RA-19-0452, Seismic Probabilistic Risk Assessment (Spra), Response to March 12, 2012, Request for Information Regarding Recommendation 2.1: Seismic, of the Near-Term Task Force Related to the Fukushima Dai-ichi Accident
| ML20084P290 | |
| Person / Time | |
|---|---|
| Site: | Robinson  |
| Issue date: | 12/12/2019 |
| From: | Kapopoulos E Duke Energy Corp |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| Valentin-Olmeda M | |
| References | |
| RA-19-0452 | |
| Download: ML20084P290 (202) | |
Text
Security-Related Information - Withhold Under 10 CFR 2.390 DUKE Ernest J. Kapopoulos, Jr.
r.. H. B. Robinson Steam
- ~ ENERGY Electric Plant Unit 2 Site Vice President Duke Energy 3581 West Entrance Road Hartsville, SC 29550 O: 843 951 1701 F: 843 951 1319 Ernie.Kapopoulos@duke-energy.com Serial: RA-19-0452 10 CFR 50.4 December 12, 2019 10 CFR 50.54(f)
ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 DOCKET NO. 50-261 / RENEWED LICENSE NO. DPR-23
SUBJECT:
H.B. Robinson Steam Electric Plant, Unit No. 2 - Seismic Probabilistic Risk Assessment (SPRA), Response to March 12, 2012, Request for Information Regarding Recommendation 2.1: Seismic, of the Near-Term Task Force Related to the Fukushima Dai-ichi Nuclear Power Plant Accident
REFERENCES:
- 1. NRC Letter, Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendations 2.1, 2.3, and 9.3 of the Near-Term Task Force review of Insights from the Fukushima Dai-ichi Accident, dated March 12, 2012 (ADAMS Accession No. ML12053A340)
- 2. EPRI 1025287, Seismic Evaluation Guidance: Screening, Prioritization and Implementation Details (SPID) for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1: Seismic, dated November 2012 (ML12333A170)
- 3. Duke Energy Letter, Submittal of Revision to Seismic Hazard Evaluation to Include New Ground Motion Response Spectra (GMRS) Using New Geotechnical Data and Shear-Wave Testing for H. B. Robinson Steam Electric Plant, Unit No. 2, dated July 17, 2015 (ML15201A006)
- 4. NRC Letter, H. B. Robinson Steam Electric Plant, Unit No. 2 - Staff Assessment of Information Provided Pursuant to Title 10 of the Code of Federal Regulations Part 50, Section 50.54(f), Seismic Hazard Reevaluations for Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident, dated October 19, 2015 (ML15280A199)
- 5. NRC Letter, Final Determination of Licensee Seismic Probabilistic Risk Assessments Under the Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(f) Regarding Recommendation 2.1: Seismic, of the Near-Term Task Force Review of Insights from the Fukushima Daiichi Accident, dated October 27, 2015 (ML15194A015)
Security-Related Information - Withhold Under 10 CFR 2.390
RA-19-0452 December 12, 2019 Page 2
- 6. Duke Energy Letter, Request for Extension of Due Date for Seismic Probabilistic Risk Assessment Submittal, for H.B. Robinson Steam Electric Plant, Unit No. 2, dated November 29, 2018 (ML18337A159)
- 7. NRC Letter, H. B. Robinson Steam Electric Plant, Unit No. 2 - Response to Request for Extension of Seismic Probabilistic Risk Assessment Submittal (EPID No. L-2018-JLD-0017), dated January 10, 2019 (ML19004A356)
- 8. Duke Energy Letter, Request for Extension of Due Date for Seismic Probabilistic Risk Assessment Submittal, for H.B. Robinson Steam Electric Plant, Unit No. 2, dated October 21, 2019 (ML19294A028)
- 9. NRC Letter, H. B. Robinson Steam Electric Plant, Unit No. 2 - Response to Request for Extension of Due Date for Seismic Probabilistic Risk Assessment Submittal (EPID NO.
L-2019-JLD-0014), dated October 28, 2019 (ML19296C623)
Ladies and Gentlemen, On March 12, 2012, the U.S. Nuclear Regulatory Commission (NRC) issued a request for information pursuant to 10 CFR 50.54(f) associated with the recommendations of the Fukushima Near-Term Task Force (NTTF) (i.e., Reference 1), requesting each licensee to reevaluate the seismic hazards at their sites using present-day NRC requirements and guidance, and to identify actions taken or planned to address plant-specific vulnerabilities associated with the updated seismic hazards.
Industry guidance was developed by EPRI that provided the screening, prioritization and implementation details (SPID) for the resolution of Fukushima Near-Term Task Force Recommendation 2.1: Seismic. The SPID (i.e., Reference 2) was used to compare the reevaluated seismic hazard to the design basis hazard. The H.B. Robinson Steam Electric Plant (HBRSEP) Unit No. 2, reevaluated seismic hazard (i.e., Reference 3) concluded that the ground motion response spectrum (GMRS) exceeded the design basis seismic response spectrum in the 1 to 10 Hz range, and therefore a seismic probabilistic risk assessment was required.
Reference 4 contains the NRC Staff Assessment of the HBRSEP Unit No. 2 reevaluated seismic hazard submittal and confirmed the conclusion that the GMRS for the Robinson site exceeds the design basis seismic response spectrum and a seismic risk evaluation is merited.
Reference 5 contains the NRC letter for the final determination of licensee seismic probabilistic risk assessments. In that letter, the NRC instructed HBRSEP Unit No. 2, to submit an SPRA by March 31, 2019.
Duke Energy requested an extension to that due date (i.e., Reference 6), and the NRC approved the due date extension to October 31, 2019 (i.e., Reference 7). Duke Energy requested a second extension request to the SPRA submittal due date (i.e., Reference 8), and the NRC approved the due date extension to December 12, 2019 (i.e., Reference 9). of this letter contains the HBRSEP, Unit No. 2, Seismic Probabilistic Risk Assessment (SPRA) Summary Report which provides the information requested in Enclosure 1, Item (8) B. of Reference 1.
This letter contains no new Regulatory Commitments.
RA-19-0452 December 12, 2019 Page 3 Should you have any questions regarding this submittal, please contact Art Zaremba - Director, Fleet Licensing, at (980) 373-2062.
I declare under penalty of perjury that the foregoing is true and correct. Executed on December 12, 2019.
Sincerely, Ernest J. Kapopoulos, Jr.
Site Vice President LJG/ljg
Enclosure:
H.B. Robinson Steam Electric Plant, Unit No. 2, Seismic Probabilistic Risk Assessment (SPRA) in Response to 50.54(f) Letter Regarding NTTF 2.1:
Seismic Summary Report cc (with enclosure)
L. Dudes, Regional Administrator USNRC Region II A Hon, NRR Project Manager - RNP M. Fannon, NRC Senior Resident Inspector - RNP
Enclosure H.B. Robinson Steam Electric Plant, Unit No. 2, Seismic Probabilistic Risk Assessment (SPRA) in Response to 50.54(f) Letter Regarding NTTF 2.1: Seismic Summary Report
H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 SEISMIC PROBABILISTIC RISK ASSESSMENT IN RESPONSE TO 50.54(F) LETTER REGARDING NTTF 2.1 SEISMIC
SUMMARY
REPORT December 2019 Page 1 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table of Contents Executive Summary ........................................................................................................ 4 1.0 Purpose and Objective .................................................................................. 5 2.0 Information Provided in This Report .............................................................. 6 3.0 Robinson Nuclear Plant Seismic Hazard and Plant Response.................... 10 3.1 Seismic Hazard Analysis ............................................................................. 11 4.0 Determination of Seismic Fragilities for the SPRA ...................................... 32 4.1 Seismic Equipment List ............................................................................... 32 4.2 Walkdown Approach ................................................................................... 35 4.3 Dynamic Analysis of Structures................................................................... 37 4.4 SSC Fragility Analysis ................................................................................. 40 5.0 Plant Seismic Logic Model .......................................................................... 43 5.1 Development of the SPRA Plant Seismic Logic Model ............................... 43 5.2 SPRA Plant Seismic Logic Model Technical Adequacy .............................. 48 5.3 Seismic Risk Quantification ......................................................................... 48 5.4 SCDF Results ............................................................................................. 49 5.5 SLERF Results............................................................................................ 62 5.6 SPRA Quantification Uncertainty Analysis .................................................. 77 5.7 SPRA Quantification Sensitivity Analysis .................................................... 79 5.8 SPRA Logic Model and Quantification Technical Adequacy ....................... 82 6.0 Conclusions ................................................................................................. 83 7.0 References .................................................................................................. 84 8.0 Acronyms ................................................................................................... 90 Appendix A .................................................................................................................... 92 A.1. Overview of Peer Review ............................................................................ 92 A.2. Summary of the Peer Review Process ........................................................ 92 A.3. Peer Review Team Qualifications ................................................................ 93 A.4. Summary of the Peer Review Conclusions .................................................. 95 A.5. Summary of the Assessment of Supporting Requirements and Findings .... 99 A.6. Summary of Technical Adequacy of the SPRA for the 50.54(f) Response 102 Page 2 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT A.7. Summary of SPRA Capability Relative to SPID Tables 6-4 through 6-6.... 102 A.8. Identification of Key Assumptions and Uncertainties Relevant to the SPRA Resultss\Results. ............................................................................ 105 A.9. Identification of Plant Changes Not Reflected in the SPRA ....................... 107 Page 3 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Executive Summary In response to the 10 CFR 50.54(f) letter issued by the NRC on March 12, 2012, a seismic PRA (SPRA) has been developed to perform the seismic risk assessment for H. B. Robinson Steam Electric Plant, Unit No. 2. The SPRA model shows the point estimate seismic Core Damage Frequency (SCDF) is 9.27x10-5/reactor-year and the seismic Large Early Release Frequency (SLERF) is 2.02x10-5/reactor-year. The SPRA reflects the as-built/as-operated Robinson Nuclear Power Plant as of the freeze date for the internal events model (June 2015). An assessment is included in Appendix A of the impact of the results of plant changes not included in the model since the model freeze date.
Due to the insights gained from the seismic risk assessment, Robinson plans to implement a means to provide Auxiliary Feedwater supplied by a modified FLEX strategy. The results of the corresponding sensitivity analysis show that the SCDF and SLERF can be reduced by approximately 40 percent and 30 percent, respectively. This modification will be implemented by the end of 2022.
Page 4 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT 1.0 Purpose and Objective Following the accident at the Fukushima Dai-ichi nuclear power plant resulting from the March 11, 2011, Great Tohoku Earthquake and subsequent tsunami, the Nuclear Regulatory Commission (NRC) established a Near Term Task Force (NTTF) to conduct a systematic review of NRC processes and regulations and to determine if the agency should make additional improvements to its regulatory system. The NTTF developed a set of recommendations intended to clarify and strengthen the regulatory framework for protection against natural phenomena. Subsequently, the NRC issued a 50.54(f) letter on March 12, 2012 [1], requesting information to assure that these recommendations are addressed by all U.S. nuclear power plants. The 50.54(f) letter requests that licensees and holders of construction permits under 10 CFR Part 50 reevaluate the seismic hazards at their sites against present-day NRC requirements and guidance.
A comparison between the reevaluated seismic hazard and the design basis for the Robinson Nuclear Power Plant has been performed, in accordance with the guidance in EPRI 1025287, Screening, Prioritization and Implementation Details (SPID) for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1: Seismic [2], and previously submitted to NRC [3]. That comparison concluded that the ground motion response spectra (GMRS), which was developed based on the reevaluated seismic hazard, exceeds the design basis seismic response spectrum in the 1 to 10 Hz range, and a seismic risk assessment is required. An SPRA has been developed to perform the seismic risk assessment for the Robinson Nuclear Power Plant in response to the 50.54(f) letter, specifically item (8) in Enclosure 1 of the 50.54(f) letter.
This report describes the seismic PRA developed for the Robinson Nuclear Power Plant and provides the information requested in item (8)(B) of Enclosure 1 of the 50.54(f) letter and in Section 6.8 of the SPID. The SPRA model has been peer reviewed (as described in Appendix A) and found to be of appropriate scope and technical capability for use in assessing the seismic risk for the Robinson Nuclear Power Plant, identifying which structures, systems, and components (SSCs) are important to seismic risk, and describing plant-specific seismic issues and associated actions planned or taken in response to the 50.54(f) letter.
This report provides summary information regarding the SPRA as outlined in Section 2.
The level of detail provided in the report is intended to enable NRC to understand the inputs and methods used, the evaluations performed, and the decisions made as a result of the insights gained from the Robinson Nuclear Power Plant seismic PRA.
Page 5 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT 2.0 Information Provided in This Report The following information is requested in the 50.54(f) letter [1], Enclosure 1, Requested Information Section, paragraph (8)B, for plants performing a SPRA.
(1) The list of the significant contributors to SCDF for each seismic acceleration bin, including importance measures (e.g., Risk Achievement Worth, Fussel-Vesely and Birnbaum)
(2) A summary of the methodologies used to estimate the SCDF and LERF, including the following:
- i. Methodologies used to quantify the seismic fragilities of SSCs, together with key assumptions ii. SSC fragility values with reference to the method of seismic qualification, the dominant failure mode(s), and the source of information iii. Seismic fragility parameters iv. Important findings from plant walkdowns and any corrective actions taken
- v. Process used in the seismic plant response analysis and quantification, including the specific adaptations made in the internal events PRA model to produce the seismic PRA model and their motivation vi. Assumptions about containment performance (3) Description of the process used to ensure that the SPRA is technically adequate, including the dates and findings of any peer reviews (4) Identified plant-specific vulnerabilities and actions that are planned or taken Note that 50.54(f) letter Enclosure 1 paragraphs 1 through 6, regarding the seismic hazard evaluation reporting, also apply, but have been satisfied through the previously submitted Robinson Nuclear Power Plant Seismic Hazard Submittal [4]. Further, 50.54(f) letter Enclosure 1 paragraph 9 requests information on the Spent Fuel Pool. Duke submitted the Spent Fuel Pool Supplemental Report to the NRC for H. B. Robinson Nuclear Power Plant [57] and has received the final staff assessment [60].
Table 2-1 provides a cross-reference between the 50.54(f) reporting items noted above and the location in this report where the corresponding information is discussed.
The SPID [2] defines the principal parts of an SPRA, and the H. B. Robinson Nuclear Power Plant SPRA has been developed and documented in accordance with the SPID.
The main elements of the SPRA performed for H. B. Robinson Nuclear Power Plant in response to the 50.54(f) Seismic letter correspond to those described in Section 6.1.1 of the SPID, i.e.:
- Seismic hazard analysis
- Seismic structure response and SSC fragility analysis
- Systems/accident sequence (seismic plant response) analysis
- Risk quantification Table 2-2 provides a cross-reference between the reporting items noted in Section 6.8 of the SPID, other than those already listed in Table 2-1, and provides the location in this report where the corresponding information is discussed.
The Robinson Nuclear Power Plant SPRA and associated documentation has been peer reviewed against the PRA Standard in accordance with the process defined in NEI 12-13
[6] as documented in the Robinson Nuclear Power Plant SPRA Peer Review Report. The Page 6 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Robinson Nuclear Power Plant SPRA, complete SPRA documentation, and details of the peer review are available for NRC review.
This submittal provides a summary of the SPRA development, results and insights, and the peer review process and results, sufficient to meet the 50.54(f) information request in a manner intended to enable NRC to understand and determine the validity of key input data and calculation models used, and to assess the sensitivity of the results to key aspects of the analysis.
The content of this report is organized as follows:
- Section 3 provides information related to the Robinson Nuclear Power Plant seismic hazard analysis.
- Section 4 provides information related to the determination of seismic fragilities for the Robinson Nuclear Power Plant SSCs included in the seismic plant response.
- Section 5 provides information regarding the plant seismic response model (seismic accident sequence model) and the quantification of results.
- Section 6 summarizes the results and conclusions of the SPRA, including identified plant seismic issues and actions taken or planned.
- Section 7 provides references.
- Section 8 provides a list of acronyms used.
Appendix A provides an assessment of SPRA Technical Adequacy for Response to NTTF 2.1 Seismic 50.54(f) Letter, including a summary of the Robinson Nuclear Power Plant SPRA peer review.
Page 7 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table 2-1 Cross-Reference for 50.54(f) Enclosure 1 SPRA Reporting 50.54(f) Letter Reporting Item Description Location in this Report 1 List of the significant contributors to Section 5 SCDF for each seismic acceleration bin, including importance measures 2 Summary of the methodologies Sections 3, 4, 5 used to estimate the SCDF and LERF 2i Methodologies used to quantify the Section 4 seismic fragilities of SSCs, together with key assumptions 2ii SSC fragility values with reference Tables 5.4-2 and 5.5-2 provide fragilities to the method of seismic (Am, r and u for fragilities following a log-qualification, the dominant failure normal distribution), failure mode mode(s), and the source of information, and method of determining information fragilities for the top risk significant SSCs based on Fussel-Vesely (F-V).
2iii Seismic fragility parameters Tables 5.4-2 and 5.5-2 provide fragilities information (Am, r and u for fragilities following a log-normal distribution) for the top risk significant SSCs based on Fussel-Vesely (F-V).
2iv Important findings from plant Section 4.2 addresses walkdowns and walkdowns and any corrective walkdown insights.
actions taken 2v Process used in the seismic plant Sections 5.1 and 5.2 provide this response analysis and information.
quantification, including specific adaptations made in the internal events PRA model to produce the seismic PRA model and their motivation 2vi Assumptions about containment Sections 4.3 and 5.5 address containment performance and related SSC performance.
3 Description of the process used to App. A describes the assessment of SPRA ensure that the SPRA is technically technical adequacy for the 50.54(f) adequate, including the dates and submittal and results of the SPRA peer findings of any peer reviews review.
4 Identified plant-specific Section 6 addresses this.
vulnerabilities and actions that are planned or taken Page 8 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table 2-2 Cross-Reference for Additional SPID Section 6.8 SPRA Reporting SPID Section 6.8 Item (1) Description Location in this Report A report should be submitted to the NRC summarizing the Entirety of the submittal addresses SPRA inputs, methods, and results. this.
The level of detail needed in the submittal should be Entirety of the submittal addresses sufficient to enable NRC to understand and determine the this. It identifies key methods of validity of all input data and calculation models used analysis and referenced codes and standards The level of detail needed in the submittal should be Entirety of the submittal addresses sufficient to assess the sensitivity of the results to all key this. Results sensitivities are aspects of the analysis discussed in the following sections:
- 5.7 (SPRA model sensitivities)
- 4.4 Fragility screening (sensitivity)
The level of detail needed in the submittal should be Entirety of the submittal template sufficient to make necessary regulatory decisions as a part addresses this.
of NTTF Phase 2 activities.
It is not necessary to submit all of the SPRA documentation Entire report addresses this. This for such an NRC review. Relevant documentation should report summarizes important be cited in the submittal, and be available for NRC review information from the SPRA, with in easily retrievable form. detailed information in lower tier documentation Documentation criteria for a SPRA are identified This is an expectation relative to throughout the ASME/ANS Standard [5 and 37]. Utilities documentation of the SPRA that the are expected to retain that documentation consistent with utility retains to support application of the Standard. the SPRA to risk-informed plant decision-making.
Note (1): The items listed here do not include those designated in SPID Section 6.8 as guidance.
Page 9 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT 3.0 Robinson Nuclear Plant Seismic Hazard and Plant Response Section 3.0 provides a high-level summary site description for the H. B. Robinson Steam Electric Plant (HBRSEP). The subsections provide brief summaries of the site hazard and response characterizations for the HBRSEP and the Lake Robinson Dam as well as discussions of the potential liquefaction evaluation and impacts.
The HBRSEP is a soil site located in Darlington County South Carolina. The following description of the general geology at the site is adapted from [16]:
The Robinson Plant is within the Coastal Plain physiographic province in South Carolina and within the upper portion of that province. At the western edge of the Coastal Plain, which is approximately 15 miles northwest of the site, pre-Cambrian basement rock of the Piedmont physiographic province is exposed. In the Piedmont, the basement rocks are covered with soil-like material weathered in place from the original granitic rocks, and the UFSAR indicates this weathered material may be present below the Coastal Plain formations at the site as well.
The Middendorf Formation of Cretaceous age overlies the Piedmont at the site.
The Middendorf Formation was formed by deposition of sediments transported by water from the west. A fluvial to deltaic depositional environment is described by Sohl and Owens, [17]. Both types of depositional environments are characterized by lateral and vertical variations in soil layers, both in composition and thickness.
Such variations were observed in the boring logs from historical and current explorations. Overall, the Middendorf is described as a sequence of alternating clay and sand layers. The sand layers vary from clean sands with some gravel zones to sands with varying proportions of silt and clay. The soils are generally hard or dense, but loose zones can exist. Indurated to partly indurated layers of clay and sand are common within the Middendorf Formation. A layer of hard clay approximately 15 to 30 feet thick is consistently present across the plant site.
A thin zone of recent soils caps the Middendorf Formation. The recent soils are sands of either alluvial, fluvial or aeolian deposition and vary in density and silt content both laterally and vertically. The boundary between the recent soils and the Middendorf Formation is not clearly identifiable and the recent soils are combined with the upper part of the Middendorf Formation for analysis based on similar shear wave velocities.
The current ground surface in the main plant area is at approximately elevation 226 feet.
Elevations in the seismic studies are referenced to the National Geodetic Vertical Datum of 1929 (NGVD29); the datum in effect when the plant was designed and constructed.
Slight amounts of cut and fill were needed to reach the current grade. The primary Category 1 structures (Reactor Building, Auxiliary Building, Turbine Generator Class 1 Building, Fuel Handling Building and New Fuel Building) are supported on driven pile foundations embedded into a hard clay layer within the Middendorf Formation. Pile tips for the reactor building are generally within the range of elevations 155 feet to 163 feet.
Other Category 1 structures are supported on shallow-depth foundations bearing in sands in the upper part of the Middendorf Formation. The intake structure is supported on a mat foundation bearing on the hard clay layer at approximately elevation 172 feet.
Detailed information regarding the HBRSEP site hazard was provided to NRC in the seismic information submitted to NRC in response to the NTTF 2.1 Seismic information request [4]. The response of [4] only considered the control point elevation in the main plant area (elevation 226 feet) for calculation of the ground motion response spectra Page 10 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT (GMRS). The probabilistic seismic hazard analysis (PSHA) described in this submittal expands on the response of [4] to consider four other control points for ground motion analyses:
- Elevation 216 feet for the base of the reactor containment building;
- Elevation 159.2 feet for the tip of the reactor building piles;
- Elevation 163 feet for the Lake Robinson Dam Spillway;
- Elevation 180 feet for the Lake Robinson dam original ground surface, and;
- Elevation 244 feet for the FLEX Building.
3.1 Seismic Hazard Analysis This section discusses the seismic hazard methodology, presents the final seismic hazard results used in the SPRA, and discusses important assumptions and important sources of uncertainty. The work follows the general guidance provided in Regulatory Guide 1.208
[25] and the SPID [2] but differs in some of the implementation details. These differences are discussed in appropriate sections of this report, and detailed information is available in [4].
