilistic-Risk-Assessment-Based Seismic Margin Analyses in Design Certification and Combined License Applications

ASME 2010 Pressure Vessels and Piping Conference: Volume 8 ◽

10.1115/pvp2010-25160 ◽

2010 ◽

Author(s):

Jim Xu

Keyword(s):

Risk Assessment ◽

Nuclear Power ◽

Power Plants ◽

Nuclear Power Plants ◽

Probabilistic Risk Assessment ◽

The United States ◽

Severe Accident ◽

Site Specific ◽

Standard Design ◽

Probabilistic Risk

The licensing process for new reactors in the United States was established in accordance with Title 10 of the Code of Federal Regulations (10 CFR) Part 52, “Licenses, Certifications, and Approvals for Nuclear Power Plants,” which provides requirements for early site permit (ESP), standard design certification (DC), and combined license (COL) applications. In this process, an application for a COL may incorporate by reference a DC, an ESP, both, or neither. This approach allows for the early resolution of safety and environmental issues. The safety issues resolved by the DC and ESP processes are not reconsidered during a COL review. However, a COL application that incorporates a DC by reference must demonstrate that pertinent site-specific characteristics are confined within the envelopes established by the DC’s site parameters. This paper provides an overview of the implementation of probabilistic risk assessment (PRA) based seismic margin analyses in DC and COL applications. In addressing the severe accident preventions and mitigations for new reactors, 10 CFR 52.47(a)(27) requires that the final safety analysis report for a DC application describe the design-specific PRA and its results. Regulatory Guide 1.206, “Combined License Applications for Nuclear Power Plants (LWR Edition),” issued June 2007, further states that the scope of this assessment should be a Level 1 and Level 2 PRA that includes internal and external hazards and addresses all plant operating modes. However, the staff recognized that it is not practical for a DC applicant to perform a seismic PRA because a DC application would not contain site-specific seismic hazard information. As an alternative approach to a seismic PRA, the staff proposed a PRA-based seismic margin analysis in SECY-93-087, “Policy, Technical, and Licensing Issues Pertaining to Evolutionary and Advanced Light-Water Reactor (ALWR) Designs,” dated April 2, 1993, and the Commission approved it in the corresponding staff requirements memorandum, dated July 21, 1993. This analysis preserves key elements of a seismic PRA to the maximum extent possible and estimates the design-specific plant seismic capacity in terms of sequence-level high confidence of low probability of failure capacities and fragility for all sequences leading to core damage or containment failures up to approximately 1.67 times the ground motion acceleration of the design-basis safe-shutdown earthquake. Using this approach, the analysis can demonstrate acceptably low seismic risk for a DC. This paper discusses the implementation aspects of PRA-based seismic margin analyses in support of a DC application and post-DC updating activities, including COL updates to incorporate site- and plant-specific features and post-COL verifications.

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Study on Nuclear Power Plant: A Recognition Category Emergency Action Level

18th International Conference on Nuclear Engineering: Volume 3 ◽

10.1115/icone18-29032 ◽

2010 ◽

Author(s):

Tao Liu

Keyword(s):

Power Plant ◽

Nuclear Power Plant ◽

Nuclear Power ◽

Power Plants ◽

Nuclear Power Plants ◽

Site Specific ◽

Consistent Basis ◽

Level Systematic

Emergency Action Level (EAL) is an effective basis and criteria for nuclear power plant emergency classification which is a pre-determined, site specific, observable threshold for a plant initiating condition that places the plant in a given emergency classification level. Systematic approaches have built a set of generic EAL guidelines, together with the basis for each, such that they could be used and adapted by each utility on a consistent basis. EALs information is presented by Recognition Categories (RC), in which “A” Recognition Category refers to abnormal Rad levels/Radiological effluent. The methodology of EALs development is reviewed first and the technical bases for “A” Recognition Category according to classic EAL methodology, i.e. NEI 99-01 series and IAEA-TECDOC-955 are introduced. The statue of Chinese nuclear power plants’ “A” Recognition Category EALs development is summarized after that, which intends to raise some challenging questions for classic “A” Recognition Category EALs development. One opinion is that some “A” Recognition Category EALs in NEI 99-01 series are not conservative. To explain the reason, the discussion mainly focuses on the interpretation and comparison of NEI 99-01 and IAEA-TECDOC-955 technical bases. Finally, it is recommended to select one suitable guideline as reference when developing plant specific EALs.