The seismic hazard analysis determines the annual frequency of exceedance for selected ground motion parameters. The analysis involves use of earthquake source models, ground motion attenuation models, characterization of the site response (e.g. soil column),
and accounts for the uncertainties and randomness of these parameters to arrive at the site seismic hazard. Detailed information regarding the HBRSEP site hazard was provided to NRC in the seismic hazard information submitted to NRC in response to the NTTF 2.1 Seismic information request [4].
The analyses for the plant area used three alternative median shear wave velocity profiles (A, B and C) as described in [4] and listed in Table 3-1. The three profiles were weighted as described in [4] based on the relative amount of data available. A different set of alternative median shear wave velocity profiles was used for the Lake Robinson dam area, as shown in Table 3-1 and described in [16]; however, the methodology for using the profiles was the same as described for the plant area in [4].
As described in [4], the characterization of epistemic uncertainty in site median Vs used for HBRSEP differs somewhat from the approach described in Appendix B of [2] in which a best estimate median Vs profile is defined and epistemic uncertainty in median Vs is represented by upper and lower bound profiles with velocities assigned based on an epistemic sigma for ln(Vs).
However, the characterization using the three alternate profiles produces comparable epistemic uncertainty. As discussed in [4], Appendix B of [2] lists a recommended value for sigma ln(median Vs) of 0.35 for sites with limited Vs data and indicates that for sites with multiple detailed shear wave velocity profiles, the appropriate value may be significantly smaller. As there is a substantial amount of velocity data for the Robinson site, the modeled epistemic uncertainty in median Vs presented in [4] is consistent with the recommendations in [2].
Page 11 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table 3-1. HBRSEP Alternative Median Shear Wave Velocity Profiles and Included SSCs Alternative Median Shear Wave Velocity Profile Represented SSCs Label A
B All within the Plant Area C
1-Dam 2-Dam Lake Robinson Dam 3-Dam Control Point Elevations The site control point elevation is 226 feet (NGVD 29) as noted in [4]. This point is the ground surface elevation within the main plant area and defines the GMRS reference point.
The reactor building foundation is a mat supported by piles. The control point for the base of the mat is elevation 216 feet. A FIRS was developed for this elevation as a geological outcrop motion as described in [16].
The pile tip elevations for the reactor vary slightly; a control point to represent the pile tips was taken at elevation 159.2 feet using the approach described in [4]. A Soil Column Outcrop Response (SCOR) FIRS was developed for this elevation as described in [16].
The Auxiliary Building and New Fuel Building are pile-supported structures with the base of the pile cap at elevation 222 feet. The GMRS at the site control point elevation of 226 feet was used for these structures.
The Lake Robinson Dam spillway is a concrete structure founded at elevation 163 feet.
That point was taken as the control point elevation for the spillway. A Truncated Soil Column Response (TSCR) FIRS was developed as described in [16].
The Lake Robinson Dam embankment is not a Category 1 structure, but consideration of potential liquefaction impacts required that a FIRS at the elevation of the original ground (elevation 180 feet) be developed. Details of that FIRS development are presented in [16].
Ground Motion Parameters The GMRS and FIRS for the control point elevations noted above are computed using the weighted alternative profiles as described above for a range of spectral frequencies between 0.5 Hz and 100 Hz. Hazard curves for the SPRA (mean and fractiles of 0.05, 0.16, 0.5, 0.84 and 0.95) are provided for the various profiles and control points for peak ground acceleration (PGA, modeled as occurring at 100 Hz).
3.1.1 Seismic Hazard Analysis Methodology For the HBRSEP, the following method is used:
- Conduct a hard-rock PSHA for the site and perform deaggregation; Page 12 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT
- Develop alternative shear wave velocity profiles as discussed above to represent the plant site and the dam site;
- Develop site amplification functions for motions at the various control points using site response analyses;
- Combine the site amplification functions with the hard rock PSHA model to produce soil hazard curves for the various control points;
Each of the above steps is summarized below.
Hard Rock PSHA A hard-rock PSHA was performed for the HBRSEP site following the guidance of NRC Regulatory Guide 1.208 [25] and the SPID [2]. The PSHA used the following input:
- Seismic Source Model The seismic source model used for the HBRSEP PSHA is based on the Central and Eastern United States Seismic Source Characterization (CEUS-SSC) for Nuclear Facilities project [18]. Distributed seismic source zones and Repeated Large Magnitude Earthquake (RLME) sources within 625 miles (1,000 km) and those RLME seismic sources at greater distances that contribute at least 1 percent to the hazard at the site are included. Two refinements to the CEUS-SSC model are incorporated as discussed in [4]. These are:
Mmax distributions for seismotectonic zones IBEB, MID-C (A through D), PEZ-N, PEZ-W and SLR are updated as included in [19] and accepted by NRC [20]
for use at HBRSEP.
Revisions to the earthquake catalog in the south-eastern US including removal of reservoir-impounding causes and aftershocks of the Charleston earthquakes of 1886 [21]. Revised recurrence rates were subsequently generated for the south-eastern US and the overall result was included in [22].
- Ground Motion Characterization (GMC)
The GMC used for the HBESEP is the GMC model of [23] which provides Ground Motion Prediction Equations (GMPEs) for seven reference spectral frequencies: PGA (100 Hz), 25, 10, 5, 2.5, 1 and 0.5 Hz. These GMPEs have been endorsed by the NRC for use in computing hazard at nuclear sites [24].
- Seismicity Catalog The seismicity catalog used for the HBRSEP is based on the CEUS-SSC catalog provided by [18] with revisions as noted above. This catalog was current through the end of 2008 and includes all known earthquakes of a magnitude relevant for assessing earthquake hazard. The CEUS-SSC catalog was updated to include earthquakes through the end of November 2014 as described in [16]. Testing of the updated catalog confirmed the adequacy of the earthquake recurrence rates provided in [18] and [22].
Hard rock hazard calculations, including deaggregation, are performed according to [25].
A minimum moment magnitude of 5.0 is used in the calculations in lieu of a cumulative absolute velocity filter. Results of the PSHA and deaggregation are used to perform the site response analysis.
Page 13 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Site Response Analysis Site response analyses are performed using suites of input acceleration time histories representing the hard rock hazard at the site. The results of the site response analyses are used to develop soil amplification functions for the site control point (Elevation 226 ft.)
The amplification functions are then used to develop soil hazard curves for the control point from which horizontal UHRS and GMRS are developed. The same process is applied for the Dam Original Ground control point (Elevation 180 ft.).
For control points at Elevations 216 ft., 159.2 ft. and 163 ft., the amplification functions are developed from the full column site response analyses following guidance given in [26]
and [27]. Details are given in [16].
Inputs to the site response analysis include:
Vs profiles. The analyses for the plant area used three alternative median shear wave velocity profiles (A, B and C) as described in [4] and shown in Figure 3-1.
The three profiles were weighted as described in [4] based on the relative amount of data available.
As described in [4], the characterization of epistemic uncertainty in site median Vs used for HBRSEP differs somewhat from the approach described in Appendix B of [2] in which a best estimate median Vs profile is defined and epistemic uncertainty in median Vs is represented by upper and lower bound profiles with velocities assigned based on an epistemic sigma for ln(Vs).
However, the characterization using the three alternate profiles produces comparable epistemic uncertainty. As discussed in [4], Appendix B of [2] lists a recommended value for sigma ln(median Vs) of 0.35 for sites with limited Vs data and indicates that for sites with multiple detailed shear wave velocity profiles, the appropriate value may be significantly smaller. As there is a substantial amount of velocity data for the Robinson site, the modeled epistemic uncertainty in median Vs presented in [4] is consistent with the recommendations in [2].
The depth to bedrock (approximately 410 feet below site ground surface) is based on results of a deep boring performed for the PSHA study as well as historical site data and geologic publications [16].
Shear modulus and damping curves. Alternative sets of non-linear material properties (G/Gmax and damping curves) were developed for the soil and weathered rock at HBRSEP using the sediment characteristics as described in
[28]. The alternative G/Gmax and damping relationships represent epistemic uncertainty in the dynamic behavior of the subsurface materials. Following the approach described in [2] for treating epistemic uncertainty in G/Gmax and damping, the G/Gmax and damping relationships developed in [28] were grouped into two sets: a Sand set representing a greater degree of nonlinearity and a Clay set representing a lesser degree of nonlinearity. This is consistent with the use by [2] of the guidance in [29] and the more linear Peninsular Ranges subset of [29] described in [2].
Page 14 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT 250
' - . - . - . - . -Ground
. . . -~Surface (Elev.226')
ALLUVIUM AND UPPER 200 MIDDENDORF FORMATION:
150 MIDDENDORF FORMATION: Hard Clay
.- *-*-*- *- *- -*-*-*-*-.
- i-. - * ~.
MIDDENDORF FORMATION:
100 Alternating layers of clay, clean sand, silty sand and clayey sand, occasional gravel zones. Thickness 50 Elevation (ft) 0
- Profile A Profile B Profile C
-50
-100
-150
-*,- *-*-*- I * *~ * * * ,- * ,- * - * - * - - . - . - . - . - .., . -. . .'
I 4686 7513 WEATHERED PIEDMONT
- * -*- *- -*-*-*-*-. -*-*-*- * - . ROCK * - * -
- 10369
-200 PIEDMONT ROCK -
METAVOLCANIC
-250 0 2000 4000 6000 8000 10000 12000 Shear Wave Velocity (ft/sec)
Figure 3-1. Plant Area Shear Wave Velocity Profiles A, B and C [4].
Page 15 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Input Rock Motions. The input rock motions were developed using site-specific information rather than the generic inputs described in Appendix B of [2].
Conditional Mean Spectra (CMS) were developed to represent the hard ground motions corresponding to AFE of 10-3, 10-4, 10-5 and 10-6. Two CMS were developed for each AFE level, one to represent earthquake scenarios contributing to hazard for high frequency motions (5 Hz) and one for earthquake scenarios contributing to hazard for low frequency motions (2.5 Hz). For each CMS, 30 time histories were loosely matched to the CMS. More details are in [16].
Kappa. As discussed in [4], the characterization of damping for the site soils does not make use of the parameter kappa. A comparison between an equivalent value of kappa computed using [30] and the value from an empirical relationship in [31]
showed a difference that is less than the standard error reported in [31], showing that the equivalent value of kappa derived from the damping relationships assigned to the HBRSEP soils is considered consistent with the recommendations in [2].
Site response analyses were performed for six analysis cases representing the epistemic uncertainty in site Vs profile (3 alternatives) and non-linear properties (2 alternatives). For each case, analyses were performed for 12 levels of input motion. The resulting amplification values were fit by a piece-wise continuous function defining the variation in ln(amplification) and its standard deviation as a function of the level of input rock motion. Figures in [4] show the results.
The analysis for other control point ground motions followed the approach described above; details and results are presented in [16].
Soil PSHA Following the approach described in [2], the ground motion values at the control point elevation are developed in a hazard-consistent manner by applying Approach 3 in [32].
Approach 3 involves characterizing the amplification of the site soils in terms of the median (mean log) amplification functions and their associated standard deviations. Rather than applying the amplification functions at a post processor on the hard rock hazard as described in [2], the PSHA for the site was recalculated by convolving the soil amplification functions with the ground motion predictions from the rock ground motion models within the hazard integral to produce mean and fractile soil hazard curves at the control point.
More details on the approach are given in [4]. Analyses for other control points followed similar methodology and details are in [16].
Horizontal UHRS and GMRS Results of the PSHA and site response analysis are used to develop the UHRS and GMRS for the site control point at elevation 226 feet. The development of the smooth UHRS for AFE of 10-4, 10-5, and 10-6 is performed in two steps. The first step involves interpolation of the mean soil hazard curves to obtain the ground motion levels at the desired AFE levels for the seven ground motion frequencies recommended in [23]. The second step involves developing smooth interpolation/extrapolation functions using the response spectra computed in the site response analyses to provide smooth UHRS for the ground motion frequency range of 0.1 to 100 Hz (PGA). The performance-based GMRS is then computed from the 10-4 and 10-5 UHRS using the formulation in [25] based on the approach given in [33] for defining a risk-consistent Design Response Spectrum (DRS).
Page 16 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT A similar approach is applied for other control points; details are in [16].
Vertical GMRS and FIRS The vertical response spectra for the HBRSEP SPRA are needed for the GMRS/FIRS.
Vertical response spectra consistent with the horizontal hazard are developed from the horizontal response spectra using vertical-to-horizontal (V/H) spectral ratios as suggested by [34]. The HBRSEP site control point at elevation 226 feet is underlain by approximately 410 feet of firm to stiff soil above hard rock. An envelope of the V/H ratios reported in [35]
and [36] for similar site conditions was used to develop the vertical GMRS for the HBRSEP site. More details are in [16].
The above approach was applied for control points at elevation 216 feet and elevation 163 feet. For the control point at the base of the reactor piles, elevation 159.2 feet, a slightly different approach is used because the piles are founded in the hard clay layer which is stiffer than the average soil and the approach described above may produce V/H ratios that are too low at low frequency. Therefore, the V/H ratios for the base of the piles are calculated using both the envelope spectral ratios discussed above and the ratios developed in [32] for CEUS hard rock conditions of PGA in the 0.2 to 0.5 g range interpolated to the frequency values at which horizontal FIRS are computed. These values and the envelope discussed above are then enveloped and used to develop a vertical FIRS for the elevation 159.2 ft. control point [48]
3.1.2 Seismic Hazard Analysis Technical Adequacy The HBRSEP SPRA hazard methodology and analysis associated with the horizontal GMRS were submitted to the NRC as part of the HBRSEP Seismic Hazard Submittal [4]
and found to be technically acceptable by NRC for application to the HBRSEP SPRA.
The analyses performed for the HBRSEP SPRA described in this Section were subject to in-process peer review against the pertinent requirements in the SPID [2] and the PRA standard [5 and 37]. Comments from the third-party reviewers were addressed and incorporated into the vendor deliverables. HBRSEP ownership of the calculations was assumed, and the vendor deliverables were issued as site calculations. Once complete, the HBRSEP hazard analysis was also subjected to an independent peer review against the pertinent requirements in the PRA standard [5 and 37]. The peer review assessment, and subsequent disposition of peer review findings, is described in Appendix A.
3.1.3 Seismic Hazard Analysis Results and Insights This section provides the final seismic hazard results used in the HBRSEP SPRA. Mean and fractile soil hazard curves for PGA at the 226 ft. elevation control point are used for quantification in the SPRA. These PGA hazard curves are provided in Figure 3-2 and Table 3-2.
Page 17 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table 3-2 HBRSEP Mean and Fractile Exceedance Frequencies for PGA at Elevation 226 ft.
Peak Ground Annual Frequency of Exceedance Acceleration (g) Mean 5th% 16th% 50th% 84th% 95th%
1.00E-03 5.70E-02 2.75E-02 3.80E-02 5.62E-02 7.59E-02 8.91E-02 2.00E-03 3.87E-02 1.95E-02 2.63E-02 4.07E-02 5.50E-02 6.46E-02 3.00E-03 3.03E-02 1.59E-02 2.14E-02 3.24E-02 4.37E-02 5.25E-02 5.00E-03 2.17E-02 1.18E-02 1.62E-02 2.34E-02 3.16E-02 3.98E-02 1.00E-02 1.29E-02 6.92E-03 9.77E-03 1.41E-02 1.91E-02 2.63E-02 2.00E-02 6.95E-03 3.31E-03 4.68E-03 7.24E-03 1.05E-02 1.59E-02 3.00E-02 4.70E-03 1.91E-03 2.82E-03 4.68E-03 7.41E-03 1.18E-02 5.00E-02 2.73E-03 7.59E-04 1.26E-03 2.46E-03 4.47E-03 7.24E-03 7.00E-02 1.81E-03 3.63E-04 6.46E-04 1.48E-03 3.09E-03 5.13E-03 1.00E-01 1.08E-03 1.45E-04 2.82E-04 7.59E-04 1.95E-03 3.39E-03 2.00E-01 2.64E-04 1.62E-05 3.63E-05 1.20E-04 4.57E-04 1.05E-03 3.00E-01 8.93E-05 3.89E-06 9.33E-06 3.31E-05 1.35E-04 3.63E-04 5.00E-01 1.75E-05 5.62E-07 1.48E-06 5.89E-06 2.40E-05 6.46E-05 7.00E-01 5.08E-06 1.38E-07 3.98E-07 1.62E-06 7.08E-06 1.86E-05 1.00E+00 1.24E-06 2.82E-08 9.12E-08 4.17E-07 1.86E-06 4.47E-06 2.00E+00 6.72E-08 8.91E-10 3.39E-09 2.24E-08 1.07E-07 2.82E-07 3.00E+00 1.43E-08 7.76E-11 3.55E-10 3.31E-09 2.14E-08 6.46E-08 5.00E+00 1.72E-09 2.51E-12 1.51E-11 2.14E-10 2.19E-09 8.13E-09 1.00E+01 6.03E-11 8.71E-15 8.13E-14 2.24E-12 5.13E-11 2.40E-10 2.00E+01 1.15E-12 9.33E-18 1.48E-16 8.71E-15 5.37E-13 3.09E-12 Page 18 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Elevation 226 ft 1.E-01 5ti1%
- 16til%
- E 1
.L 50 ti1 % *i-
.L I
841ti1 % *r 1.E-03 951ti1 %
- 3::::3:::I:: _
i----,---t-- --,-,-
J----J---i-- -i-
~---L.-J--1--1-i : I : I 111
*----* ---3/4--~-T-T --------r----r--,-,-T*rr I
I I
I I
I I I I I I I
~----~---r-~--,--,-T_.
*----*-----~-----*--
c----c--:c-J-:c:c1:c
r----r--+-J*-++++
L----L---L-J--L...L.i-L
L----L---L- I I--L-Li-L
r----r----~-*1TTT
- I I 111 I I I I
=======J====J===i==t~=i=i:Ci --------~----~-----~-----*--
c----c--:c-J-:c:c1:c
J----J---i--
_______ J ____ J ___ i __ -i-i-Li
-i-i-L-i --------L----L---L-.J--L-L~-L
L----L---L-.J--L-L~-L
----~---!-- -i-i 1'
]'----,---T--c-T-TI' -------~----~--I'
l----l-- -~-I*
_J_
I' !+I'
- I I
1----,---r--~1-r-r*T I
I I I 11 O I I I
r----r--r-,-
I I I TTT I I I I I I I
~~~~~~~~~~~~~~~~I~~t~I~tl!
- j::::i:::i::t:+/-:i+/-i I I
1'----~---+--r*-+-++T I I I I I I
~-----*--
I O I I I I
-~-~--~--
- c:J::C:Cl:C
--L-.J--L-L~-L
,-,-,,TT
--L-.J--L-L~-L I I I I I I 1 I 111
-rTTTT 1.E-08 -------J----J---*--
l----l---1--c-1-1:Cl
J----J---i--L-i-i-L-i
- ~::::~:::f::~:t:t++
J----J---*--r-*-*--*
_______ J____ J___1___ 1_1.11 II II II II II 1.E-09 iO.:O ,0_1 1 10 Peak Ground A~!1llle:ra *on (g)
Figure 3-2. HBRSEP Mean and Fractile Soil Hazard Curves for PGA for Elevation 226 ft.
Sources of Uncertainty The soil hazard fractiles are produced by the combination of the epistemic uncertainty in the CEUS rock hazard and the epistemic uncertainty in the site response model parameters characterized by the logic trees presented in [16]. The epistemic uncertainty is quantified by the variance in AFE computed for ground motion levels of AFE from 10-3 to 10-7 as shown on Figures 5-15 through 5-21 in [16].
Page 19 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT For low frequency ground motions, the largest contributions to the uncertainty in the hazard are from the uncertainty in the RLME magnitudes and recurrence rates. For the HBRSEP site, this is primarily the uncertainty in the Charleston RLME characterization.
The other major contributors are the uncertainty in the ground motion median models in
[23]. For high frequency ground motions, the uncertainty in the ground motion median models in [23] becomes the largest contributor to the uncertainty in AFE. In addition, uncertainty in the distributed seismicity sources seismicity parameters has a contribution.
The uncertainty in the characterization of the site dynamic properties has a small contribution to the total uncertainty in AFE.
3.1.4 Horizontal and Vertical GMRS This section provides the control point horizontal and vertical GMRS.
The horizontal UHRS and GMRS and the vertical GMRS described in Section 3.1.1 above are provided in Table 3-3 and Figure 3-3. The V/H ratios used as described in Section 3.1.1 are shown in Figure 3-4 and tabulated in Table 3-4. Horizontal and vertical FIRS were also developed for the reactor foundation level at elevation 216 ft., the base of the Lake Robinson dam spillway at elevation 163 ft., and a vertical FIRS for the base of the reactor building piles at elevation 159.2 ft. These FIRS are provided in [16].
Page 20 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table 3-3 Smoothed UHRS for Elevation 226 ft. and Horizontal and Vertical GMRS Spectral Acceleration (g)
Frequency Horizontal Horizontal Horizontal Horizontal Vertical (Hz) 10-4 UHRS 10-5 UHRS 10-6 UHRS GMRS GMRS 100.000 (PGA) 2.88E-01 5.82E-01 1.05E+00 3.03E-01 3.03E-01 90.090 2.88E-01 5.91E-01 1.12E+00 3.07E-01 3.24E-01 83.333 2.88E-01 6.33E-01 1.20E+00 3.25E-01 3.57E-01 66.667 2.88E-01 7.43E-01 1.48E+00 3.69E-01 4.57E-01 60.241 2.88E-01 7.99E-01 1.64E+00 3.91E-01 4.81E-01 50.000 2.98E-01 9.35E-01 1.84E+00 4.46E-01 5.42E-01 40.000 3.34E-01 9.89E-01 1.95E+00 4.77E-01 5.57E-01 33.333 3.63E-01 9.87E-01 1.91E+00 4.85E-01 5.46E-01 25.000 4.45E-01 9.43E-01 1.78E+00 4.87E-01 4.98E-01 20.000 5.00E-01 1.01E+00 1.86E+00 5.24E-01 4.93E-01 16.667 5.42E-01 1.05E+00 1.92E+00 5.52E-01 4.83E-01 13.333 5.90E-01 1.12E+00 2.01E+00 5.92E-01 4.84E-01 11.111 6.15E-01 1.19E+00 2.10E+00 6.24E-01 4.89E-01 10.000 6.32E-01 1.23E+00 2.15E+00 6.44E-01 4.91E-01 8.333 6.00E-01 1.20E+00 2.11E+00 6.28E-01 4.62E-01 6.667 5.82E-01 1.17E+00 2.06E+00 6.11E-01 4.32E-01 5.882 5.95E-01 1.19E+00 2.11E+00 6.23E-01 4.31E-01 5.000 6.12E-01 1.22E+00 2.17E+00 6.36E-01 4.27E-01 4.000 6.39E-01 1.26E+00 2.27E+00 6.59E-01 4.28E-01 3.333 6.38E-01 1.22E+00 2.21E+00 6.45E-01 4.19E-01 3.000 5.96E-01 1.18E+00 2.13E+00 6.17E-01 4.01E-01 2.500 5.25E-01 1.07E+00 1.94E+00 5.59E-01 3.63E-01 2.000 4.70E-01 1.03E+00 1.85E+00 5.28E-01 3.43E-01 1.667 4.54E-01 9.71E-01 1.77E+00 5.01E-01 3.25E-01 1.333 4.05E-01 8.61E-01 1.59E+00 4.44E-01 2.89E-01 1.111 3.30E-01 7.29E-01 1.40E+00 3.73E-01 2.42E-01 1.000 2.84E-01 6.42E-01 1.30E+00 3.27E-01 2.13E-01 0.667 1.39E-01 3.94E-01 9.08E-01 1.92E-01 1.25E-01 0.500 8.38E-02 2.60E-01 6.78E-01 1.24E-01 8.09E-02 0.333 4.77E-02 1.47E-01 4.18E-01 7.03E-02 4.57E-02 0.250 3.66E-02 1.07E-01 2.90E-01 5.20E-02 3.38E-02 0.200 2.86E-02 8.53E-02 2.30E-01 4.11E-02 2.67E-02 0.167 2.27E-02 7.04E-02 1.86E-01 3.37E-02 2.19E-02 0.133 1.78E-02 5.39E-02 1.52E-01 2.59E-02 1.68E-02 0.111 1.35E-02 4.34E-02 1.24E-01 2.06E-02 1.34E-02 0.100 1.16E-02 3.67E-02 1.05E-01 1.75E-02 1.14E-02 Page 21 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT 10 -+--+--~-+-+- ---+--;--+-~-+- -------------~-------+----+-**i***+*-+--+-i*
1----!---+--+: _;__.;.__;_.;._;_1----!---+--+___.;.__i__.;._i--i- -------------i-------.;.____.;.___i---i--i--i-i.
1----1---+--+,--~--+--~-+-+- ----i---l--*--~--~- -------------~-------+----+---~---+--+--+-l t----1---+--+,,--:--: --:- :-:- --- :--:-- :-:--: - -------------:------- :----: ---:---:--:--:-:-
- I I I l I : : l I I I l I I I I I l t----1---+--+*- t----t---+--+--- -- -- - -- - ------------- ------- ---- --- --- -- -- - -
t----t---+--+--- -- -- - *- - ------------- ------- ---- --- --- -- -- - -
t----1----+--+*-- -- -- - -- - ******-*------ ------ ---- --- ... -- -- - -
--Cl 1 ---
C 0
ns Q)
Q) 0
-~ =i* itiii!' . . . . . i~iiJl\IIIi:iiil 0
<( 0.1
, -,--,-,-,-~,*--,--,--,-1**,-*--**--*--**-1**--**-,--**,-*-r-:--,-*,-,*
...ns 0
Q)
C.
Cl) i i ! i i ! l i ! ! l i l i i iii l
! ! i ! ! !i! ! i i i ! i ! ! ! ! !i i i ! i i i 1i i : : 1 1 !l 0.01 rr-T-
.. .. 1e-4UHRS H-I----->--+----<:---;.->*>: -:-t----1-----+--+ t-1" 1----~--+--~;_-j::tJt:t :---- - H t----1---+--+:----:--: --:-: -:-t----1-----+--+
1e-5 UHRS
- - - Horizontal GMRS 1----1---+--+*---- -- -- - - -I----!---+--+ Vertical GMRS ii
!i 0.001 I 0.1 1 10 100 Frequency (Hz)
Figure 3-3. Horizontal Smoothed UHRS for MAFEs of 1e-4 and 1e-5, and the Horizontal and Vertical GMRS (5% damping) at Elevation 226 ft.
Page 22 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table 3-4. V/H Ratios for GMRS at Elevation 226 ft.
Frequency V/H Ratio (Hz) 100.000 1.000 90.090 1.056 83.333 1.101 66.667 1.238 60.241 1.230 50.000 1.216 40.000 1.166 33.333 1.128 25.000 1.023 20.000 0.939 16.667 0.875 13.333 0.817 11.111 0.783 10.000 0.763 8.333 0.737 6.667 0.708 5.882 0.692 5.000 0.672 4.000 0.650 3.333 0.650 3.000 0.650 2.500 0.650 2.000 0.650 1.667 0.650 1.333 0.650 1.111 0.650 1.000 0.650 0.667 0.650 0.500 0.650 0.333 0.650 0.250 0.650 0.200 0.650 0.167 0.650 0.133 0.650 0.111 0.650 0.100 0.650 Page 23 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT 1.4 i----;:::================,-,-----;---,--,---;--,---,-----,----~----,~T7il 1.2 1** ** ** **** ** *** **** .
0.8 I
t -----+---+---+----i!-----:----:----!-..!
0.6
............*..*****....: l'. l***~.Ht 0.4
+'----------- ..******.......... - ... .
0.2 ----------- -----------a,-*--<--* --r 1
t -----+---+---+---+-**-*--+*-+**+**i-------,e---+--+--+--
0 +-----+--~-----,---+--+-t----+---+--~--+--+---+------~-+--+--+--+-~--1 0.1 10 100 Frequency (Hz)
Figure 3-4. V/H Ratios for Two Soil Sites Compared to the Hard Rock V/H Ratios from [31] for PGA of 0.2 to 0.5 g.
3.1.5 Liquefaction Hazard Evaluation - Plant Area In accordance with HLR-SHA-I [5], the HBRSEP is screened for potential secondary seismic hazards. Except for potential for soil liquefaction, no secondary seismic hazards are identified. This section briefly describes the liquefaction evaluation and results. More details are provided in [16].
Liquefaction Screening The broad process described in [38] for screening and evaluation of liquefaction potential is used in the HBRSEP approach with some deviations. The liquefaction evaluation is part of the SPRA project which is not a design-basis assessment, but an assessment that includes behavior under extreme events. Thus, use of conservative parameter selection as would be required for the design basis approach described in [38] is not necessary for the SPRA. Instead, best estimates for the input parameters to calculate factor of safety of liquefaction are used based on the data, and the triggering factor of safety for initiation of liquefaction is taken as 1.00, instead of the value of 1.1 used in [38].
The upper part of the Middendorf Formation in the HBRSEP site profile described previously includes sands with variable densities below the water table. Such conditions cause the HBRSEP site to be screened in for potential for liquefaction; however, it is noted that available records do not show liquefaction as having occurred at the HBRSEP site or nearby. Details of the screening are in [16].
Page 24 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Liquefaction Evaluation Methodology In summary, assessment of liquefaction triggering is done by developing a cyclic resistance ratio, CRR, and a cyclic stress ratio, CSR. The soil characteristics entering into both ratios are developed from site specific data using the best estimate parameters as listed in [16]. Adjustments to the CSR and CRR are made to account for differing earthquake magnitudes, overburden effects and geologic age as described in [16].
Earthquake ground motions representing annual frequencies of exceedance for the four target hazard levels (10-4, 10-5, 10-6 and GMRS) are used. Both high frequency motions and low frequency motions from the site response analysis discussed previously are used in the evaluation.
The CRR is computed using the methods of Boulanger and Idriss [39] with a probability of liquefaction of 50 percent. The calculation methodologies are summarized in [16] with details in [40] and [41].
Liquefaction Evaluation Results - Plant Area Factors of safety were computed to evaluate liquefaction triggering for all exploration points in [42]. The explorations include 30 Standard Penetration Test borings (SPT), three Cone Penetration Test probes (CPT), and two Geophysics boreholes with shear wave velocity (Vs) measurements, for a total of 35 points. Samples above the water table, below the top of the hard clay stratum, or having factors of safety greater than 2.00 are represented with a factor of safety of 2.00.
Table 3-5 summarizes the results of the factor of safety computations. See [40] for detailed figures with results. Samples with factors of safety 1.00 occur with increasing frequency as the hazard level decreases. For example, there are only 12 exploration points at the GMRS hazard level (HF) that have points with factors of safety 1.00, while there are 32 exploration points at the 10-6 hazard level (LF). Instances of liquefaction triggering occur at variable vertical depths within the exploration points, separated by zones without liquefaction.
Page 25 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table 3-5. Summary of Liquefaction Triggering Results in Plant Area Summary of Liquefaction Results (LF)
Number of Liquefaction Zones Target Hazard None 1 2 3 4 or more Level Number of Explorations (Max thickness, Min. thickness) 10-4 AFE 30 5 (27, 2) 0 0 0 GMRS 25 7 (27, 2) 3 (6, 3) 0 0 10-5 AFE 3 14 (40, 3) 10 (19, 1) 7 (27, 1) 1 (9, 4) 10-6 AFE 3 4 (11, 5) 12 (45, 3) 6 (27, 1) 10 (17, 1)
Summary of Liquefaction Results (HF)
Number of Liquefaction Zones Target Hazard Level None 1 2 3 4 or more Number of Explorations (Max thickness, Min. thickness) 10-4 AFE 28 6 (25, 2) 1 (6, 6) 0 0 GMRS 23 10 (25, 2) 2 (9, 3) 0 0 10-5 AFE 11 8 (27, 1) 11 (21, 1) 4 (9, 1) 1 (12, 2) 10-6 AFE 7 10 (40, 1) 12 (27, 1) 1 (8, 3) 5 (12, 1)
While liquefaction could potentially occur at a single exploration, not every sample in the exploration may indicate liquefaction triggering, and in adjacent explorations, liquefaction triggering may not be indicated at the same elevations. As shown in [42], there is not an indication at any of the four hazard levels of a liquefiable soil layer that is continuous across the protected area.
Impacts of Liquefaction - Plant Area Potential impacts of liquefaction are ground settlement and lateral displacement of soils toward the lake.
Ground Settlement. Using the factors of safety for liquefaction determined as described above, probabilistic factors of safety and evaluation of seismic settlement are performed as described in [43]. The impacts to seismic fragilities are discussed in Section 4.
Lateral Displacement. Lateral spreading occurs when a soil mass slides laterally on a liquefied layer, and gravitational and inertial forces cause the layer, and the overlying non-liquefied material, to move in a downslope direction. In order for the lateral spreading to mobilize there must be a continuous layer of liquefying soils so that a failure surface can form and connect to an outlet, such as the free face of a Page 26 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT slope. The locations of free faces are at the Lake Robinson shore and at the discharge canal. Cross sections between locations of the priority Structures, Systems and Components (SSC) and these two features were used to obtain lateral distances and vertical slope heights as discussed in [42].
Groupings of borings that could represent soil conditions along the cross sections are selected from the available site data and a criterion for determining if a cross section has a continuous layer of liquefiable soils is created. That criterion and the results of probabilistic factor of safety evaluations in [43] are used to compute Lateral Displacement Indices with the methodology described in [44].
Non-exceedance probabilities of lateral spreading displacements were evaluated in a step-wise manner as described in [46]. The probabilities of a continuous layer are shown in Table 3-6 and the numerical results are summarized in Tables 3-7, 3-8 and 3-9. The lateral displacement results are used to evaluate fragility estimates as described in Section 4.
Page 27 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table 3-6. Probability of a Continuous Layer of Liquefaction Occurring, Plant Area Prob. of a Continuous Layer Target 3 to 4 8 to 9 13 to 14 Hazard borings (140- borings (490- borings (840-Level 280 feet) 630 feet) 980 feet) 10-4 3.2% 0.0% 0.0%
GMRS 9.6% 0.4% 0.0%
10-5 34.6% 5.7% 1.4%
10-6 47.6% 9.7% 3.6%
Table 3-7. Lateral Displacement (LD) for 140 to 280-foot Cross Section (3-4 borings), Plant Area Target AFE Statistical Value 10-4 GMRS 10-5 10-6 LD (in) LD (in) LD (in) LD (in)
Triggering LD 76.4 (96.9th 46.1 (90.4th 46.0 (65.4th 57.6 (52.4th and Percentile percentile) percentile) percentile) percentile)
Minimum 0.0 0.0 0.0 0.0 2nd Percentile 0.0 0.0 0.0 0.0 16th Percentile 0.0 0.0 0.0 0.0 84th Percentile 0.0 0.0 309.1 370.2 98th Percentile 243.9 388.1 618.8 680.4 Maximum 592.4 739.1 903.3 949.1 Page 28 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table 3-8. Lateral Displacement (LD) for 450 to 630-foot Cross Section (8-9 borings), Plant Area Target AFE Statistical Value 10-4 GMRS 10-5 10-6 LD (in) LD (in) LD (in) LD (in)
Triggering LD 50.3 (99.6th 30.7 (94.3th 20.0 (90.4th None and Percentile percentile) percentile) percentile)
Minimum 0.0 0.0 0.0 0.0 2nd Percentile 0.0 0.0 0.0 0.0 16th Percentile 0.0 0.0 0.0 0.0 84th Percentile 0.0 0.0 0.0 0.0 98th Percentile 0.0 0.0 78.3 86.0 Maximum 0.0 71.5 110.5 115.2 Table 3-9. Lateral Displacement (LD) for 840 to 980-foot Cross Section (13-14 borings), Plant Area Target AFE Statistical Value 10-4 GMRS 10-5 10-6 LD (in) LD (in) LD (in) LD (in)
Triggering LD 73.3 (98.6th 63.8 (96.4th None None and Percentile percentile) percentile)
Minimum 0.0 0.0 0.0 0.0 2nd Percentile 0.0 0.0 0.0 0.0 16th Percentile 0.0 0.0 0.0 0.0 84th Percentile 0.0 0.0 0.0 0.0 98th Percentile 0.0 0.0 0.0 124.5 Maximum 0.0 0.0 160.6 169.0 Page 29 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT 3.1.6 Page 30 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Page 31 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT 4.0 Determination of Seismic Fragilities for the SPRA This section provides a summary of the process for identifying and developing fragilities for SSCs that participate in the plant response to a seismic event for the RNP SPRA. The subsections provide brief summaries of these elements.
4.1 Seismic Equipment List For the RNP SPRA, a seismic equipment list (SEL) was developed that includes those SSCs that are important to achieving safe shutdown following a seismic event, and to mitigating radioactivity release if core damage occurs, and that are included in the SPRA model. The methodology used to develop the SEL is generally consistent with the guidance provided in Seismic Probabilistic Risk Assessment Implementation Guide, EPRI 3002000709, December 2013 [11] and Screening, Prioritization and Implementation Details (SPID) for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1: Seismic, EPRI Repor1025287 [2].
4.1.1 SEL Development The RNP SPRA SEL is developed by using the RNP existing full-power PRA models as the starting point. Use of the PRA models as a starting point for SSCs to consider for fragility analysis is a rational starting point as the PRA models have already identified and modeled SSCs that cover all the critical safety functions and are appropriate for modeling in PRA core damage frequency (CDF) and release frequency models. The process begins by extracting the basic events from the respective models. A tabular list of all the RNP PRA basic events was used as input to this SEL development [64]. In addition, reviews of RNP drawings and RNP PRA System Notebook model boundary diagrams were performed as part of the initial SEL development to confirm that the PRA models are a sufficiently detailed input for the SEL development. These drawing reviews also assisted in locating equipment and identifying various passive failure items not contained in the PRA models.
Once the unique basic event list is generated, the basic events are then reviewed to disposition from further detailed consideration those basic events that need not be carried further in the SEL development process. Such events include:
- Non-applicable initiating events
- Type A and B HEP basic events
- Dependent HEP basic events
- Functional recovery and repair basic events
- Test and Maintenance basic events
- Common cause failure (CCF) basic events
- Flag basic events
- Other basic events SSCs not to be credited in the SPRA include BOP equipment not powered by emergency AC (e.g., Main Feedwater, Condensate and Main Circulating Water). The SEL line items related to these systems are screened out from consideration of future seismic fragility analysis activities.
On-site structures and passive equipment was also reviewed for potential inclusion on the SEL. Structures that house or spatially interact with identified SSCs, as well as those that involve ex-Control Room actions credited in the SPRA, are included in the SEL for future consideration. The following buildings and structures are identified for inclusion on the SEL:
Page 32 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT
- Reactor Containment Building (RCB)
- RB Internal Structure
- Reactor Auxiliary Building (RAB)
- RHR Pump Pit
- Turbine Building Class 1 Bay
- Turbine Building Class 3 Portion
- Fuel Building (#215)
- Maintenance Fab Shop (#390)
- Unit 2 Intake Structure
- DS Diesel Generator Bldg
- AFW "C" Diesel Generator Bldg
- SW Pipe Enclosure
- Radwaste Building (#210)
- RCB Sump The following earthen structures are identified for inclusion on the SEL:
- Intake Pool Submerged Dike
- Discharge Canal
- Robinson Dam The following large above ground storage tanks are identified for inclusion on the SEL:
- Condensate Storage Tank (CST)
- Refueling Water Storage Tank (RWST)
- Diesel Fuel Oil Storage Tank (DFOST)
The following buried items are identified for inclusion on the SEL:
- Deepwell Pumps and Buried Discharge Piping
- Deepwell Pumps A, B and C Buried Electrical
- Deepwell Pump D Buried Electrical
- DS DG Buried Electrical
- Service Water Buried Piping
- Service Water Buried Electrical
- Fire Water Buried Piping
- EDG Fuel Oil Transfer Buried Piping
- EDG Fuel Oil Transfer Buried Electrical In addition, per the EPRI SPRA Implementation Guide, items associated with reactor scram function (reactor internals), offsite power and primary system LOCA were reviewed and added to the SEL.
Previous analyses have been completed in support of determining the seismic risk at all nuclear power plants. As a result, these analyses are used to supplement the basic event review performed in the previous steps to determine if the additional SEL line items are warranted. The following seismic evaluations were reviewed to identify any potential additional SSCs not yet included in the previous steps:
Page 33 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Seismic-induced fragility of the main control room ceiling has been considered in past SPRAs as a potential impact on the human error probabilities used in the PRA. As such, the main control room ceiling is added to the SEL for future consideration. The identification of the Ex-Control Room actions support a review of the pathway availability to the action location and the location itself was performed. This review is documented in the SEL notebook. The following items were added to the SEL for consideration for operator ex-MCR post-initiator action access issues:
- Sliding door to DG Room A
- Sliding door to DG Room B
- CO2 tanks next to MCC-1 Certain types of equipment are inherently rugged such that it need not be considered for fragility modeling in an SPRA. This includes check valves, manually operated valves, disconnect switches and inline piping items (e.g., nozzles, orifices, filters). These items are not included in the SEL based on their very high seismic capacity and their passive nature.
Equipment that is captured through rule-of-the-box considerations, e.g., equipment contained on a skid or in a cabinet that can be subsumed into the major skid equipment or into the cabinet, was also not explicitly included on the SEL. For such equipment, the seismic fragilities for the containing equipment consider all of the equipment in the box.
A walkdown for the RNP SPRA was performed in support of the initial SEL development.
This walkdown covered confirmation of SSC locations, identification of missing items, Rule of the Box grouping, Ex-Control Room Operator Action Access Blockage items, identification of inherently rugged equipment, Seismic, fire, and flood sources identification. In addition, separate seismic capacity walkdowns were performed and documented in the station calculation for the seismic capacity walkdown [53]. These walkdowns identified additional SEL items that were fed back into the SEL.
As a final check, the RNP SEL was compared for reasonableness with the SEL developed for the Surry Power Station during the EPRI Surry SPRA Pilot study (Reference [89], EPRI 1020756, Surry Seismic Probabilistic Risk Assessment Pilot Plant Review, Electric Power Research Institute, July 2010.). This comparison did not identify SSCs inappropriately omitted from the RNP SEL development.
The resulting SEL includes about 339 component entries (Table 4 of the station calculation for the SEL development) [64]. The final SEL was documented for the SPRA.
Page 34 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT 4.1.2 Relay Evaluation During a seismic event, vibratory ground motion can cause relays to chatter. The chattering of relays potentially can result in spurious signals to equipment. Most relay chatter is either acceptable (does not impact the associated equipment), is self-correcting, or can be recovered by operator action. An extensive relay chatter evaluation was performed for the Robinson SPRA, in accordance with SPID [2], Section 6.4.2 and ASME/ANS PRA Standard [5], Section 5-2.2. The results of this assessment show that the chatter of the vast majority of devices is of no consequence to the accident mitigation function of the components supported by the devices, or may require reset of a component from the Main Control Room or locally [72]. Chatter of some devices, typically in a lockout or seal-in type of application, can result in the need to reset a component outside the Main Control Room. The operator action to reset a component outside the Main Control Room may have an impact on the seismic risk and may need to be modeled in the seismic PRA.
The results of the assessment found a relatively small number of unique combinations of relays requiring further assessment in the RNP SPRA. This unique set is summarized in the station calculation for the essential relay development [72]. This listing reflects a total of 90 individual devices/relays involving a total of 114 contact pairs.
4.2 Walkdown Approach This section provides a summary of the methodology and scope of the seismic walkdowns performed for the RNP SPRA. Walkdowns were performed by personnel with appropriate qualifications as defined in the SPID [2]. Walkdowns of those SSCs included on the seismic equipment list were performed to assess the as-installed condition of these SSCs for determining their seismic capacity and performing initial screening.
Previous walkdowns from USI A-46 [48] and IPEEE [51] were used to inform the walkdowns for the RNP SPRA. The components of the RNP A-46 Safe Shutdown Equipment List (SSEL) required a thorough walkdown and review. The essential information for the equipment included in the USI A-46 SSEL had been assembled into data files using the Seismic Qualification Utility Group, Generic Implementation Procedure (GIP) for Seismic Verification of Nuclear Plant Equipment, Revision 2 [50], screening evaluation worksheets (SEWS), anchorage calculations, outlier sheets, photographs, drawings, test reports, and other background documentation. In addition, RNP completed a Seismic Margin Assessment (SMA) for the IPEEE [51]. As part of the IPEEE effort, extensive walkdowns were conducted in 1993 and 1994 for RNP Unit 2 consistent with the intent of the guidelines described in EPRI NP-6041-SL [8]. Because the screening rules for the USI A46 walkdown per the SQUG GIP [50] are similar to the rules for seismic margin walkdown per EPRI NP-6041-SL [8], the components common to both USI A-46 and IPEEE did not need an additional detailed walkdown for IPEEE. The IPEEE seismic capability team reviewed the USI A-46 equipment data files and then performed a walk-by of the equipment for seismic/fire, seismic/flood, and spatial interactions applicable to beyond design basis seismic events, including block walls upgraded under I&E Bulletin 80-11 [52]. The previous walkdowns were particularly useful to obtain information for components with restricted access. Information from those prior walkdowns was used where the appropriate level of detail needed for the SPRA was available.
The walkdowns were performed in accordance with the EPRI SMA methodology report (EPRI NP-6041-SL [8]. Based on EPRI NP-6041-SL guidelines, the Seismic Review Teams (SRTs) reviewed a reasonable sample from each group of similar SEL items in full detail (full scope walkdown) and reviewed the remaining reasonably accessible SEL items via walk-by.
Page 35 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT In full scope walkdowns, the SRT collected detailed notes on the equipment configuration and tag information (e.g., weight, model number), and performed a detailed evaluation against the walkdown caveats and criteria from Appendix F to EPRI NP-6041-SL [8]. The team also collected photographs, measurements, sketches, and any other data that could be used to inform fragility calculations. The walkdowns focused on the seismic ruggedness of the equipment, anchorage capacity, mounting of internal devices, and potential spatial systems interaction concerns. The walkdowns were documented in the Robinson Unit 2 Walkdown Database containing the detailed walkdown findings and observations and SEWS forms.