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Research on Improved Seismic Instrumentation System for Nuclear Power Plants

Energies ◽

10.3390/en14144262 ◽

2021 ◽

Vol 14 (14) ◽

pp. 4262

Author(s):

Liang Li ◽

Xiuli Du ◽

Rong Pan ◽

Xiuyun Zhu ◽

Haiyan Luan

Keyword(s):

Nuclear Power ◽

Power Plants ◽

Nuclear Power Plants ◽

Response Spectrum ◽

Ground Acceleration ◽

Cumulative Absolute Velocity ◽

Safety Regulations ◽

Seismic Instrumentation ◽

Three Components

According to the requirements of nuclear safety regulations, nuclear power plants must be equipped with seismic instrumentation systems, which are mainly used for monitoring alarm and automatic shutdown alarm during an earthquake. Both the second and third generation NPPs adopt Peak Ground Acceleration (PGA). However, among the seismic acceleration characteristics, isolated and prominent single high frequency acceleration peaks have no decisive influence on the seismic response. Especially when the earthquake monitoring alarm is at 1 out of 7, it is likely to cause a false alarm or false shutdown. In addition, it usually takes one month or more for the NPPs to restart after the shutdown. In this paper, an improved seismic instrumentation system based on the existing system is proposed. For high intensity areas, three components resultant acceleration is used to judge the 2 out of 4 logic of the automatic seismic trip system(ASTS). For low intensity areas, the seismic failure level is evaluated quickly by using three components resultant acceleration, seismic instrument intensity, cumulative absolute velocity, floor response spectrum and other multi-parameters, avoiding unnecessary and long-term shutdown inspection.

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Site-Specific Seismic Ground Response Study for Nuclear Power Plants and Ports in Mumbai

Natural Hazards Review ◽

10.1061/(asce)nh.1527-6996.0000177 ◽

2015 ◽

Vol 16 (4) ◽

pp. 04015002 ◽

Cited By ~ 25

Author(s):

Sarika S. Desai ◽

Deepankar Choudhury

Keyword(s):

Nuclear Power ◽

Power Plants ◽

Nuclear Power Plants ◽

Ground Response ◽

Site Specific ◽

Seismic Ground Response

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A Shaking Table Test for the Evaluation of Floor Response Spectrum of Seismic Isolated NPP Structure

Volume 8: Seismic Engineering ◽

10.1115/pvp2016-63728 ◽

2016 ◽

Cited By ~ 1

Author(s):

Min Kyu Kim ◽

Jung Han Kim

Keyword(s):

Nuclear Power ◽

Power Plants ◽

Nuclear Power Plants ◽

Shaking Table Test ◽

Response Spectrum ◽

Low Frequency ◽

Shaking Table ◽

Frame Structure ◽

Design Spectrum ◽

Seismic Input

The development of a floor response spectrum (FRS) is very important for a seismic risk assessment of nuclear power plants. In the case of non-isolated nuclear power plants, the methodology regarding FRS generation has already been developed. But in the case of seismic isolated NPP structure, the methodology for developing floor response spectrum is not been determined yet. Therefore, in this study, shaking table tests for a seismic isolated frame structure were performed for the development of the floor response spectrum. For the shaking table test, a two-story artificial frame structure was manufactured. For the isolation devices, a lead rubber bearing and an EradiQuake system (EQS) were used. An artificial input seismic motion, which was generated for the NRC Reg. guide 1.60 design spectrum, was used but low-frequency range should be cut off for a decrease of the shaking table displacement. One-, two-, and three-dimensional seismic input motions were considered for the assessment of the horizontal bidirectional and vertical directional effects. Through this test, whether a horizontal bi-directional seismic input can make a difference in the floor response, and moreover, whether the vertical input motion can make a change in the horizontal floor response, were investigated.

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