After a lead item or a sample of similar equipment had been thoroughly reviewed via full scope walkdown, other similar components were reviewed to confirm similarity with the lead items and to verify that there are no anomalies in installation or interaction concerns.
The SRT members judged items to be similar based on equipment construction, dimensions, location, seismic qualification requirements, anchorage type, and configuration. The abbreviated, confirmatory walkdowns are termed walk-bys. The level of detail of the review depended on what was observed during the walk-by. For example, if the anchorage was found to differ substantially from the lead item, the team may have taken new measurements or made new sketches as necessary to adequately describe the differences. In accordance with EPRI NP-6041-SL [8] the walk-by components were documented on the SEWS forms by identifying the equipment number of the component (lead item) to which it is similar and briefly describing any unique seismic interaction issues and/or differences from respective lead items.
EPRI NP-6041-SL [8] guidance states that 100% walk-by of all SEL items is not necessary for equipment classes that have excessively large numbers of like elements, which would include classes such as local instruments and distribution systems such as piping, cable trays, and HVAC ducting. The SRTs performed the walk-bys for these items on an area or sampling-basis as described in EPRI NP-6041-SL [8] Section 2 and Appendix D. The SRTs reviewed selected samples of such items in the vicinity of SEL components to establish the consistency of construction and general robustness of their supports.
4.2.1 Significant Walkdown Results and Insights Consistent with the guidance from EPRI NP-6041-SL [8], no significant findings were noted during the RNP seismic walkdowns. The walkdowns are documented in the station calculation for the seismic capacity walkdowns [53].
Components on the SEL were evaluated for seismic anchorage and interaction effects, effects of component degradation, such as corrosion and concrete cracking, for consideration in the development of SEL fragilities. The potential for seismic-induced fire and flooding scenarios was assessed. Potential internal flood scenarios were incorporated into the RNP SPRA model. The walkdown observations were appropriate for use in developing the SSC fragilities for the SPRA and any seismic interactions which were determined to be credible failure modes were documented and included as failure modes for the associated SEL components.
To support the iterative process for the risk assessment, the seismic capacity of each component was ranked by expert judgment as to whether the component is Rugged or exhibits a High, Medium, or Low seismic capacity. The capacity ranking of the SSCs is based on whether the SSC itself meets the guidelines in EPRI NP-6041-SL [8],
engineering judgment regarding the capacity of the anchorage, and engineering judgment regarding any potential spatial systems interactions. SSCs ranked Rugged are judged by Page 36 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT the SRT to have very high seismic capacity, such that they are not expected to significantly contribute to seismic risk. SSCs ranked High satisfy the applicable EPRI NP-6041-SL [8]
walkdown criteria for 1.2g Sa ground motion and have robust anchorage. SSCs ranked Medium satisfy the applicable EPRI NP-6041-SL [8] walkdown criteria for 1.2g Sa ground motion, but their failure is likely to be governed by anchorage failure modes. SSCs ranked Low have obvious seismic deficiencies as judged by the SRT. The capacity estimates were used for preliminary screening and prioritization.
Electrical cabinets typically met EPRI NP-6041-SL [8] criterion, but anchorage frequently controlled. Heating, Ventilating & Air Conditioning (HVAC) and cable raceways generally have robust support systems and were found to have relatively high seismic capacities.
General observations on potential interaction sources include that lighting fixtures in safety-related areas are generally well supported, HVAC ducting and non-safety piping is generally well supported, fire protection piping is of good construction and generally well supported.
Components on the SEL were evaluated for seismic anchorage and interaction effects, effects of component degradation, such as corrosion and concrete cracking, for consideration in the development of SEL fragilities. In addition, walkdowns were performed on operator pathways, and the potential for seismic-induced fire and flooding scenarios was assessed. The walkdown observations were adequate for use in developing the SSC fragilities for the SPRA.
4.2.2 Seismic Equipment List and Seismic Walkdowns Technical Adequacy The RNP SPRA SEL development and walkdowns were subjected to an independent peer review against the pertinent requirements (i.e., the relevant SFR and SPR requirements)
Capability Category II of Part 5 of the ASME/ANS PRA Standard [37].
A summary of the peer review assessment, and subsequent disposition of peer review findings, is described in Appendix A, and establishes that the RNP SPRA SEL and seismic walkdowns are suitable for this SPRA application.
4.3 Dynamic Analysis of Structures This section summarizes the dynamic analyses of structures that contain systems and components important to achieving a safe shutdown. Modeling of structures is documented in the station calculation for structural modeling [54] and response analyses are documented in the station calculation for seismic response analysis [55]. Table 4-2 lists the structures that support systems and components on the SEL and the type of model used to perform the dynamic analysis, whether finite element model (FEM) or lumped mass stick model (LMSM). The table also lists whether effects of soil-structure interaction (SSI) were included in the dynamic response analysis, the structure damping used, and the parameters varied in the analysis. Effects of ground motion incoherence (GMI) on structure responses were evaluated in the station calculation for the RAB incoherence sensitivity study [56] and found to be insignificant.
Page 37 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table 4-2 Description of Structures and Dynamic Analysis Methods for RNP SPRA Best-Model Estimate Building SSI Varied Parameters Type Structure Damping Soil properties, structure Reactor stiffness, pile-to-soil Containment New FEM* Yes 4%
interface stiffness, Building (RCB) damping Soil properties, structure Reactor Auxiliary stiffness, pile-to-soil New FEM Yes 4%
Building (RAB) interface stiffness, damping Class I Turbine Revised No 2% Structure stiffness Building FEM Class III Turbine New No 7%** Structure stiffness Building FEM**
Partial Unit 2 Intake Building No N/A None Structure FEM Model***
- The new FEM incorporates an existing LMSM of the containment shell that was revised to meet SPID modeling criteria
- A separate model and damping was used to compute the building fragility
- A modal analysis was performed on a model of the Unit 2 Intake Structure roof slab to demonstrate that the load path to the SSCs mounted in the structure is not flexible enough to significantly amplify the input motion In addition to the buildings listed above, analysis was performed on the RNP site to determine probabilistic distributions of liquefaction-induced settlement to support fragility analyses of SSCs sensitive to ground settlement. This analysis is documented in the station calculation for liquefaction settlement [43].
4.3.1 Cracking Analysis Before performing response analyses to compute in-structure response spectra (ISRS),
each uncracked building model is evaluated for effects of concrete cracking.
In a preliminary study on pile modeling for the RAB and RCB, the fragility vendor noted that the lateral flexibility in the pile foundations causes these buildings to behave as rigid structures on a flexible pile foundation in an earthquake. Due to this behavior, stresses in the RAB and RCB superstructures are low, and cracking of concrete structural elements does not occur at the site reference earthquake.
Cracking in the Turbine Building structures is evaluated using a response spectrum or time history analysis with the applicable input motion and the appropriate damping level based on Table 3-2 of ASCE/SEI 43-05 [58]. Stresses from the cracking analysis are reviewed to identify areas in the building that are expected to crack due to the reference Page 38 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT earthquake and result in softening that could affect the input motion seen by SEL equipment. Structure stiffnesses are determined following Section 3.3 of ASCE/SEI 4-16
[59] with stiffness reductions in elements where cracking is expected.
Judgment is exercised when assessing whether cracking is sufficient to warrant stiffness reduction. Very localized cracking that does not significantly affect overall structure seismic response and force distribution may not require stiffness reduction. Conversely, stiffness reduction may be appropriate for localized cracking that affects local structure response of significance to SSC fragilities.
4.3.2 Fixed-base Analyses Fixed-base dynamic response analyses are performed for the Class I and III Turbine Building structures. Response uncertainties are considered by analyzing models with best estimate (BE), lower bound (LB), and upper bound (UB) structure stiffness.
To obtain ISRS, each of the three structure stiffness models are analyzed for each of the applicable five earthquake input acceleration time history sets.
For each of the three structure stiffness models, the average of ISRS for the five sets of earthquake acceleration time histories is determined. The median ISRS is taken as the average of values for the three structure stiffness cases and the 84% non-exceedance probability (NEP) ISRS as the envelope of values for the three structure stiffness cases with valleys between the peaks of the median and 84% NEP ISRS filled in.
The ISRS are generated at an array of damping values ranging from 0.5% to 15% to cover damping values that might be needed for system and component fragility evaluation.
4.3.3 Soil Structure Interaction Analyses SSI analyses are performed for the RCB and RAB using computer program SASSI. These structures are supported on piles and the SSI effects on their seismic responses are non-negligible. Pile elements are modeled with horizontal pile-to-soil springs connect the pile beam elements to the soil to represent the local flexibility and horizontal interaction between the piles and soil. The stiffness of these springs is based on pile lateral load test data.
The pile test data is judged to be stiffer than the expected response during seismic loading.
To account for the local soil degradation due to cyclic loading effect, the computed scale factors are reduced to obtain the stiffness for the pile-to-soil springs.
Separate sets of SSI analyses are performed to permit determination of structure response variability due to uncertainty in soil properties, structure stiffness, pile-to-soil spring stiffness, and damping. For each varied parameter, median and 84% NEP ISRS are obtained similar to the process described for varying structural stiffness in fixed-base analyses using BE properties for all other modeling parameters. Only BE and LB damping values are considered for structural damping. For both buildings, effects of structural damping are found to be insignificant.
The median ISRS for seismic fragility evaluation is the average of (1) the median ISRS considering only variability in soil properties, (2) the median ISRS considering only variability in structure stiffness, and (3) the median ISRS considering only variability in pile-to-soil interface stiffness.
Spectra are computed for several damping levels, as required for the subsequent fragility evaluations. In addition to the ISRS calculated for seismic fragility evaluation, spectra are Page 39 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT computed in the free field for evaluation of SSSI effects on the buildings adjacent to the RCB and RAB. It is found that the only building that is affected by the SSSI effects is the Class 1 Turbine Building.
4.3.4 Structure Response Models The type of models used for dynamic response analyses are listed in Table 4-2. To meet SPID [2] modeling criteria for these structures, we developed new state of the art FEMs with the following exceptions. The model of the RCB included a new FEM representation of the internal structure as well as a recreated FEM of the Nuclear Steam Supply System (NSSS) and a revised version of an existing LMSM representation of the containment shell. The recreated NSSS model meets SPID [2] criteria and the original LMSM of the containment shell was modified to bring it into compliance with SPID [2] requirements by revising the modulus of elasticity, refining model discretization, and including mass moments of inertia. We used an existing FEM of the Class I Turbine Building with certain modifications to make it a median-centered model satisfying SPID [2] modeling criteria.
4.3.5 Seismic Structure Response Analysis Technical Adequacy The RNP SPRA Seismic Structure Response and Soil Structure Interaction Analysis were subjected to an independent peer review against the pertinent requirements of Capability Category II of Part 5 of the ASME/ANS PRA Standard [37].
The peer review assessment, and subsequent disposition of peer review findings, is described in Appendix A, and establishes that the RNP SPRA Seismic Structure Response and SSI Analysis are suitable for this SPRA application.
4.4 SSC Fragility Analysis The SSC seismic fragility analysis considers the impact of seismic events on the probability of SSC failures at a given value of a seismic motion parameter, such as peak ground acceleration (PGA), peak spectral acceleration, floor spectral acceleration, etc.
The fragilities of the SSCs that participate in the SPRA accident sequences, i.e., those included on the seismic equipment list (SEL), are addressed in the model. Seismic fragilities for the significant risk contributors, i.e., those which have an important contribution to plant risk, are intended to be generally realistic and plant-specific based on actual current conditions of the SSCs in the plant, as confirmed through the detailed walkdown of the plant.
This section summarizes the fragility analysis methodology, and the calculation method and failure modes) for those SSCs determined to be sufficiently risk important, based on the final SPRA quantification (as summarized in Section 5). Important assumptions and important sources of uncertainty, and any particular fragility-related insights identified, are also discussed.
4.4.1 SSC Screening Approach A screening level that would result in a seismic core damage frequency (SCDF) of 5E-7/reactor-year was developed in accordance with SPID [2] guidance. This screening level for rugged SSCs is a high-confidence-of-low-probability-of-failure (HCLPF) capacity of 0.75g. Rugged SSCs are not removed from the SPRA model but are modeled with their respective fragilities, at least the screening level fragility. The screening-level fragility is assigned to several seismically rugged SSCs whose seismic capacity is determined to be equal to or greater than the screening level based on analysis or engineering judgement coupled with walkdown observations. Items commonly ranked rugged include local Page 40 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT instruments, sensors, transmitters, and large-diameter valves. The items must be obviously well anchored and free of interaction concerns. The screening level is validated in Section 4.2 of the station calculation for the logic model quantification notebook [13].
For the RNP SPRA, the initial quantification was performed with representative fragilities that were conservatively biased, resulting in conservative SCDF and SLERF estimates.
Then, for the remainder of the RNP SPRA project, significant efforts were expended to eliminate or minimize conservatisms in top risk contributor fragilities to avoid a situation where overall risk results and insights were unnecessarily masked by conservatively biased fragilities. This necessitated numerous refinement iterations in risk quantification.
For each iteration, risk insights were gained, and top contributors were identified. For the newly identified top contributors that were conservatively biased, more detailed fragilities were developed as SOV fragilities. The revised fragilities enabled a new iteration of risk quantification. The refinement process continued until reasonably converged quantification results were obtained with respect to risk results and insights, conditional upon a holistic review of fragility quality that top contributor fragilities were either realistic or, if they were not, the degree of conservatism was justified for its relative impact on the overall risk results (i.e., quantification results are not sensitive to potential fragility improvements).
4.4.2 SSC Fragility Analysis Methodology For the Robinson SPRA, the following methods were used to determine seismic fragilities for SSCs included in the SPRA:
The fragility evaluation effort began with the development of representative fragilities for all SSCs on the SEL as documented in the station calculation for representative fragility development [61]. These representative fragilities were used in an initial risk quantification to identify the most important SSCs and thus focus subsequent fragility evaluation efforts.
In many cases, the representative fragilities are conservatively biased and/or based on generic data (e.g., earthquake experience). Following each risk quantification conducted for the RNP SPRA, meetings were scheduled with the SPRA team to jointly determine the optimal path forward in terms of fragility refinements. Fragilities were refined for the most dominant risk contributors using increasingly sophisticated techniques to make them more realistic and plant-specific. Using the refined fragilities, the systems analyst then re-quantified the seismic risk and reprioritized the SSCs. This iterative process was repeated until the top risk contributors were ultimately characterized with a realistic and plant-specific fragility.
The final fragilities used to quantify the seismic risk of the plant are a combination of representative fragilities and more detailed fragilities (SOV fragilities).
4.4.2.1 Representative Fragilities Representative fragilities are based on a combination of the following:
- Design information and calculations
- Seismic evaluations from the Unresolved Safety Issue (USI) A-46 [48]
and Individual Plant Examination for External Events (IPEEE) [51]
programs
- Judgements and rankings made during the seismic walkdowns
- Past fragility estimates and experience Page 41 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT
- Recent assessments made at RNP as part of the Near-Term Task Force (NTTF) requirements of all U.S. nuclear plants (NTTF 2.1 and 2.3 requirements)
- Response analyses described in Section 4.3 In developing the initial representative fragilities, a slight conservative bias was incorporated to reduce the potential for changes to the risk ranking during subsequent risk quantifications.
4.4.2.2 Detailed fragilities In most cases, detailed fragilities were performed following the SOV methodology documented in EPRI TR-103959 [62] along with more recent guidance from EPRI 1019200 [63] and EPRI 1025287 [2]. This involves identifying critical failure modes, computing median capacities (Am) including effects of inelastic energy absorption, and computing lognormal standard deviations for randomness and uncertainty (r and u).
Liquefaction-induced settlement or deformation related fragilities do not fit a double lognormal distribution and required to be expressed as mean conditional probabilities of failures at the multiple hazard levels.
4.4.3 SSC Fragility Analysis Results and Insights Refer to Section 5, Tables 5.4-2 and 5.5-2, for a tabulation of the fragilities for those SSCs (or correlated SSC groups) determined to be risk important, based on the final SPRA quantification. Tables 5.4-2 and 5.5-2 provide the risk important fragilities for SCDF and SLERF, respectively. The tables provide for each listed fragility, the median capacity and uncertainties (e.g., Am, r, u), the calculation method, and the failure mode(s) addressed in the model.
4.4.4 SSC Fragility Analysis Technical Adequacy The RNP SPRA SSC Fragility Analysis was subjected to an independent peer review against the pertinent requirements Capability Category II of Part 5 of the ASME/ANS PRA Standard [37].
The peer review assessment, and subsequent disposition of peer review findings, is described in Appendix A, and establishes that the RNP SPRA SSC Fragility Analysis is suitable for this SPRA application.
Page 42 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT 5.0 Plant Seismic Logic Model The seismic plant response analysis models the various combinations of structural, equipment, and human failures given the occurrence of a seismic event that could initiate and propagate a seismic core damage or large early release sequence. This model is quantified to determine the overall SCDF and SLERF and to identify the important contributors, e.g., important accident sequences, SSC failures, and human actions. The quantification process also includes an evaluation of sources of uncertainty and provides a perspective on how such sources of uncertainty affect SPRA insights.
5.1 Development of the SPRA Plant Seismic Logic Model The Robinson Nuclear Power Plant seismic response model was developed by starting with the Robinson Nuclear Power Plant internal events at power PRA model of record as of June 2015, and adapting the model in accordance with guidance in the SPID [2] and PRA Standard [5 and 37], including adding seismic fragility-related basic events to the appropriate portions of the internal events PRA, eliminating some parts of the internal events model that do not apply or that were screened-out, and adjusting the internal events PRA model human reliability analysis to account for response during and following a seismic event.
The general PRA modeling elements addressed in the development of the SPRA model are listed below, followed by a short description summarizing the treatment of each element.
- Initiating Event Analysis (IE)
- Accident Sequence Analysis (AS)/Success Criteria (SC)
- Systems Analysis (SY)
- Human Reliability Analysis (HR)
- Data Analysis (DA)
- Quantification (QU) 5.1.1 Initiating Event Analysis (IE)
Initiating events and consequential events that can be caused by a seismic event were considered by examining the SEL. Each SSC was examined to determine the plant impact from its failure. Since the plant impacts are already addressed by existing event trees, no new initiators needed to be considered. Passive failures, which may not have been represented in the internal events PRA, were given special attention, especially building failures.
Additionally, plant-specific seismic risk evaluations for the Surry [89] and Oconee [90]
nuclear plants were reviewed to ensure all applicable initiating events have been accounted for in the RNP SPRA. These evaluations considered events, such as, seismic-induced flooding, seismic-induced fires, and seismic-induced failures of structures and dams, which are similarly evaluated in the RNP SPRA. No new initiating events or accident sequences were included in the RNP SPRA based on the review of the Surry and Oconee SPRAs.
Plant-specific seismic events are required to be considered in the RNP SPRA, if applicable. The intent of the requirement is to ensure the SPRA includes plant-specific seismic events and appropriately models the plant response to the event. No plant-specific events have occurred.
Page 43 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT 5.1.2 Accident Sequence Analysis (AS)/Success Criteria (SC)
This element consists of modeling the plant response using event trees, and identifying those functions and systems used to mitigate the modeled initiator. This PRA element starts with the review of the initiators identified in element (IE). No new accident sequences were developed, and so it was not necessary to modify the success criteria of the base model for seismic events. However, there is one variance in the success criteria for the SPRA due to crediting the Diverse and Flexible Coping Strategies (FLEX) systems and strategies. As discussed in the FLEX PRA system notebook, for events mitigated by FLEX strategies, the success criteria are limited to Phase 1 and Phase 2 FLEX strategies. Per MAAP analyses, core damage is prevented, and the plant is stable for at least 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> following the implementation of the Phase 2 FLEX strategies [73 and 66].
The PRA Standard requires examining the effect of including a small-small (also known as a very small) LOCA as an additional fault within each sequence in the Seismic PRA model. Very small LOCA (VSLOCA) break sizes of less than 0.35 were screened out of the internal events PRA because they could be mitigated by the highly reliable charging system. In the SPRA, these very small breaks are considered because the charging pumps are highly correlated and can fail concurrently during a seismic event. Thermal-hydraulic analysis [73] demonstrated a VSLOCA can be conservatively modeled with the same timing and success criteria as a Small LOCA, and no new event tree development was necessary. FLEX strategies are credited with mitigating a VSLOCA in the SPRA. In addition, FLEX is also credited for CST makeup and alternate suction to the SDAFW pump from the Condenser Inlet Waterbox.
Some SSCs included in the internal events PRA are not credited for the SPRA model; for example, Circulating Water pumps, primary water equipment, Fire Water equipment, and primary instrument air are not credited for the SPRA model due to their power source not being backed by a diesel generator and no credit is given for LOOP recovery for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
5.1.3 Systems Analysis (SY)
The Internal Events PRA [88] system fault trees developed for the event tree functions were used as a starting point for the SPRA. Seismic-related events were inserted into the fault tree, and they were linked to 10 seismic acceleration levels. In general, at least 6 intervals are used in SPRAs and using 10 intervals for RNP should accurately represent the hazard curve. For SSCs which have no corresponding Internal Events PRA Basic Event (BE), the SSC was tied to events or gates which represented the same plant impact as the seismically failed SSC. Many SSCs are seismically rugged, and therefore are unchanged from the internal events PRA: e.g., check valves, manual valves, filters, dampers, and sensors; no seismic failure mode is added, only the nominal failure probability is included.
Relay chatter during a seismic event may cause SSCs to inadvertently actuate or lockout from actuating during or following the seismic-induced shaking. The plant specific assessment of relay chatter is provided in the station calculation for relay chatter systems evaluation [72].
FLEX equipment has been credited and modeled in the Robinson SPRA for mitigating seismic-induced failures. The key functions of FLEX credited for the Robison SPRA are the SG makeup and RCS makeup capabilities [77].
Page 44 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT 5.1.4 Human Reliability Analysis (HR)
Pre-initiator actions are not affected by seismic events and so their assessments were not changed from the internal events PRA model. The list of post initiator human actions for the internal events model was analyzed for modification due to seismic impacts. Only human failure events (HFE) associated with the sequence models used to represent seismic initiators were retained in the model and any new operator actions added for the SPRA.
Post-initiator HFEs retained in the SPRA cutsets were evaluated for the impacts of seismic events. The degree of impact was assumed dependent on the seismic acceleration level. At very high accelerations, the human error probabilities (HEPs) were set to 1.0. The seismic impacts on every post-initiator HFE in the SPRA models is accounted for by the HFE specific, performance shaping factors, and selected minimal values that increase with acceleration as a function of plant damage as documented in the station calculation for HRA [65].
5.1.5 Data Analysis (DA)
Equipment failure data for random failures, test and maintenance unavailabilities, and plant configuration data are unchanged from the internal events PRA model. The increasing SSC seismic failure probabilities with acceleration interval are computed from the fragility curves developed in the fragility analysis. Equations are developed in terms of the Am, r, and u parameters of the SSC seismic fragility curves, and the failure probability of each new seismic event is evaluated using these equations as a function of the seismic acceleration level that applies.
5.1.6 Quantification (QU)
The 10 seismic interval frequencies are included as separate initiating events. Each SSC seismic failure is combined in a single OR gate with the associated seismic initiating event under an AND gate. The placements of these OR gates in the single linked fault tree is dictated by the BE(s) associated with the SSC. An SSC may impact more than one BE, there may be more than one failure mode for a basic event, and a single BE in the linked fault tree may be impacted by more than one SSC failed by a seismic event.
The quantification of the SPRA fault tree is accomplished using standard EPRI developed software used to analyze risk from external events at power plants. The seismic PRA quantification notebook [13] describes the use of this software for seismic events in detail.
5.1.7 Seismic Sequence Model Seismic Initiating Event Impacts The seismic initiating events are addressed by the seismic master event tree in the station calculation for plant logic modeling [77]. This event tree maps a seismic initiating event to an internal events initiating event based on the availability of certain SSCs. The seismic master event tree transfers to the following internal events initiating events or combination of internal events initiating events:
- Reactor / Turbine Trip (%T1)
- Loss of Main Feedwater (%T4)
- Loss of Offsite Power (%T5G)
Page 45 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT
- Loss of Service Water (%T9)
- S1LOCA (%S1)
- S2LOCA (%S2)
- Medium LOCA (%M)
- Large LOCA (%A)
- Vessel Rupture (%EXLOCA) (surrogate for direct core damage from failure of NSSS components beyond mitigation capability of the ECCS)
Seismic Initiating Event Frequencies The seismic hazard curve is shown in Figure 3-2. The 100 Hz spectral acceleration is selected to represent the zero period acceleration or PGA. From the hazard curve the RNP Safe Shutdown Earthquake (SSE) at 0.20g has a mean hazard exceedance frequency of 2.64E-4 per year. The hazard exceedance frequency is at 1.75E-5 at 0.5g and 1.24E-6 per year at 1.0g.
The seismic initiating event frequencies and their associated acceleration intervals are found in Table 5.1-1. The lowest acceleration chosen was 0.1g. Relatively narrow acceleration intervals were selected for those ranges of acceleration where the conditional core damage probability was expected to change most quickly, and to aid in the demonstration that adding new SSC fragilities with higher capacity would not significantly impact the computed CDF. The higher end of the range of accelerations were retained to evaluate LERF.
Table 5.1-1: Seismic Initiating Event Intervals IE PGA Lower PGA Upper IE Frequency
%G01 0.10 0.15 5.67E-04
%G02 0.15 0.20 2.49E-04
%G03 0.20 0.25 1.16E-04
%G04 0.25 0.30 5.86E-05
%G05 0.30 0.35 3.26E-05
%G06 0.35 0.40 1.94E-05
%G07 0.40 0.50 1.98E-05
%G08 0.50 0.60 8.47E-06
%G09 0.60 0.90 7.16E-06
%G10 0.90 --- 1.88E-06 Fault Tree Model Changes Using FRANX minimizes the need to make changes to the RNP fault tree to model the plant response to a seismic event. However, some logic adjustment is needed to control the fault tree logic used by FRANX for the SPRA.
- Basic Event S-DAM is added to model dam failure failing fire water and service water.
Page 46 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT
- Basic Event DUMMY_CD was inserted as input to the core damage top event to serve as a placeholder to map seismic failures that lead (or are assumed to lead) directly to core damage, such as structural failure of the containment building.
Logic was added to the fault tree to credit the FLEX system. The FLEX system is credited with providing backup suction for the SDAFW pump, alternate reactor coolant system (RCS) boration and makeup, and CST makeup. Logic was also added to the fault tree for FLEX alternate auxiliary feedwater to the steam generators.
A complete list of fault tree changes is documented in the station calculation for the plant logic modeling [77].
5.1.8 SSC Correlation SSCs retained in the model were assigned to correlation sets by their seismic capacities and failure modes. Equipment is considered correlated for seismically induced failures if they meet all of the following four conditions: 1) located in the same building, 2) located on the same elevation in the building, 3) identical or essentially identical equipment and
- 4) the same orientation.
The model assumes complete correlation, which means that if one equipment item in the fragility group fails seismically, all others in that set are also assumed to fail. This 100%
correlation approach conservatively minimizes the advantages of redundancy. Groups of equipment with different failure modes were split into different fragility groups because different failure modes are not correlated; for example, a functional failure of a panel and failure due to failure of the block wall on which it is mounted were put in different fragility groups. Further, potentially risk-significant fragility groups with significantly different fragilities were split into multiple fragility groups based on their seismic capacity and analyst judgement, supported by justifications provided by the fragility analysts. Once fragility groups were finalized, fragility values larger than the lowest value within the group were changed to match the most conservative value in the group. The final list of fragility groups and their numerical characterization are located in the station calculation for the plant logic modeling [77].
Relays were correlated, if they were 1) located in the same host equipment, 2) were the same relay make and 3) were the same relay model. If relays failed any of these checks, they were considered uncorrelated.
5.1.9 Seismic-Induce Floods and Fires The fire PRA analysis existing at the time of the fire/flood seismic walkdowns was used to develop the list of fire ignition sources for the seismic fire assessment for RNP SPRA.
The approach taken for seismic, focused on a sub-set of fire ignition sources identified from the Fire PRA that have the potential to become fire sources in the seismic context.
This took advantage of insights gained from a systematic review of fire PRA results and past fire evaluations with respect to the type of ignition sources relevant to seismic and resulted in identifying flammable gases and liquids as representative seismic-fire ignition sources to be considered for SPRA. These flammable gas/liquid sources were determined to be Hydrogen Piping, Station Transformers, Diesel Fuel Oil Storage Tank, Diesel Fuel Oil Day Tank, Turbine Lube Oil Reserve and Lube Oil Storage Tank, Engine-Driven Fire Pump and Flammable Material Storage Cabinet. In addition, general plant areas and rooms were walked down as documented in the station calculation for the SPRA walkdowns [53] to assess whether any additional flammable gas/liquid sources existed in the plant that were not already identified from the fire PRA analysis existing at Page 47 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT the time. When the SRT walked by each area, they looked for any credible seismic fire concerns related to flammable gas/liquid sources and documented their observations and findings in a SEWS. Based on the results of the seismic induced fire assessment, there were no seismic induced fire scenarios identified for inclusion with the SPRA.
Seismic component walkdowns, area walkdowns and reviews of past internal flood analysis [80] were systematically utilized to support the identification of potential seismic-induced flood sources to be added to the SEL. However, the SPRA approach was not to assess/screen flood sources just based on the Internal Flooding PRA, but to use the walkdowns to identify piping with the unusual features that could cause failure from a seismic event. In other words, the intent is not necessarily to itemize and disposition every passive component in the plant. For example, small tanks of limited volume that pose no significant flood hazard, and do not fail systems of interest to the SPRA, need not be itemized and individually dispositioned. To this end, area flooding walkdown reviews were conducted by the SRT and SEWS developed for Fire protection piping and other flooding sources beyond the SEL items. It should be noted that, based on experience with other SPRAs, particular attention was given to the Fire Water lines in RAB, since there can be the potential for threaded joints and piping interactions for the non-seismic design piping, and the size of the flood can be large. The seismic capacity walkdown team observed that this piping is adequately supported and free of vulnerabilities such as flexible headers with stiff branch lines and insufficient clearance between sprinkler heads and adjacent objects. For sprays, the equipment walkdowns included assessing potential piping failures near the SEL equipment. The SEL has entries for flood and fire potential for all of the area walkdowns. After a series of qualitative and quantitative screening steps taken to disposition all identified flooding sources on the SEL, the Fire Water lines in RAB was identified as one of flooding sources that were ultimately retained in the SPRA for quantification purposes as documented in the station calculation for the Seismic Induced Fire/Flood Assessment
[69].
5.2 SPRA Plant Seismic Logic Model Technical Adequacy The Robinson Nuclear Power Plant SPRA seismic plant response methodology and analysis were subjected to an independent peer review against the pertinent requirements in the PRA Standard [37].
The peer review assessment, and subsequent disposition of peer review findings, is described in Appendix A, and establishes that the Robinson Nuclear Power Plant SPRA seismic plant response analysis is suitable for this SPRA application.
5.3 Seismic Risk Quantification In the SPRA risk quantification the seismic hazard is integrated with the seismic response analysis model to calculate the frequencies of core damage and large early release of radioactivity to the environment. This section describes the SPRA quantification methodology and important modeling assumptions.
5.3.1 SPRA Quantification Methodology The Robinson SPRA model has been created using the linked fault tree approach. The EPRI R&R workstation code package is used to support development and quantification of the model. The analytic tools for the development of a quantified model are the EPRI CAFTA code suite augmented by EPRIs ACUBE Binary Decision Diagram (BDD) software.
Page 48 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT 5.3.2 SPRA Model and Quantification Assumptions The following assumptions were used in development of the RNP SPRA model:
All assumptions and success criteria in the Internal Events PRA are retained in the SPRA for the portions of the sequence models that apply [77]. This assumption provides continuity between the Internal Events PRA and the SPRA.
Any future changes to the Internal Events PRA success criteria would be addressed as part of the maintenance and update process of the integrated PRA.
- A plant trip is assumed for all seismic initiating events.
- The portions of the internal events PRA model that apply to seismic events are the following: transients (including loss of offsite power (LOOP), secondary line breaks, loss of service water, loss of CCW, loss of Instrument Air, loss of CVCS, and loss of feedwater), loss of coolant accidents (LOCAs) (small, medium, large, and interfacing-system), reactor vessel rupture, internal fire, and internal flood.
- The screening criterion for excluding structures, systems, and components (SSCs) from the SPRA model is an SSC High Confidence of a Low Probability of Failure (HCLPF) peak ground acceleration (PGA) of 0.75g or higher as documented in the station calculation for the logic modeling notebook [77].
- Seismic SSC failures are assumed to be complete failures, in that the SSC fails to perform its function. Degraded states of equipment for the period following the seismic initiator are not represented.
- The assumed SSC seismic failure mode depends on the SSC type and whether the fragility applies to functional failure, structural failure, or block wall or other interaction failure.
- Many SSCs are seismically rugged, and therefore the failure probability is unchanged from the internal events PRA (e.g., check valves, manual valves, filters, dampers, and sensors).
- If the seismic event results in a LOOP, recovery of offsite power is not credited in the SPRA during the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the seismic event.
- Building fragilities are included in the model. Failure of key structures (Auxiliary Building, Reactor Containment Building, and Reactor Containment Building) are conservatively assumed to lead to core damage. Failure of key structures is also assumed to lead to a large early release. Other building failures lead to failure of the equipment function in the building (e.g., failure of the dam leads to loss of service water and fire water).
- Non-rugged/offsite power dependent equipment screened from the SEL [64] is not credited in the model. Therefore, this equipment is assumed failed in the model. These include such SSCs as Circulating Water pumps, Main Feedwater and Condensate equipment, primary instrument air train, D instrument air train, Fire Water equipment, and primary water equipment.
5.4 SCDF Results The seismic PRA performed for Robinson Nuclear Power Plant shows that the point estimate mean seismic CDF is 9.27xE-05/reactor-year. A discussion of the mean SCDF with uncertainty distribution reflecting the uncertainties in the hazard, fragilities, and model data is presented in Section 5.6. Important contributors are discussed in the following paragraphs.
Page 49 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT The top SCDF cutsets are documented in the SPRA quantification report [13]. These are briefly summarized in Table 5.4-1.
SSCs with the most significant seismic failure contributions to SCDF are listed in Table 5.4-2, sorted by Fussell-Vesely (FV). The seismic fragilities for each of the significant contributors is also provided in Table 5.4-2, along with the corresponding limiting seismic failure mode and method of fragility calculation.
The FV for each fragility basic event, has been calculated by summing the criticality in each interval from the ACUBE output. The top ten events for the SCDF are described here.
As can be seen from Table 5.4-2, the RNP SCDF is dominated (43.71%) by the pounding-induced cracking failure of the Class III Turbine Building (TB3); this is a loss of structural integrity due to the Turbine Building Mezzanine floor pounding into the RAB.
Failure of TB3 leads to a LOOP and failure of AFW C, as well as a high chance of failure of the Class I Turbine Building (TB1), which houses the piping required to provide flow to the steam generators and piping from the CST. The second highest contributor (12.57%)
is the liquefaction-induced settlement failure of the DFOST piping leading to failure of the EDGs. Contributor #3 (11.81%) is the failure of the TB gantry crane which has a high probability of interacting with and failing the CST. Contributor #4 (7.01%) is the liquefaction-induced settlement failure of the SDAFW pump piping.
. Contributor #6 (2.59%) is the liquefaction-induced settlement failure of underground cable trays running from the Intake Structure to the RAB, which contain cables for the SW system and is assumed to fail SW. Contributors #7 and #8 (1.87%) are two separate relay chatter groups which cause a loss of the EDGs.
Contributor #9 (1.70%) is the liquefaction-induced settlement failure of North Header SW piping at the Intake Structure, which leads to the loss of one of two SW headers.
Contributor #10 (1.54%) is failure of SSCs associated with offsite power.
Page 50 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table 5.4-1: Summary of Top SCDF Cutsets Cutset Event
- Event Description Cutset Description Prob Prob Seismic Initiating Event (0.4g to 1 8.32E-06 1.98E-05 %G07
<0.5g)
SEISMIC FRAGILITY FOR 9.13E-01 SF-TB_CRANE-C-G07 %G07: Turbine Building Gantry Bin %G07 seismic event causes a plant trip. There is a seismic failure of the Crane Class 3 Turbine Building (TB3) due to pounding-induced cracking, which leads to SEISMIC FRAGILITY FOR a LOOP and failure of AFW C. Liquefaction-induced settlement failure of the SF-TB-CLASS 9.52E-01 %G07: Turbine Building Class 3 DFOST fails EDGs; thus, loss of all AC power. The TB Gantry Crane fails and POUND-C-G07
- Pounding-induced cracking interacts with and fails the CST failing the normal supply to the SDAFW; failure of SF-TK-DG-FOSTRG- DFOST Liquefaction-Induced TB3 prevents alignment of the Condenser Water Box to the SDAFW pump 7.20E-01 TNK_SETTLE_G07 Settlement suction. Loss of all injection/recirculation and SSHR leads to core damage.
TB Crane Failure Interacts with 7.50E-01 TBCRANE-CST CST Causing Failure 8.96E-01 X-POWEROP Plant Capacity Factor Seismic Initiating Event (0.4g to 2 6.08E-06 1.98E-05 %G07
<0.5g)
SEISMIC FRAGILITY FOR Bin %G07 seismic event causes a plant trip. There is a seismic failure of the SF-TB-CLASS 9.52E-01 %G07: Turbine Building Class 3 Class 3 Turbine Building (TB3) due to pounding-induced cracking, which leads to POUND-C-G07
- Pounding-induced cracking a LOOP and failure of AFW C. Liquefaction-induced settlement failure of the SF-TK-DG-FOSTRG- DFOST Liquefaction-Induced DFOST fails EDGs; thus, loss of all AC power. TB3 interacts with and fails TB1, 7.20E-01 TNK_SETTLE_G07 Settlement which fails the SDAFW pump. Loss of all injection/recirculation and SSHR leads 50% chance TB Class 3 to core damage.
5.00E-01 TB-CLASS3-1 interacts with TB Class 1 8.96E-01 X-POWEROP Plant Capacity Factor Seismic Initiating Event (0.3g to 3 5.77E-06 3.26E-05 %G05
<0.35g)
SEISMIC FRAGILITY FOR Bin %G05 seismic event causes a plant trip. There is a seismic failure of the SF-TB-CLASS 6.98E-01 %G05: Turbine Building Class 3 Class 3 Turbine Building (TB3) due to pounding-induced cracking, which leads to POUND-C-G05
- Pounding-induced cracking a LOOP and failure of AFW C. Liquefaction-induced settlement failure of the SF-TK-DG-FOSTRG- DFOST Liquefaction-Induced DFOST fails EDGs; thus, loss of all AC power. TB3 interacts with and fails TB1, 5.66E-01 TNK_SETTLE_G05 Settlement which fails the SDAFW pump. Loss of all injection/recirculation and SSHR leads 50% chance TB Class 3 to core damage.
5.00E-01 TB-CLASS3-1 interacts with TB Class 1 8.96E-01 X-POWEROP Plant Capacity Factor Seismic Initiating Event (0.3g to Bin %G05 seismic event causes a plant trip. There is a seismic failure of the 4 5.50E-06 3.26E-05 %G05
<0.35g) Class 3 Turbine Building (TB3) due to pounding-induced cracking, which leads to SEISMIC FRAGILITY FOR a LOOP and failure of AFW C. Liquefaction-induced settlement failure of the 6.36E-01 SF-TB_CRANE-C-G05 %G05: Turbine Building Gantry DFOST fails EDGs; thus, loss of all AC power. The TB Gantry Crane fails and Crane interacts with and fails the CST, failing the normal supply to the SDAFW; failure of Page 51 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table 5.4-1: Summary of Top SCDF Cutsets Cutset Event
- Event Description Cutset Description Prob Prob SEISMIC FRAGILITY FOR TB3 prevents alignment of the Condenser Water Box to the SDAFW pump SF-TB-CLASS 6.98E-01 %G05: Turbine Building Class 3 suction. Loss of all injection/recirculation and SSHR leads to core damage.
POUND-C-G05
- Pounding-induced cracking SF-TK-DG-FOSTRG- DFOST Liquefaction-Induced 5.66E-01 TNK_SETTLE_G05 Settlement TB Crane Failure Interacts with 7.50E-01 TBCRANE-CST CST Causing Failure 8.96E-01 X-POWEROP Plant Capacity Factor Seismic Initiating Event (0.35g 5 5.48E-06 1.94E-05 %G06 to <0.4g)
SEISMIC FRAGILITY FOR 7.88E-01 SF-TB_CRANE-C-G06 %G06: Turbine Building Gantry Bin %G06 seismic event causes a plant trip. There is a seismic failure of the Crane Class 3 Turbine Building (TB3) due to pounding-induced cracking, which leads to SEISMIC FRAGILITY FOR a LOOP and failure of AFW C. Liquefaction-induced settlement failure of the SF-TB-CLASS 8.48E-01 %G06: Turbine Building Class 3 DFOST fails EDGs; thus, loss of all AC power. The TB Gantry Crane fails and POUND-C-G06
- Pounding-induced cracking interacts with and fails the CST, failing the normal supply to the SDAFW; failure of SF-TK-DG-FOSTRG- DFOST Liquefaction-Induced TB3 prevents alignment of the Condenser Water Box to the SDAFW pump 6.30E-01 TNK_SETTLE_G06 Settlement suction. Loss of all injection/recirculation and SSHR leads to core damage.
TB Crane Failure Interacts with 7.50E-01 TBCRANE-CST CST Causing Failure 8.96E-01 X-POWEROP Plant Capacity Factor Seismic Initiating Event (0.4g to 6 5.20E-06 1.98E-05 %G07
<0.5g)
SEISMIC FRAGILITY FOR Bin %G07 seismic event causes a plant trip. There is a seismic failure of the SF-TB-CLASS 9.52E-01 %G07: Turbine Building Class 3 Class 3 Turbine Building (TB3) due to pounding-induced cracking, which leads to POUND-C-G07
- Pounding-induced cracking a LOOP and failure of AFW C. Liquefaction-induced settlement failure of the SF-TK-DG-FOSTRG- DFOST Liquefaction-Induced DFOST fails EDGs; thus, loss of all AC power. Liquefaction-induced settlement 7.20E-01 TNK_SETTLE_G07 Settlement failures of the SDAFW pump. Loss of all injection/recirculation and SSHR leads to SF-TP-SDAFW- SDAFW Liquefaction-Induced core damage.
4.28E-01 PMP_SETTLE_G07 Settlement 8.96E-01 X-POWEROP Plant Capacity Factor Seismic Initiating Event (0.25g 7 5.11E-06 5.86E-05 %G04 Bin %G04 seismic event causes a plant trip. There is a seismic failure of the to <0.3g)
Class 3 Turbine Building (TB3) due to pounding-induced cracking, which leads to SEISMIC FRAGILITY FOR SF-TB-CLASS a LOOP and failure of AFW C. Liquefaction-induced settlement failure of the 4.69E-01 %G04: Turbine Building Class 3 POUND-C-G04 DFOST fails EDGs; thus, loss of all AC power. TB3 interacts with and fails TB1,
- Pounding-induced cracking which fails the SDAFW pump. Loss of all injection/recirculation and SSHR leads SF-TK-DG-FOSTRG- DFOST Liquefaction-Induced 4.15E-01 to core damage.
TNK_SETTLE_G04 Settlement Page 52 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table 5.4-1: Summary of Top SCDF Cutsets Cutset Event
- Event Description Cutset Description Prob Prob 50% chance TB Class 3 5.00E-01 TB-CLASS3-1 interacts with TB Class 1 8.96E-01 X-POWEROP Plant Capacity Factor Seismic Initiating Event (0.6g to 8 4.89E-06 7.16E-06 %G09
<0.9g)
SEISMIC FRAGILITY FOR Bin %G09 seismic event causes a plant trip. There is a seismic failure of the SF-TB-CLASS 1.00E+00 %G09: Turbine Building Class 3 Class 3 Turbine Building (TB3) due to pounding-induced cracking, which leads to POUND-C-G09
- Pounding-induced cracking a LOOP and failure of AFW C. Liquefaction-induced settlement failure of the SEISMIC FRAGILITY FOR DFOST fails EDGs; thus, loss of all AC power. CST seismically fails leading to SF-TK-COND-STRG-7.82E-01 %G09: Condensate Storage failure of the normal supply to the SDAFW; failure of TB3 prevents alignment of TNK-C-G09 Tank the Condenser Water Box to the SDAFW pump suction. Loss of all SF-TK-DG-FOSTRG- DFOST Liquefaction-Induced injection/recirculation and SSHR leads to core damage.
9.76E-01 TNK_SETTLE_G09 Settlement 8.96E-01 X-POWEROP Plant Capacity Factor Page 53 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table 5.4-1: Summary of Top SCDF Cutsets Cutset Event
- Event Description Cutset Description Prob Prob Seismic Initiating Event (0.6g to 12 4.69E-06 7.16E-06 %G09
<0.9g)
SEISMIC FRAGILITY FOR 9.98E-01 SF-TB_CRANE-C-G09 %G09: Turbine Building Gantry Bin %G09 seismic event causes a plant trip. There is a seismic failure of the Crane Class 3 Turbine Building (TB3) due to pounding-induced cracking, which leads to SEISMIC FRAGILITY FOR a LOOP and failure of AFW C. Liquefaction-induced settlement failure of the SF-TB-CLASS 1.00E+00 %G09: Turbine Building Class 3 DFOST fails EDGs; thus, loss of all AC power. The TB Gantry Crane fails and POUND-C-G09
- Pounding-induced cracking interacts with and fails the CST, failing the normal supply to the SDAFW; failure of SF-TK-DG-FOSTRG- DFOST Liquefaction-Induced TB3 prevents alignment of the Condenser Water Box to the SDAFW pump 9.76E-01 TNK_SETTLE_G09 Settlement suction. Loss of all injection/recirculation and SSHR leads to core damage.
TB Crane Failure Interacts with 7.50E-01 TBCRANE-CST CST Causing Failure 8.96E-01 X-POWEROP Plant Capacity Factor Seismic Initiating Event (0.6g to 13 4.66E-06 7.16E-06 %G09
<0.9g)
Bin %G09 seismic event causes a plant trip. There is a seismic failure of the SEISMIC FRAGILITY FOR 9.52E-01 SF-TB-CLASS-3-C-G09 Class 3 Turbine Building (TB3), which leads to a LOOP and failure of AFW C.
%G09: Turbine Building Class 3 Liquefaction-induced settlement failure of the DFOST fails EDGs; thus, loss of all SEISMIC FRAGILITY FOR SF-TK-COND-STRG- AC power. CST seismically fails leading to failure of the normal supply to the 7.82E-01 %G09: Condensate Storage TNK-C-G09 SDAFW; failure of TB3 prevents alignment of the Condenser Water Box to the Tank SDAFW pump suction. Loss of all injection/recirculation and SSHR leads to core SF-TK-DG-FOSTRG- DFOST Liquefaction-Induced 9.76E-01 damage.
TNK_SETTLE_G09 Settlement 8.96E-01 X-POWEROP Plant Capacity Factor Seismic Initiating Event (0.5g to Bin %G08 seismic event causes a plant trip. There is a seismic failure of the 14 4.66E-06 8.47E-06 %G08
<0.6g) Class 3 Turbine Building (TB3) due to pounding-induced cracking, which leads to Page 54 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table 5.4-1: Summary of Top SCDF Cutsets Cutset Event
- Event Description Cutset Description Prob Prob SEISMIC FRAGILITY FOR a LOOP and failure of AFW C. Liquefaction-induced settlement failure of the 9.77E-01 SF-TB_CRANE-C-G08 %G08: Turbine Building Gantry DFOST fails EDGs; thus, loss of all AC power. The TB Gantry Crane fails and Crane interacts with and fails the CST failing the normal supply to the SDAFW; failure of SEISMIC FRAGILITY FOR TB3 prevents alignment of the Condenser Water Box to the SDAFW pump SF-TB-CLASS 9.91E-01 %G08: Turbine Building Class 3 suction. Loss of all injection/recirculation and SSHR leads to core damage.
POUND-C-G08
- Pounding-induced cracking SF-TK-DG-FOSTRG- DFOST Liquefaction-Induced 8.45E-01 TNK_SETTLE_G08 Settlement TB Crane Failure Interacts with 7.50E-01 TBCRANE-CST CST Causing Failure 8.96E-01 X-POWEROP Plant Capacity Factor Seismic Initiating Event (0.35g 15 4.64E-06 1.94E-05 %G06 to <0.4g)
SEISMIC FRAGILITY FOR Bin %G06 seismic event causes a plant trip. There is a seismic failure of the SF-TB-CLASS 8.48E-01 %G06: Turbine Building Class 3 Class 3 Turbine Building (TB3) due to pounding-induced cracking, which leads to POUND-C-G06
- Pounding-induced cracking a LOOP and failure of AFW C. Liquefaction-induced settlement failure of the SF-TK-DG-FOSTRG- DFOST Liquefaction-Induced DFOST fails EDGs; thus, loss of all AC power. TB3 interacts with and fails TB1, 6.30E-01 TNK_SETTLE_G06 Settlement which fails the SDAFW pump. Loss of all injection/recirculation and SSHR leads 50% chance TB Class 3 to core damage.
5.00E-01 TB-CLASS3-1 interacts with TB Class 1 8.96E-01 X-POWEROP Plant Capacity Factor Page 55 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table 5.4-1: Summary of Top SCDF Cutsets Cutset Event
- Event Description Cutset Description Prob Prob Seismic Initiating Event (0.6g to 19 4.51E-06 7.16E-06 %G09
<0.9g)
SEISMIC FRAGILITY FOR Bin %G09 seismic event causes a plant trip. There is a seismic failure of the SF-TB-CLASS 1.00E+00 %G09: Turbine Building Class 3 Class 3 Turbine Building (TB3) due to pounding-induced cracking, which leads to POUND-C-G09
- Pounding-induced cracking a LOOP and failure of AFW C. Liquefaction-induced settlement failure of the SF-TK-DG-FOSTRG- DFOST Liquefaction-Induced DFOST fails EDGs; thus, loss of all AC power. Liquefaction-induced settlement 9.76E-01 TNK_SETTLE_G09 Settlement failures of the SDAFW pump. Loss of all injection/recirculation and SSHR leads to SF-TP-SDAFW- SDAFW Liquefaction-Induced core damage.
7.21E-01 PMP_SETTLE_G09 Settlement 8.96E-01 X-POWEROP Plant Capacity Factor Seismic Initiating Event (0.6g to 20 4.46E-06 7.16E-06 %G09
<0.9g)
SEISMIC FRAGILITY FOR 9.98E-01 SF-TB_CRANE-C-G09 %G09: Turbine Building Gantry Bin %G09 seismic event causes a plant trip. There is a seismic failure of the Crane Class 3 Turbine Building (TB3), which leads to a LOOP and failure of AFW C.
Liquefaction-induced settlement failure of the DFOST fails EDGs; thus, loss of all SEISMIC FRAGILITY FOR 9.52E-01 SF-TB-CLASS-3-C-G09 AC power. The TB Gantry Crane fails and interacts with and fails the CST failing
%G09: Turbine Building Class 3 the normal supply to the SDAFW; failure of TB3 prevents alignment of the SF-TK-DG-FOSTRG- DFOST Liquefaction-Induced 9.76E-01 Condenser Water Box to the SDAFW pump suction. Loss of all TNK_SETTLE_G09 Settlement injection/recirculation and SSHR leads to core damage.
TB Crane Failure Interacts with 7.50E-01 TBCRANE-CST CST Causing Failure 8.96E-01 X-POWEROP Plant Capacity Factor Page 56 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table 5.4-1: Summary of Top SCDF Cutsets Cutset Event
- Event Description Cutset Description Prob Prob Seismic Initiating Event (0.6g to 22 4.30E-06 7.16E-06 %G09
<0.9g)
SEISMIC FRAGILITY FOR Bin %G09 seismic event causes a plant trip. There is a seismic failure of the 9.52E-01 SF-TB-CLASS-3-C-G09
%G09: Turbine Building Class 3 Class 3 Turbine Building (TB3), which leads to a LOOP and failure of AFW C.
SF-TK-DG-FOSTRG- DFOST Liquefaction-Induced Liquefaction-induced settlement failure of the DFOST fails EDGs; thus, loss of all 9.76E-01 TNK_SETTLE_G09 Settlement AC power. Liquefaction-induced settlement failures of the SDAFW pump. Loss of SF-TP-SDAFW- SDAFW Liquefaction-Induced all injection/recirculation and SSHR leads to core damage.
7.21E-01 PMP_SETTLE_G09 Settlement 8.96E-01 X-POWEROP Plant Capacity Factor Page 57 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Page 58 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table 5.4-2: Top 10 SCDF Importance Measures Ranked by FV - Seismic Failures Median Fragility Component Description FV r u Failure Mode2 Capacity (g) Method2 Turbine Building RAB pounding induced cracking and SF-TB-CLASS Class 3 - Pounding- 43.71% 0.28 0.13 0.25 splitting of the mezzanine floor slab SOV POUND induced cracking resulting in loss of structural integrity.
Failure of EDG-B pipe at RAB penetration.
Probability of failure at PGA:
SF-TK-DG-FOSTRG- DFOST Liquefaction-12.57% N/A1 N/A N/A PGA(g) - Probability SOV TNK_SETTLE Induced Settlement 0.265g - 0.388 0.325g - 0.569 0.582g - 0.887 0.717g - 0.966 Turbine Building SF-TB_CRANE 11.81% 0.29 0.21 0.24 Failure of A-frame anchor bolts SOV Gantry Crane Failure of AFW discharge piping where the 4 in. diameter pipe meets the 6x4 reducing elbow.
SF-TP-SDAFW- SDAFW Liquefaction- Probability of failure at PGA:
7.01% N/A1 N/A N/A SOV PMP_SETTLE Induced Settlement PGA(g) - Probability 0.265g - 0.168 0.325g - 0.263 0.582g - 0.609 0.717g - 0.708 Page 59 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table 5.4-2: Top 10 SCDF Importance Measures Ranked by FV - Seismic Failures Median Fragility Component Description FV r u Failure Mode2 Capacity (g) Method2 Liquefaction-induced settlement failure at Intake Structure.
Underground Cable Probability of failure at PGA:
SF-TR-UG- Trays at Intake 2.59% N/A1 N/A N/A PGA(g) - Probability Rep.
INTAKE_SETTLE Liquefaction-Induced 0.265g - 0.110 Settlement 0.325g - 0.180 0.582g - 0.450 0.717g - 0.570 Relay Chatter - DG-A,B-AUX-Relay malfunction due to earthquake SF-RC-20 PNL_Barksdale 1.87% 0.95 0.27 0.9 SOV shaking Controls_D2T-M18SS Relay Chatter - DG-A.B-ENG-Relay malfunction due to earthquake SF-RC-21 PNL_Barksdale 1.87% 0.95 0.27 0.9 SOV shaking Controls_D2T-M80SS North Header - Mechanism 3.
Probability of failure at PGA:
North Header Intake SF-PIP-UG- PGA(g) - Probability Mech 3 Liquefaction- 1.70% N/A1 N/A N/A SOV N_INTAKE3_SETTLE 0.265g - 0.223 Induced Settlement 0.325g - 0.390 0.582g - 0.722 0.717g - 0.824 Seismic-Induced SF-SLOSP 1.54% 0.3 0.3 0.45 SPRAIG values for offsite power. Rep.
Loss of Offsite Power Note 1: Liquefaction-Induced Settlement failures do not utilize median capacity or -values. The probability of failure at a given PGA is provided in the Failure Mode column.
Note 2: Fragility Mode and Fragility Method per [82].
Page 60 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT The most significant non-seismic SSC failures (e.g., random failures of modeled components during the SPRA mission time) are listed in Table 5.4-3. A summary of the SCDF results for each seismic hazard interval is presented in Table 5.4-4.
Table5.4-3: Top 10 SCDF Importance Measures Ranked by FV - Non-Seismic Failures Component Description and Failure Mode FV FPT1XSABFR TURBINE-DRIVEN PUMP FAILS TO RUN 0.73%
Failure to align and start pre-staged pumps OPER-61 for SG makeup - Condenser Inlet Waterbox 0.68%
(FLEX)
CONDENSER WATERBOX INLET MOTOR FPMCIW-LFS 0.57%
PUMP CIW-L FAILS TO START (FLEX)
FPT1XSABFS TURBINE-DRIVEN PUMP FAILS TO START 0.43%
Failure to align and start portable pumps to OPER-64 lake for long-term water source - SG makeup 0.42%
(FLEX)
Dependent HEP for OPER-35,OPER-XOPERC-1 0.33%
68,OPER-18B-S1,OPER-64,OPER-01S OPERATOR FAILS TO TRANSFER POWER OPER-14 0.33%
TO DEEPWELL PUMP DIESEL FTMSDPTRXM AFW TD PUMP TRAIN C UNAVAILABLE 0.30%
PORTABLE LAKE PUMP FAILS TO START FPMPLAKEFS 0.26%
(FLEX)
OPER-13S OPERATOR FAILS TO ALIGN AFW PUMP C 0.18%
Page 61 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT A summary of the SCDF results for each seismic hazard interval is presented in Table 5.4-4.
Table 5.4-4 Contribution to SCDF by Acceleration Interval Hazard Interval Description SCDF % of Total Cumulative SCDF CDF Hazard Curve: HAZARD - PGA Range: 0.1g to 0.15g 5.17E-07 1% 5.17E-07 Hazard Curve: HAZARD - PGA Range: 0.15g to 0.2g 2.13E-06 2% 2.64E-06 Hazard Curve: HAZARD - PGA Range: 0.2g to 0.25g 9.14E-06 10% 1.18E-05 Hazard Curve: HAZARD - PGA Range: 0.25g to 0.3g 1.66E-05 18% 2.84E-05 Hazard Curve: HAZARD - PGA Range: 0.3g to 0.35g 1.77E-05 19% 4.61E-05 Hazard Curve: HAZARD - PGA Range: 0.35g to 0.4g 1.41E-05 15% 6.02E-05 Hazard Curve: HAZARD - PGA Range: 0.4g to 0.5g 1.69E-05 18% 7.72E-05 Hazard Curve: HAZARD - PGA Range: 0.5g to 0.6g 7.54E-06 8% 8.46E-05 Hazard Curve: HAZARD - PGA Range: 0.6g to 0.9g 6.42E-06 7% 9.10E-05 Hazard Curve: HAZARD - PGA Range: > 0.9g 1.68E-06 2% 9.27E-05 5.5 SLERF Results The seismic PRA performed for Robinson Nuclear Power Plant shows that the point estimate mean seismic LERF is 2.02E-05/reactor-year. A discussion of the mean SLERF with uncertainty distribution reflecting the uncertainties in the hazard, fragilities, and model data is presented in Section 5.6. Important contributors are discussed in the following paragraphs.
The top SLERF cutsets are documented in the SPRA quantification report [13]. These are briefly summarized in Table 5.5-1.
SSCs with the most significant seismic failure contribution to SLERF are listed in Table 5.5-2, sorted by FV. The seismic fragilities for each of the significant contributors is also provided in Table 5.5-2, along with the corresponding limiting seismic failure mode and method of fragility calculation.
The top ten events for the SLERF are described here. As can be seen from Table 5.5-2, the RNP SLERF is dominated (29.59%) by the pounding-induced cracking failure of the Class III Turbine Building (TB3); this is a loss of structural integrity due to the Turbine Building Mezzanine floor pounding into the RAB. Failure of TB3 leads to a LOOP and failure of AFW C, as well as a high chance of failure of the Class I Turbine Building (TB1),
which houses the piping required to provide flow to the steam generators and piping from the CST. The second highest contributor (11.85%) is the failure of the Reactor Containment Building, which is assumed to lead directly to LERF. Contributor #3 (8.03%)
is the liquefaction-induced settlement failure of the DFOST piping leading to failure of the EDGs. Contributor #4 (7.67%) is the failure of the TB gantry crane which has a high probability of interacting with and failing the CST. Contributor #5 (5.02%) is the liquefaction-induced lateral spreading of Displacement Category 2, which includes the Reactor Auxiliary Building; and thus, is assumed to lead directly to LERF. Contributor #6 Page 62 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Contributor #7 (4.19%) is the liquefaction-induced settlement failure of the SDAFW pump piping. Contributor #8 (2.74%) is the failure of the Reactor Auxiliary Building, which is assumed to lead directly to LERF. Contributor #9 (2.45%) are two separate relay chatter groups which cause a loss of the EDGs. Contributor #9 (2.45%) is the liquefaction-induced settlement failure of underground cable trays running from the Intake Structure to the RAB, which contain cables for the SW system and is assumed to fail SW. Contributor #10 (2.07%) is the liquefaction-induced failure of Deepwell D piping.
Page 63 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table 5.5-1: Summary of Top SLERF Cutsets Cutset Event
- Event Description Cutset Description Prob Prob Seismic Initiating Event (0.6g to 1 2.12E-06 7.16E-06 %G09
<0.9g)
SEISMIC FRAGILITY FOR
%G09: REACTOR Bin %G09 seismic event causes a plant trip. There is a seismic failure of the 3.31E-01 SF-RCB-C-G09 RCB, which is assumed to lead directly to core damage and LERF.
CONTAINMENT BUILDING (RCB) 8.96E-01 X-POWEROP Plant Capacity Factor 2 1.14E-06 1.88E-06 %G10 Seismic Initiating Event (>0.9g)
SEISMIC FRAGILITY FOR
%G10: REACTOR Bin %G10 seismic event causes a plant trip. There is a seismic failure of the 6.77E-01 SF-RCB-C-G10 CONTAINMENT BUILDING RCB, which is assumed to lead directly to core damage and LERF.
(RCB) 8.96E-01 X-POWEROP Plant Capacity Factor Seismic Initiating Event (0.4g to 3 8.22E-07 1.98E-05 %G07
<0.5g)
SEISMIC FRAGILITY FOR 9.13E-01 SF-TB_CRANE-C-G07 %G07: Turbine Building Gantry Bin %G07 seismic event causes a plant trip. There is a seismic failure of the Crane Class 3 Turbine Building (TB3) due to pounding-induced cracking, which leads SEISMIC FRAGILITY FOR to a LOOP and failure of AFW C. Liquefaction-induced settlement failure of the SF-TB-CLASS-3-POUND-9.52E-01 %G07: Turbine Building Class 3 DFOST fails EDGs; thus, loss of all AC power. The TB Gantry Crane fails and C-G07
- Pounding-induced cracking interacts with and fails the CST failing the normal supply to the SDAFW; failure SF-TK-DG-FOSTRG- DFOST Liquefaction-Induced of TB3 prevents alignment of the Condenser Water Box to the SDAFW pump 7.20E-01 TNK_SETTLE_G07 Settlement suction. Loss of all injection/recirculation and SSHR leads to core damage. Loss TB Crane Failure Interacts with of power leads to loss of containment coolers and containment spray injection, 7.50E-01 TBCRANE-CST which leads to LERF.
CST causing Failure 9.88E-02 XFL_PDS3P Plant Damage State 3P 8.96E-01 X-POWEROP Plant Capacity Factor Seismic Initiating Event (0.6g to 4 7.49E-07 7.16E-06 %G09
<0.9g)
SEISMIC FRAGILITY FOR Bin %G09 seismic event causes a plant trip. There is a seismic failure of the 1.17E-01 SF-RAB-C-G09 %G09: Reactor Auxiliary RAB, which is assumed to lead directly to core damage and LERF.
Building 8.96E-01 X-POWEROP Plant Capacity Factor Seismic Initiating Event (0.5g to Bin %G08 seismic event causes a plant trip. There is a seismic failure of the 5 7.09E-07 8.47E-06 %G08
<0.6g) RCB, which is assumed to lead directly to core damage and LERF.
Page 64 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table 5.5-1: Summary of Top SLERF Cutsets Cutset Event
- Event Description Cutset Description Prob Prob SEISMIC FRAGILITY FOR
%G08: REACTOR 9.34E-02 SF-RCB-C-G08 CONTAINMENT BUILDING (RCB) 8.96E-01 X-POWEROP Plant Capacity Factor 6 6.11E-07 1.88E-06 %G10 Seismic Initiating Event (>0.9g)
SEISMIC FRAGILITY FOR Bin %G10 seismic event causes a plant trip. There is a seismic failure of the 3.64E-01 SF-RAB-C-G10 %G10: Reactor Auxiliary RAB, which is assumed to lead directly to core damage and LERF.
Building 8.96E-01 X-POWEROP Plant Capacity Factor Seismic Initiating Event (0.4g to 7 6.01E-07 1.98E-05 %G07
<0.5g)
SEISMIC FRAGILITY FOR SF-TB-CLASS-3-POUND- Bin %G07 seismic event causes a plant trip. There is a seismic failure of the 9.52E-01 %G07: Turbine Building Class 3 C-G07 Class 3 Turbine Building (TB3) due to pounding-induced cracking, which leads
- Pounding-induced cracking to a LOOP and failure of AFW C. Liquefaction-induced settlement failure of the SF-TK-DG-FOSTRG- DFOST Liquefaction-Induced 7.20E-01 DFOST fails EDGs; thus, loss of all AC power. TB3 interacts with and fails TB1, TNK_SETTLE_G07 Settlement which fails the SDAFW pump. Loss of all injection/recirculation and SSHR leads 50% chance TB Class 3 5.00E-01 TB-CLASS3-1 to core damage. Loss of power leads to loss of containment coolers and interacts with TB Class 1 containment spray injection, which leads to LERF.
9.88E-02 XFL_PDS3P Plant Damage State 3P 8.96E-01 X-POWEROP Plant Capacity Factor Seismic Initiating Event (0.3g to 8 5.70E-07 3.26E-05 %G05
<0.35g)
SEISMIC FRAGILITY FOR SF-TB-CLASS-3-POUND- Bin %G05 seismic event causes a plant trip. There is a seismic failure of the 6.98E-01 %G05: Turbine Building Class 3 C-G05 Class 3 Turbine Building (TB3) due to pounding-induced cracking, which leads
- Pounding-induced cracking to a LOOP and failure of AFW C. Liquefaction-induced settlement failure of the SF-TK-DG-FOSTRG- DFOST Liquefaction-Induced 5.66E-01 DFOST fails EDGs; thus, loss of all AC power. TB3 interacts with and fails TB1, TNK_SETTLE_G05 Settlement which fails the SDAFW pump. Loss of all injection/recirculation and SSHR leads 50% chance TB Class 3 5.00E-01 TB-CLASS3-1 to core damage. Loss of power leads to loss of containment coolers and interacts with TB Class 1 containment spray injection, which leads to LERF.
9.88E-02 XFL_PDS3P Plant Damage State 3P 8.96E-01 X-POWEROP Plant Capacity Factor Seismic Initiating Event (0.3g to Bin %G05 seismic event causes a plant trip. There is a seismic failure of the 9 5.44E-07 3.26E-05 %G05
<0.35g) Class 3 Turbine Building (TB3) due to pounding-induced cracking, which leads Page 65 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table 5.5-1: Summary of Top SLERF Cutsets Cutset Event
- Event Description Cutset Description Prob Prob SEISMIC FRAGILITY FOR to a LOOP and failure of AFW C. Liquefaction-induced settlement failure of the 6.36E-01 SF-TB_CRANE-C-G05 %G05: Turbine Building Gantry DFOST fails EDGs; thus, loss of all AC power. The TB Gantry Crane fails and Crane interacts with and fails the CST, failing the normal supply to the SDAFW; failure SEISMIC FRAGILITY FOR of TB3 prevents alignment of the Condenser Water Box to the SDAFW pump SF-TB-CLASS-3-POUND-6.98E-01 %G05: Turbine Building Class 3 suction. Loss of all injection/recirculation and SSHR leads to core damage. Loss C-G05
- Pounding-induced cracking of power leads to loss of containment coolers and containment spray injection, SF-TK-DG-FOSTRG- DFOST Liquefaction-Induced which leads to LERF.
5.66E-01 TNK_SETTLE_G05 Settlement TB Crane Failure Interacts with 7.50E-01 TBCRANE-CST CST causing Failure 9.88E-02 XFL_PDS3P Plant Damage State 3P 8.96E-01 X-POWEROP Plant Capacity Factor Seismic Initiating Event (0.35g 10 5.42E-07 1.94E-05 %G06 to <0.4g)
SEISMIC FRAGILITY FOR 7.88E-01 SF-TB_CRANE-C-G06 %G06: Turbine Building Gantry Bin %G06 seismic event causes a plant trip. There is a seismic failure of the Crane Class 3 Turbine Building (TB3) due to pounding-induced cracking, which leads SEISMIC FRAGILITY FOR to a LOOP and failure of AFW C. Liquefaction-induced settlement failure of the SF-TB-CLASS-3-POUND-8.48E-01 %G06: Turbine Building Class 3 DFOST fails EDGs; thus, loss of all AC power. The TB Gantry Crane fails and C-G06
- Pounding-induced cracking interacts with and fails the CST, failing the normal supply to the SDAFW; failure SF-TK-DG-FOSTRG- DFOST Liquefaction-Induced of TB3 prevents alignment of the Condenser Water Box to the SDAFW pump 6.30E-01 TNK_SETTLE_G06 Settlement suction. Loss of all injection/recirculation and SSHR leads to core damage. Loss TB Crane Failure Interacts with of power leads to loss of containment coolers and containment spray injection, 7.50E-01 TBCRANE-CST which leads to LERF.
CST causing Failure 9.88E-02 XFL_PDS3P Plant Damage State 3P 8.96E-01 X-POWEROP Plant Capacity Factor Seismic Initiating Event (0.4g to 11 5.14E-07 1.98E-05 %G07
<0.5g)
SEISMIC FRAGILITY FOR SF-TB-CLASS-3-POUND- Bin %G07 seismic event causes a plant trip. There is a seismic failure of the 9.52E-01 %G07: Turbine Building Class 3 C-G07 Class 3 Turbine Building (TB3) due to pounding-induced cracking, which leads
- Pounding-induced cracking to a LOOP and failure of AFW C. Liquefaction-induced settlement failure of the SF-TK-DG-FOSTRG- DFOST Liquefaction-Induced 7.20E-01 DFOST fails EDGs; thus, loss of all AC power. Liquefaction-induced settlement TNK_SETTLE_G07 Settlement failures of the SDAFW pump. Loss of all injection/recirculation and SSHR leads SF-TP-SDAFW- SDAFW Liquefaction-Induced 4.28E-01 to core damage. Loss of power leads to loss of containment coolers and PMP_SETTLE_G07 Settlement containment spray injection, which leads to LERF.
9.88E-02 XFL_PDS3P Plant Damage State 3P 8.96E-01 X-POWEROP Plant Capacity Factor Page 66 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table 5.5-1: Summary of Top SLERF Cutsets Cutset Event
- Event Description Cutset Description Prob Prob Seismic Initiating Event (0.25g 12 5.05E-07 5.86E-05 %G04 to <0.3g)
SEISMIC FRAGILITY FOR SF-TB-CLASS-3-POUND- Bin %G04 seismic event causes a plant trip. There is a seismic failure of the 4.69E-01 %G04: Turbine Building Class 3 C-G04 Class 3 Turbine Building (TB3) due to pounding-induced cracking, which leads
- Pounding-induced cracking to a LOOP and failure of AFW C. Liquefaction-induced settlement failure of the SF-TK-DG-FOSTRG- DFOST Liquefaction-Induced 4.15E-01 DFOST fails EDGs; thus, loss of all AC power. TB3 interacts with and fails TB1, TNK_SETTLE_G04 Settlement which fails the SDAFW pump. Loss of all injection/recirculation and SSHR leads 50% chance TB Class 3 5.00E-01 TB-CLASS3-1 to core damage. Loss of power leads to loss of containment coolers and interacts with TB Class 1 containment spray injection, which leads to LERF.
9.88E-02 XFL_PDS3P Plant Damage State 3P 8.96E-01 X-POWEROP Plant Capacity Factor Seismic Initiating Event (0.6g to 13 4.83E-07 7.16E-06 %G09
<0.9g)
SEISMIC FRAGILITY FOR Bin %G09 seismic event causes a plant trip. There is a seismic failure of the SF-TB-CLASS-3-POUND-1.00E+00 %G09: Turbine Building Class 3 Class 3 Turbine Building (TB3) due to pounding-induced cracking, which leads C-G09
- Pounding-induced cracking to a LOOP and failure of AFW C. Liquefaction-induced settlement failure of the SEISMIC FRAGILITY FOR DFOST fails EDGs; thus, loss of all AC power. CST seismically fails, leading to SF-TK-COND-STRG-TNK-7.82E-01 %G09: Condensate Storage failure of the normal supply to the SDAFW; failure of TB3 prevents alignment of C-G09 Tank the Condenser Water Box to the SDAFW pump suction. Loss of all SF-TK-DG-FOSTRG- DFOST Liquefaction-Induced injection/recirculation and SSHR leads to core damage. Loss of power leads to 9.76E-01 TNK_SETTLE_G09 Settlement loss of containment coolers and containment spray injection, which leads to 9.88E-02 XFL_PDS3P Plant Damage State 3P LERF.
8.96E-01 X-POWEROP Plant Capacity Factor Seismic Initiating Event (0.4g to 14 4.77E-07 1.98E-05 %G07
<0.5g)
SEISMIC FRAGILITY FOR
%G07: REACTOR Bin %G07 seismic event causes a plant trip. There is a seismic failure of the 2.69E-02 SF-RCB-C-G07 RCB, which is assumed to lead directly to core damage and LERF.
CONTAINMENT BUILDING (RCB) 8.96E-01 X-POWEROP Plant Capacity Factor Page 67 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table 5.5-1: Summary of Top SLERF Cutsets Cutset Event
- Event Description Cutset Description Prob Prob Page 68 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table 5.5-1: Summary of Top SLERF Cutsets Cutset Event
- Event Description Cutset Description Prob Prob Seismic Initiating Event (0.6g to Bin %G09 seismic event causes a plant trip. There is a seismic failure of the 20 4.63E-07 7.16E-06 %G09
<0.9g) Class 3 Turbine Building (TB3) due to pounding-induced cracking, which leads Page 69 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table 5.5-1: Summary of Top SLERF Cutsets Cutset Event
- Event Description Cutset Description Prob Prob SEISMIC FRAGILITY FOR to a LOOP and failure of AFW C. Liquefaction-induced settlement failure of the 9.98E-01 SF-TB_CRANE-C-G09 %G09: Turbine Building Gantry DFOST fails EDGs; thus, loss of all AC power. The TB Gantry Crane fails and Crane interacts with and fails the CST failing the normal supply to the SDAFW; failure SEISMIC FRAGILITY FOR of TB3 prevents alignment of the Condenser Water Box to the SDAFW pump SF-TB-CLASS-3-POUND-1.00E+00 %G09: Turbine Building Class 3 suction. Loss of all injection/recirculation and SSHR leads to core damage. Loss C-G09
- Pounding-induced cracking of power leads to loss of containment coolers and containment spray injection, SF-TK-DG-FOSTRG- DFOST Liquefaction-Induced which leads to LERF.
9.76E-01 TNK_SETTLE_G09 Settlement TB Crane Failure Interacts with 7.50E-01 TBCRANE-CST CST causing Failure 9.88E-02 XFL_PDS3P Plant Damage State 3P 8.96E-01 X-POWEROP Plant Capacity Factor Seismic Initiating Event (0.6g to 22 4.60E-07 7.16E-06 %G09
<0.9g) Bin %G09 seismic event causes a plant trip. There is a seismic failure of the SEISMIC FRAGILITY FOR Class 3 Turbine Building (TB3), which leads to a LOOP and failure of AFW C.
9.52E-01 SF-TB-CLASS-3-C-G09
%G09: Turbine Building Class 3 Liquefaction-induced settlement failure of the DFOST fails EDGs; thus, loss of all SEISMIC FRAGILITY FOR AC power. CST seismically fails, leading to failure of the normal supply to the SF-TK-COND-STRG-TNK-7.82E-01 %G09: Condensate Storage SDAFW; failure of TB3 prevents alignment of the Condenser Water Box to the C-G09 Tank SDAFW pump suction. Loss of all injection/recirculation and SSHR leads to core SF-TK-DG-FOSTRG- DFOST Liquefaction-Induced damage. Loss of power leads to loss of containment coolers and containment 9.76E-01 TNK_SETTLE_G09 Settlement spray injection, which leads to LERF.
9.88E-02 XFL_PDS3P Plant Damage State 3P Page 70 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table 5.5-1: Summary of Top SLERF Cutsets Cutset Event
- Event Description Cutset Description Prob Prob 8.96E-01 X-POWEROP Plant Capacity Factor Seismic Initiating Event (0.5g to 23 4.60E-07 8.47E-06 %G08
<0.6g)
SEISMIC FRAGILITY FOR 9.77E-01 SF-TB_CRANE-C-G08 %G08: Turbine Building Gantry Bin %G08 seismic event causes a plant trip. There is a seismic failure of the Crane Class 3 Turbine Building (TB3) due to pounding-induced cracking, which leads SEISMIC FRAGILITY FOR to a LOOP and failure of AFW C. Liquefaction-induced settlement failure of the SF-TB-CLASS-3-POUND-9.91E-01 %G08: Turbine Building Class 3 DFOST fails EDGs; thus, loss of all AC power. The TB Gantry Crane fails and C-G08
- Pounding-induced cracking interacts with and fails the CST failing the normal supply to the SDAFW; failure SF-TK-DG-FOSTRG- DFOST Liquefaction-Induced of TB3 prevents alignment of the Condenser Water Box to the SDAFW pump 8.45E-01 TNK_SETTLE_G08 Settlement suction. Loss of all injection/recirculation and SSHR leads to core damage. Loss TB Crane Failure Interacts with of power leads to loss of containment coolers and containment spray injection, 7.50E-01 TBCRANE-CST which leads to LERF.
CST causing Failure 9.88E-02 XFL_PDS3P Plant Damage State 3P 8.96E-01 X-POWEROP Plant Capacity Factor Seismic Initiating Event (0.35g Bin %G06 seismic event causes a plant trip. There is a seismic failure of the 25 4.59E-07 1.94E-05 %G06 to <0.4g) Class 3 Turbine Building (TB3) due to pounding-induced cracking, which leads SEISMIC FRAGILITY FOR to a LOOP and failure of AFW C. Liquefaction-induced settlement failure of the SF-TB-CLASS-3-POUND-8.48E-01 %G06: Turbine Building Class 3 DFOST fails EDGs; thus, loss of all AC power. TB3 interacts with and fails TB1, C-G06
- Pounding-induced cracking which fails the SDAFW pump. Loss of all injection/recirculation and SSHR leads SF-TK-DG-FOSTRG- DFOST Liquefaction-Induced to core damage. Loss of power leads to loss of containment coolers and 6.30E-01 TNK_SETTLE_G06 Settlement containment spray injection, which leads to LERF.
Page 71 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table 5.5-1: Summary of Top SLERF Cutsets Cutset Event
- Event Description Cutset Description Prob Prob 50% chance TB Class 3 5.00E-01 TB-CLASS3-1 interacts with TB Class 1 9.88E-02 XFL_PDS3P Plant Damage State 3P 8.96E-01 X-POWEROP Plant Capacity Factor Page 72 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table 5.5-2: Top 10 SLERF Importance Measures Ranked by FV - Seismic Failures Median Fragility Component Description FV r u Failure Mode2 Capacity (g) Method2 Turbine Building RAB pounding induced cracking and SF-TB-CLASS-3-POUND Class 3 - Pounding- 29.59% 0.28 0.13 0.25 splitting of the mezzanine floor slab SOV induced cracking resulting in loss of structural integrity.
Nonlinear rotation of the socketed like REACTOR connection at the underside of the SF-RCB CONTAINMENT 11.85% 0.85 0.18 0.28 SOV basemat due to lateral displacement of the BUILDING (RCB) structure Failure of EDG-B pipe at RAB penetration.
Probability of failure at PGA:
DFOST SF-TK-DG-FOSTRG- PGA(g) - Probability Liquefaction- 8.03% N/A1 N/A N/A SOV TNK_SETTLE 0.265g - 0.388 Induced Settlement 0.325g - 0.569 0.582g - 0.887 0.717g - 0.966 Turbine Building SF-TB_CRANE 7.67% 0.29 0.21 0.24 Failure of A-frame anchor bolts SOV Gantry Crane Liquefaction-Induced Lateral Spreading.
Liquefaction- Probability of failure at PGA:
Induced Lateral PGA(g) - Probability SEISMIC_SPREAD_DC2 5.02% N/A1 N/A N/A Rep.
Spreading Distance 0.265g - 0.000 Category 2 0.325g - 0.004 0.582g - 0.043 0.717g - 0.061 Page 73 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table 5.5-2: Top 10 SLERF Importance Measures Ranked by FV - Seismic Failures Median Fragility Component Description FV r u Failure Mode2 Capacity (g) Method2 SOV Failure of AFW discharge piping where the 4 in. diameter pipe meets the 6x4 reducing elbow.
SDAFW SF-TP-SDAFW- Probability of failure at PGA:
Liquefaction- 4.19% N/A1 N/A N/A SOV PMP_SETTLE PGA(g) - Probability Induced Settlement 0.265g - 0.168 0.325g - 0.263 0.582g - 0.609 0.717g - 0.708 Reactor Auxiliary SF-RAB 2.74% 1.12 0.24 0.26 Flexural failure of piles. Rep.
Building Liquefaction-induced settlement failure at Intake Structure.
Underground Cable Probability of failure at PGA:
SF-TR-UG- Trays at Intake 2.45% N/A1 N/A N/A Rep.
INTAKE_SETTLE Liquefaction- PGA(g) - Probability Induced Settlement 0.265g - 0.110 0.325g - 0.180 0.582g - 0.450 0.717g - 0.570 Page 74 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table 5.5-2: Top 10 SLERF Importance Measures Ranked by FV - Seismic Failures Median Fragility Component Description FV r u Failure Mode2 Capacity (g) Method2 Bending failure of a bolted flange connection due to liquefaction-induced settlement.
Deepwell Pump D Probability of failure at PGA:
SF-WP-DPW-PMP-Liquefaction- 2.07% N/A1 N/A N/A SOV D_SETTLE Induced Settlement PGA(g) - Probability 0.265g - 0.546 0.325g - 0.732 0.582g - 0.952 0.717g - 0.968 Note 1: Liquefaction-Induced Settlement failures do not utilize median capacity or -values. The probability of failure at a given PGA is provided in the Failure Mode column.
Note 2: Fragility Mode and Fragility Method per [82].
Page 75 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT The most significant non-seismic SSC SLERF contributors (e.g., random failures of modeled components during the SPRA mission time) are listed in Table 5.5-3.
Table 5.5-3: Top 10 SLERF Importance Measures Ranked by FV - Non-Seismic Failures Component Description and Failure Mode FV PERSONNEL HATCH INNER DOOR GINRDOORSL 1.09%
GASKETS FAILS FPT1XSABFR TURBINE-DRIVEN PUMP FAILS TO RUN 0.32%
FAILURE OF PERSONNEL HATCH DOOR GDOORSEALS 0.32%
SEALS GELPENFO ELECTRICAL PENETRATIONS FAILS OPEN 0.32%
Failure to align and start pre-staged pumps OPER-61 for SG makeup - Condenser Inlet Waterbox 0.30%
(FLEX)
PRE-INITIATOR IMPORTANCE SCOPING GOPER-PRE3 EVENT FOR CI - P-44/45 BYPASS LEFT 0.29%
OPEN FPT1XSABFS TURBINE-DRIVEN PUMP FAILS TO START 0.18%
Failure to align and start portable pumps to OPER-64 lake for long-term water source - SG makeup 0.16%
(FLEX)
OPER-18B-CST Failure to supply AFW with SW 0.15%
PRE-INITIATOR IMPORTANCE SCOPING GOPER-PRE4 EVENT FOR CI- PERSN HATCHES LEFT 0.14%
OPEN Page 76 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT 5.6 SPRA Quantification Uncertainty Analysis A parametric uncertainty propagation for the SPRA base seismic CDF and LERF was performed.
Probability distribution types and associated variance parameters (e.g., error factors (EFs) in the case of lognormal distributions) are assigned to each of the basic events in the PRA. This is a basic step in the PRA development process and much of this distribution information already exists in the PRA database used for the SPRA as the SPRA database is built upon the internal events PRA database. Distribution information had to be added for modeling elements for SPRA that do not already exist in the internal events-based PRA database; these include seismic hazard interval initiators, seismic fragility basic events and seismic-adjusted HEPs.
The Monte Carlo sampling process was selected for the parametric analysis, with 20,000 samples and a /C value of 37,000 cutsets for SCDF and 29,195 for SLERF. 100%
Binary Decision Diagram (BDD) was not possible due to insufficient computer memory; therefore, there is some percent of over counting in the uncertainty calculations and that is reasonable in the context of parametric uncertainty analysis since the primary purpose of understanding the spread of the distribution in the final point estimate result. The results are provided in Table 5.6-1, and Figures 5.6-1 through 5.6-2, each of which shows the curves of cumulative probability and probability density function.
Table 5.6-1 Parametric Uncertainty Analysis Results SCDF SLERF
/reactor-year /reactor-year 5% 50% 95% 5% 50% 95%
MC 5.81E-06 5.50E-05 5.07E-04 7.71E-07 8.74E-06 9.91E-05 Model uncertainty is introduced when assumptions are made in the SPRA model and inputs to represent plant response, when there may be alternative approaches to particular aspects of the modeling, or when there is no consensus approach for a particular issue. For the SPRA, the important model uncertainties are addressed through the sensitivity studies described in Section 5.7 to determine the potential impact on SCDF or SLERF.
In terms of completeness uncertainty, the SPRA scope and level of detail is evaluated through the SPRA Peer Review to support the technical adequacy needed for risk-informed decision making.
Page 77 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT
~
.,v 0.8
~
/
0.6
~
!I 1
/
0.4
~ 0.2
__.,,,,, /
0 1E-07 1E-06 1E-05 1E-04 1E-03 1E-02 1E-01 1E+OO
/
0.8
~
"'\.
1 I
0.6
~
~ 0.4
"!,I ~
~ 0.2 J
_/ i'-...
0 1E-07 1E-06 1E-05 1E-04 1E-03 1E-02 1E-01 1E+OO Figure 5.6 RNP SPRA SCDF Parametric Uncertainty (Monte Carlo, 20K Samples)
/ ~
/
0.8 1
/'
0.6
!I 1
/
0.4
~ 0.2 0
/
1E-08 1E-07 1E*06 1E-05 1E*04 1E-03 1E*02 1E-01 1E+OO J
0.8 1 ~
/ \
0.6
~
/
~ 0.4
"!,I
~
~
0.2
_J -
0 1E-08 1E-07 1E*06 1E-05 1E*04 1E-03 1E*02 1E-01 1E+OO Figure 5.6 RNP SPRA LERF Parametric Uncertainty (Monte Carlo, 20K Samples)
Page 78 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT 5.7 SPRA Quantification Sensitivity Analysis Various sources of model uncertainties were reviewed and examined to identify sources that may have a significant impact on the SCDF and SLERF. A detailed description of each sensitivity is provided in the station calculation documenting the sensitivity notebook
[70]. Table 5.7-1 shows a summary of the SPRA sensitivity analysis.
Page 79 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table 5.7-1: Summary of RNP SPRA Sensitivity Cases Sensitivity Delta % Delta Delta % Delta Description CDF LERF Case # CDF CDF LERF LERF Base Case Base Case 9.31E-052 -- -- 2.02E-05 -- --
Use the Upper Bound percentile hazard IE-1a 3.58E-04 2.65E-04 285% 7.67E-05 5.65E-05 280%
curve (95th percentile)
Use the Lower Bound percentile hazard IE-1b 4.35E-06 -8.88E-05 -95% 8.14E-07 -1.94E-05 -96%
curve (5th percentile)
IE-1c Use 12 hazard intervals 9.31E-05 -2.48E-08 -0.03% 2.00E-05 -1.57E-07 -0.8%
SY-1a Half median capacity of OSP 9.48E-05 1.63E-06 1.7% 2.05E-05 2.58E-07 1.3%
Screen fragility groups with HCLFP SY-1b 9.30E-05 -8.93E-08 -0.1% 2.02E-05 -1.05E-08 -0.1%
0.75g Increase Demand of Liquefaction-Induced SY-1c 1.03E-04 9.91E-06 11% 2.19E-05 1.67E-06 8.3%
Settlement Failures Increase Capacity of DFOST Piping to SY-1d Resist Liquefaction-Induced Settlement 8.98E-05 -3.32E-06 -3.6% 1.98E-05 -4.34E-07 -2.1%
Failure Decrease Capacity of Deepwell Pumps SY-1e A/B/C to Resist Liquefaction-Induced 9.44E-05 1.29E-06 1.4% 2.03E-05 1.34E-07 0.7%
Settlement Failure Increase Capacity of Deepwell Pump D to SY-1f Resist Liquefaction-Induced Settlement 9.31E-05 -2.23E-08 -0.02% 2.01E-05 -7.55E-08 -0.4%
Failure Increase Capacity of SDAFW Pump to SY-1g Resist Liquefaction-Induced Settlement 8.97E-05 -3.39E-06 -3.6% 1.98E-05 -4.31E-07 -2.1%
Failure Decrease Capacity of SW Piping to SY-1h Resist Liquefaction-Induced Settlement 9.37E-05 5.86E-07 0.6% 2.03E-05 1.24E-07 0.6%
Failure SY-1i Increase Capacity of TBCrane 8.86E-05 -4.54E-06 -4.9% 1.94E-05 -7.57E-07 -3.7%
SY-1j Decrease Capacity of RCB 1.34E-04 4.12E-05 44% 6.14E-05 4.12E-05 204%
Page 80 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table 5.7-1: Summary of RNP SPRA Sensitivity Cases Sensitivity Delta % Delta Delta % Delta Description CDF LERF Case # CDF CDF LERF LERF SY-2a Decrease TB3-TB1 interaction probability 8.62E-05 -6.89E-06 -7.4% 1.93E-05 -9.09E-07 -4.5%
SY-2b Increase TB3-TB1 interaction probability 1.00E-04 6.88E-06 7.4% 2.11E-05 9.35E-07 4.6%
Decrease TBCrane-CST interaction SY-2c 8.95E-05 -3.67E-06 -3.9% 1.97E-05 -5.16E-07 -2.6%
probability Set TBCrane-CST interaction probability SY-2d 8.20E-05 -1.11E-05 -12% 1.86E-05 -1.55E-06 -7.7%
to zero SY-2e Assume TB3 fails SDAFW 1.06E-04 1.31E-05 14% 2.21E-05 1.88E-06 9.3%
Assume Lateral spreading to DC2 does SY-2f N/A1 N/A N/A 1.92E-05 -1.01E-06 -5.0%
not lead directly to LERF SY-3a Assume no relay chatter scenarios 8.84E-05 -4.76E-06 -5.1% 1.96E-05 -5.78E-07 -2.9%
SY-5a Uncorrelate seismic fragility groups 9.25E-05 -6.68E-07 -0.7% 1.66E-05 -3.56E-06 -18%
Correlate liquefaction-induced settlement SY-5b 1.20E-04 2.71E-05 29% 2.42E-05 3.96E-06 20%
failures HR-2a Credit Portable FLEX SG Makeup 9.16E-05 -1.51E-06 -1.6% 1.98E-05 -3.53E-07 -1.7%
All HEPs and JHEPS set to 95%
HR-2b 9.50E-05 1.92E-06 2.1% 2.03E-05 9.97E-08 0.5%
percentile HR-2c All HEPs and JHEPS set to 5% percentile 9.19E-05 -1.25E-06 -1.3% 2.01E-05 -1.03E-07 -0.5%
Note 1: Seismic CDF is not quantified as this only affects the LERF model.
Note 2: This value was based on the combined cutsets while 9.27E-05 was per the summation of the hazard bin results.
Page 81 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT 5.8 SPRA Logic Model and Quantification Technical Adequacy The Robinson Nuclear Power Plant SPRA risk quantification and results interpretation methodology were subjected to an independent peer review against the pertinent requirements in the ASME/ANS PRA Standard [37].
The peer review assessment, and subsequent disposition of peer review findings, is described in Appendix A, and establishes that the SPRA seismic plant response analysis is suitable for this SPRA application.
Page 82 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT 6.0 Conclusions A seismic PRA has been performed for Robinson Nuclear Power Plant in accordance with the guidance in the PRA Standard [5 and 37] and the SPID [2]. The Robinson SPRA shows that the point estimate SCDF is 9.27x10-5/reactor-year and the SLERF is 2.02x10-5/reactor-year. The SPRA as described in this submittal reflects the as-built/as-operated Robinson Nuclear Power Plant as of the freeze date - June 2015 [77]. An assessment is included in Appendix A of the impact of the results of plant changes not included in the model since the model freeze date.
The insights from this study reveal the SCDF and SLERF are dominated by the seismic failure of the Turbine Building Class 3 caused by building pounding between the RAB and the mezzanine floor portion of the Turbine Building Class 3. To mitigate this unique seismic vulnerability, Robinson will implement a plant modification to provide additional protection from seismic hazards. The modification involves changes to the existing FLEX strategy to provide AFW flow to the SGs (See Table 6-1). The results of the corresponding sensitivity analysis show the SCDF and SLERF reductions to be approximately 40 percent and 30 percent, respectively. The following action(s) will be performed as a result of the SPRA.
Table 6-1 Planned Actions Action System Description Action Description Completion Date Auxiliary Feedwater This modified FLEX strategy By 12/31/2022 1 provided by a modified involves SG makeup using FLEX strategy intermediate pressure AFW pumps with available water sources at the site (e.g.,
Lake Robinson or alternate water sources, such as existing or new tanks).
Page 83 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT 7.0 References
- 1) NRC (E Leeds and M Johnson) Letter to All Power Reactor Licensees et al.,
Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(f) Regarding Recommendations 2.1, 2.3 and 9.3 of the Near-Term Task Force Review of Insights from the Fukushima Dai-Ichi Accident, March 12, 2012.
- 2) EPRI 1025287, Seismic Evaluation Guidance: Screening, Prioritization and Implementation Details (SPID) for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1: Seismic. Electric Power Research Institute, Palo Alto, CA: February 2013.
- 3) Duke Energy Letter, Seismic Hazard Evaluation, Response to NRC 10 CFR 50.54(f) Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(f) Regarding Recommendations 2.1, 2.3 and 9.3 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident, dated March 31, 2014 (ML14099A204)
- 4) Duke Energy Letter, Submittal of Revision to Seismic Hazard Evaluation to Include New Ground Motion Response Spectra (GMRS) Using New Geotechnical Data and Shear-Wave Testing for H. B. Robinson Steam Electric Plant, Unit No. 2, dated July 17, 2015 (ML15201A006)
- 5) ASME/ANS RA-S-2008, Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, including Addenda [specify Addenda A, 2009 or Addenda B, 2013 as appropriate],
American Society of Mechanical Engineers, New York, [February 2, 2009] or
[September 30, 2013]
- 6) NEI-12-13, External Hazards PRA Peer Review Process Guidelines, Revision 0, Nuclear Energy Institute, Washington, DC, August 2012
- 7) Peer Review of the Robinson Unit 2 Seismic Probabilistic Risk Assessment, PWROG-18063-P Rev 0, March 2019
- 8) EPRI NP 6041-SL, A Methodology for Assessment of Nuclear Power Plant Seismic Margin, Rev. 1. Electric Power Research Institute, Palo Alto, CA, August 1991.
- 9) Duke Energy Letter, Compliance Letter and Final Integrated Plan in Response to the March 12, 2012 Commission Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events (Order Number EA-12-049) For H. B. Robinson Steam Electric Plant, Unit No. 2, dated August 19, 2015 (ML152321A007)
- 10) Duke Energy Letter, H. B. Robinson Steam Electric Plant, Unit No. 2 Expedited Seismic Evaluation Process Report (CEUS Sites), Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident, dated December 17, 2014 (ML143651A105)
- 11) EPRI 3002000709, Seismic PRA Implementation Guide, Electric Power Research Institute, Palo Alto, CA, December 2013
- 12) Regulatory Guide 1.200, Revision 2, An Approach For Determining The Technical Adequacy Of Probabilistic Risk Assessment Results For Risk-Informed Activities, U.S. Nuclear Regulatory Commission, March 2009 Page 84 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT
- 13) Station calculation for the Robinson Seismic Probabilistic Risk Assessment Quantification Notebook
- 14) NUREG-1855, Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decision Making, Rev. 0, March 2009
- 15) EPRI 1016737, Treatment of Parameter and Modeling Uncertainty for Probabilistic Risk Assessments, Electric Power Research Institute, Palo Alto, CA, December 2008
- 16) Station calculation for the Final Seismic Analysis
- 17) Sohl, Norman F. and Owens, James P., 1991. Cretaceous Stratigraphy of the Carolina Coastal Plain, in: The Geology of the Carolinas, Carolina Geological Society Fiftieth Anniversary Volume, Chapter 11, University of Tennessee Press, pp 191-199.
- 18) Electric Power Research Institute (EPRI), U.S. Department of Energy, and U.S.
Nuclear Regulatory Commission, 2012. Central and Eastern United States (CEUS) Seismic Source Characterization for Nuclear Facilities, EPRI Report 1021097/ DOE/NE-0140/ NUREG-2115
- 19) Electric Power Research Institute (EPRI), 2015a. Central and Eastern United States Seismic Source Characterization for Nuclear Facilities: Maximum Magnitude Distribution Evaluation,Technical Report 3002005684, June.
- 20) U. S. Nuclear Regulatory Commission (US NRC), 2015. Letter from Frankie G.
Vega, Project Manager, Hazards Management Branch, Japan Lessons Learned Division, Office of Nuclear Reactor Regulation, H. B. Robinson Steam Electric Plant, Unit No. 2 - Staff Assessment of Information Provided Pursuant to Title 10 of the Code of Federal Regulations for Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-Ichi Accident (TAC No.
MF3724, October 10, ADAMS Accession No, ML15280A199.
- 21) Youngs, R, 2014. Review of EPRI 1021097 Earthquake Catalog for RIS Earthquakes in the Southeastern US and Earthquakes in South Carolina Near the Time of the 1886 Charleston Earthquake Sequence. Report transmitted by EPRI letter on March 5, 2014 by J. Richards.
- 22) Electric Power Research Institute (EPRI), 2015b. Central and Eastern United States Seismic Source Characterization for Nuclear Facilities: Review for Reservoir-Induced Seismicity (RIS) in the Southeast and Earthquakes in South Carolina Near the 1886 Charleston Earthquake, Technical Report 3002005288, September.
- 23) Electric Power Research Institute (EPRI), 2013. EPRI (2004, 2006) Ground-Motion Model (GMM) Review Project, Report 3002000717.
- 24) U. S. Nuclear Regulatory Commission (US NRC), 2013. Letter from David L.
Skeen, Director, Japan Lessons-Learned Project Directorate, Office of Nuclear Reactor Regulation, to Ms. Kimberly A. Keithline, Nuclear Energy Institute,
Subject:
Approval of Electric Power Research Institute Ground Motion Model Review Project Final Report for Use by Central and Eastern United States Nuclear Power Plants, August 28 Page 85 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT
- 25) U. S. Nuclear Regulatory Commission (NRC), 2007. Regulatory Guide 1.208: A Performance-Based Approach to Define the Site-Specific Earthquake Ground Motion, Office of Nuclear Regulatory Research, March.
- 26) U. S. Nuclear Regulatory Commission (US NRC), 2010. Final DC/COL-ISG-017, Ensuring Hazard-Consistent Seismic Input for Site Response and Soil Structure Interaction Analyses, ADAMS Accession No. ML100570203.
- 27) Nuclear Energy Institute (NEI), 2009. Consistent site response/soil-structure interaction analysis and evaluation [White Paper]: NEI, June 12
- 28) Station calculation for the Plant Area Material Properties.
- 29) Electric Power Research Institute (EPRI), 1993. Guidelines for Determining Design Basis Ground Motions, EPRI Report TR-102293, Research Project 3302, Final Report, November, Palo Alto, CA.
- 30) Silva, W., and R.B. Darragh, 1995. Engineering Characterization of Strong Ground Motion Records from Rock Sites, Electric Power Research Institute (EPRI) Report TR-102262, June.
- 31) Campbell, K.W., 2009. Estimates of shear-wave Q and 0 for unconsolidated and semiconsolidated sediments in eastern North America, Bulletin of the Seismological Society of America, v. 99, n. 4, p. 2365-2392.
- 32) McGuire, R.K., W.J. Silva, and C.J. Costantino. 2001. Technical Basis for Revision of Regulatory Guidance on Design Ground Motions: Hazard- and Risk-Consistent Ground Motion Spectra Guidelines, NUREG/CR-6728, U.S. Nuclear Regulatory Commission, Washington, DC.
- 33) American Society of Civil Engineers/Structural Engineering Institute (ASCE/SEI),
2005. Standard 43-05, Seismic Design Criteria for Structures, Systems and Components in Nuclear Facilities.
- 34) Gülerce, Z., and N.A. Abrahamson. 2011. Site-specific design spectra for vertical ground motions. Earthquake Spectra, v. 27, no. 4, pp. 1023-1047.
- 35) Detroit Edison (DTE), 2014a. Fermi 3 COLA Final Safety Analysis Report (Section 2.5.2), ADAMS ML14309A443.
- 36) Exelon Generation Company, 2006. Appendix B to Site Safety Analysis Report, Early Site Permit, Rev 4, ADAMS ML061100308, ML061100309, ML061100310, and ML061100311.
- 37) U. S. Nuclear Regulatory Commission, Letter dated March 12, 2018, Acceptance of ASME/ANS RA-S Case 1, Adams Accession Number ML18017A964.
- 38) U.S. Nuclear Regulatory Commission (USNRC), 2003. Regulatory Guide 1.198:
Procedures and Criteria for Assessing Seismic Soil Liquefaction at Nuclear Power Plant Sites.
- 39) Boulanger, R.W. and Idriss, I.M., 2012, Probabilistic Standard Penetration Test--
Based LiquefactionTriggering Procedure. Journal of Geotechnical and Geoenvironmental Engineering, ASCE, Vol. 138, No. 10, pp. 1185-1195.
- 40) Station calculation for the Liquefaction Analysis for Robinson Plant
- 41) Station calculation for the Liquefaction Sensitivity Analysis for Robinson Plant.
- 42) Station calculation for the Site-Wide Liquefaction Evaluation.
Page 86 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT
- 43) Station calculation for the Liquefaction Settlement Calculation.
- 44) Idriss, I.M., and Boulanger, R.W., 2008, Soil Liquefaction during earthquake.
Monograph MNO-12, Earthquake Engineering Research Institute, Oakland CA.
- 45) Zhang, G., Robertson, P.K., Brachman, R.W.I., 2004. Estimating Liquefaction-Induced Lateral Displacements Using the Standard Penetration Test or Cone Penetration Test, Journal of Geotechnical and Geoenvironmental Engineering, Volume 130, Number 8, August.
- 46) Station calculation for the Probability of Lateral Displacement of Continuous Liquefaction Layer.
- 47) Station calculation for the Geotechnical Analysis Report.
- 48) U.S. Nuclear Regulatory Commission, Verification and Seismic Adequacy of Mechanical and Electrical Equipment in Operating Reactors, Unresolved Safety Issue (USI) A-46 (Generic Letter 87-03), 27 Feb. 1987.
- 49) U.S. Nuclear Regulatory Commission, Procedural and Submittal Guidance for the Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities, NUREG-1407, Washington, DC, June 1991.
- 50) Seismic Qualification Utility Group, Generic Implementation Procedure (GIP) for Seismic Verification of Nuclear Plant Equipment, Revision 2, Office of Standards Development, Washington, DC, 1992.
- 51) EQE International, Seismic IPEEE H. B. Robinson Steam Electric Plant Unit No.
2, Report No. 52212-R-001, Revision 0, 23 June, 1995.
- 52) U.S. Nuclear Regulatory Commission, Masonry Wall Design, IE Bulletin No. 80-11, Washington, D.C., 8 May 1980.
- 53) Station calculation for the Seismic Capacity Walkdown Report.
- 54) Station calculation for the Structural Modeling Notebook.
- 55) Station calculation for the Response Analysis Notebook.
- 56) Station calculation for the RAB Incoherence Sensitivity Study.
- 57) Duke Energy Letter to the NRC, Spent Fuel Pool Evaluation Supplemental Report, Response to NRC Request for Information Pursuant to 10 CFR 50.54(f)
Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident August 2016 ML16215A376
- 58) American Society of Civil Engineers, Seismic Design Criteria for Structures, Systems, and Components in Nuclear Facilities, ASCE/SEI 43-05, American Society of Civil Engineers, Reston, VA, 2005.
- 59) American Society of Civil Engineers, Seismic Analysis of Safety-Related Nuclear Structures and Commentary, Draft ASCE/SEI 4-16, American Society of Civil Engineers, Reston, VA, 2017.
- 60) NRC Letter to Duke Energy, H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 - RESPONSE TO REQUEST FOR EXTENSION OF SEISMIC PROBABILISTIC RISK ASSESSMENT SUBMITIAL (EPID NO. L-2018-JLD-0017), Jan 2019, ML19004A356
- 61) Station calculation for the Robinson SPRA Representative Fragilities.
Page 87 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT
- 62) Electric Power Research Institute, Methodology for Developing Seismic Fragilities, EPRI TR-103959, Palo Alto, CA, 1994.
- 63) Electric Power Research Institute, Seismic Fragility Applications Guide Update, EPRI 1019200, Palo Alto, CA, Dec. 2009.
- 64) Station calculation for the Seismic Equipment List
- 65) Station calculation for the Robinson Seismic Probabilistic Risk Assessment Human Reliability Analysis Notebook
- 66) Station Calculation for the Robinson PRA System Notebook
- 67) Station calculation for the Seismically Induced Liquefaction Fragility Evaluation of Robinson Dam
- 68) Station calculation for the Seismic Fragility of Pounding between the Class III Turbine Bldg. and the Reactor Aux. Bldg. by Nonlinear Analysis
- 69) Station calculation for the Robinson Nuclear Power Plant Seismic-Induced Flood and Fire Assessment
- 70) Station calculation for the Robinson Seismic Probabilistic Risk Assessment Uncertainty and Sensitivity
- 72) Station calculation for the Robinson Nuclear Power Plant Relay Contact Chatter Analysis
- 73) Station Calculation for the VSLOCA Success Criteria Development
- 74) Not Used
- 75) Not Used
- 76) Station calculation for the Containment Safeguards and Plant Damage States
- 77) Station calculation for H.B. Robinson Seismic Probabilistic Risk Assessment Model Notebook
- 78) Not Used
- 79) Not Used
- 80) Station calculation for Assessment of Internally Initiated Flood Events
- 81) Station calculation for Fire PRA Quantification
- 82) Station calculation for Seismic Fragility Evaluation Notebook
- 83) Station calculation for the Geotechnical Analysis Report
- 84) Not Used
- 85) Station calculation for the Seismic Fragility Evaluation of the Condensate Storage Tank (CST) by the Separation of Variables Method
- 86) LTR-RAM-19-106 rev 0, Robinson Focused Scope Peer Review of LERF and Dam Fragility, November 6, 2019
- 87) Station calculation for the Response Analysis of the Robinson Unit 2 Intake Structure: Phase 2 SPRA Project Page 88 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT
- 88) Station calculation for Internal Events PRA Model
- 89) EPRI Technical Report 1020756, Surry Seismic Probabilistic Risk Assessment Pilot Plant Review, July 2010
- 90) Station calculation for the Oconee SPRA Model Notebook
- 91) NRC, NUREG/CR-6595 Rev 1, An Approach for Estimating the Frequencies of Various Containment Failure Modes and Bypass Events, Washington, D.C. 2004
- 92) NRC, Regulatory Guide 1.174, Rev3, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant Specific Changes to the Licensing Basis, January 2018
- 93) Electric Power Research Institute, Methodology for Seismically Induced Internal Fire and Flood Probabilistic Risk Assessment, 3002012980, June 27, 2018
- 94) Station calculation for the Seismic Fragility Evaluation of the RCB
- 95) Station calculation for the Reactor Auxiliary Building Soil Structure Interaction Analysis: SPRA Project
- 96) Station calculation for the Reactor Building Soil Structure Interaction Analysis
- 97) Choudhury, K.H., 2019, Evaluation of the State of Practice Regarding Nonlinear Seismic Deformation Analyses of Embankment Dams Subject to Soil Liquefaction Based on Case Histories,: PhD dissertation, University of California at Berkley, Spring
- 98) R. Fell, 2002, Damage and Cracking of Embankment Dams by Earthquake for Internal Erosion and Piping, proceedings of the 20th Congress on Large Dams, International Commission on Large Dams.
Page 89 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT 8.0 Acronyms AFE Annual Frequency of Exceedance AFW Auxiliary Feedwater ANS American Nuclear Society ASCE American Society of Civil Engineers ASME American Society of Mechanical Engineers ATWS Anticipated Transient Without Scram (also ATWT, Anticipated Transient Without Trip)
BDD Binary Decision Diagram BOP Balance of Plant CDFM Conservative Deterministic Failure Model CEUS Central and Eastern United States CMS Conditional Mean Spectra CSR Cyclic Stress Ratio CRR Cyclic Resistance Ratio DRS Design Response Spectrum ECCS Emergency Core Cooling System EPRI Electric Power Research Institute ESEP Expedited Seismic Evaluation Program FEM Finite Element Model FIRS Foundation Input Response Spectra FLEX Diverse and Flexible Coping Strategies FV Fussel-Vesely GIP Generic Implementation Procedure GMC Ground Motion Characterization GMI Ground Motion Incoherence GMRS Ground Motion Response Spectra HBRSEP H.B. Robinson Steam Electric Plant, Unit No. 2 HEP Human Error Probability HFE Human Failure Event IPEEE Individual Plant Examination for External Events ISRS In-Structure Response Spectra HF High Frequency LF Low Frequency LMSM Lumped Mass Stick Model LOCA Loss of Coolant Accident MAFE Mean Annual Frequency of Exceedance MDAFW Motor Driven Auxiliary Feedwater NEI Nuclear Energy Institute NGVD National Geodetic Vertical Datum NRC Nuclear Regulatory Commission NSSS Nuclear Steam Supply System NTTF Near Term Task Force PGA Peak Ground Acceleration PRT Peer Review Team PSHA Probabilistic Seismic Hazard Analysis Page 90 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT RLME Repeated Large Magnitude Earthquake RNP Robinson Nuclear Power Plant RPS Reactor Protection System SBO Station Blackout SCDF Seismic Core Damage Frequency SCOR Soil Column Outcrop Response SDAFW Steam Driven Auxiliary Feedwater SEL Seismic Equipment List SEWS Screening Evaluation Worksheets SFP Spent Fuel Pool SFR Seismic Fragility Element Within ASME/ANS PRA Standard SG Steam Generator SHA Seismic Hazard Analysis Element Within ASME/ANS PRA Standard SHS Seismic Hazard Submittal SLERF Seismic Large Early Release Frequency SMA Seismic Margin Assessment SOV Separation of Variables SPID Screening, Prioritization and Implementation Details SPR Seismic PRA Modeling Element Within ASME/ANS PRA Standard SPRA Seismic Probabilistic Risk Assessment SQUG Seismic Qualification Utility Group SRSS Square Root of the Sum of the Squares SRT Seismic Review Team SSC Structure, System or Component SSEL Safe Shutdown Equipment List SSHAC Senior Seismic Hazard Analysis Committee SSI Soil Structure Interaction TSCR Truncated Soil Column Response UHRS Uniform Hazard Response Spectra UHS Ultimate Heat Sink USI Unresolved Safety Issue Page 91 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Appendix A Summary of SPRA Peer Review and Assessment of PRA Technical Adequacy for Response to NTTF 2.1 Seismic 50.54(f)
Letter This Appendix has two purposes:
- 1. Provide a summary of the SPRA peer review
- 2. Provide the bases for why the SPRA is technically adequate for the 50.54(f) response.
The Robinson PRA was subjected to an independent peer review against the pertinent requirements of ASME / ANS RA-S Case 1 [37] (Code Case), which is an accepted alternate approach to Part 5 (Seismic) of Addenda B of the PRA Standard [5].
The information presented here establishes that the Robinson SPRA has been peer reviewed by a team with adequate credentials to perform the assessment, establishes that the peer review process meets the intent of the peer review characteristics and attributes in Table 16 of Regulatory Guide 1.200 R2 [12] and the requirements in Section 1-6 of the ASME/ANS PRA Standard [5 and 37], and presents the significant results of the peer review.
A.1. Overview of Peer Review The peer review assessment [7], and subsequent disposition of peer review findings, is summarized here. The scope of the review encompassed the set of technical elements and supporting requirements (SR) for the SHA (seismic hazard), SFR (seismic fragilities), and SPR (seismic PRA modeling) elements for seismic CDF and LERF. The peer review therefore addressed the set of SRs identified in Tables 6-4 through 6-6 of the SPID [2].
The Robinson SPRA peer review was conducted during the week of November 11, 2018 at the Duke offices in Charlotte North Carolina. As part of the peer review, a walk-down of portions of Robinson Nuclear Plant was performed on November 12, 2018 by selected members of the peer review team.
A Focused Scope Peer Review, FSPR, of LERF and Dam Failure Fragility was also performed remotely September 25-29, 2019. The scope of the review was the SRs related to the revised LERF hydrogen analysis as well as the refined Robinson Dam fragility.
A.2. Summary of the Peer Review Process The peer review was performed against the requirements of ASME / ANS RA-S Case 1 [37],
which is an accepted alternate approach to Part 5 (Seismic) of Addenda B of the PRA Standard
[5]. The team utilized the peer review process defined in NEI 12-13 [6]. The review was conducted over a 30-day period, including 3 weeks of offsite review prior to a five day on-site portion of the review.
The SPRA peer review process defined in [6] involves an examination by each reviewer of their assigned PRA technical elements against the requirements in the Code Case to ensure the robustness of the model relative to all of the requirements.
Implementing the review involves a combination of a broad scope examination of the PRA elements within the scope of the review and a deeper examination of portions of the PRA elements based on what is found during the initial review. The supporting requirements (SRs) provide a structure which, in combination with the peer reviewers PRA experience, provides the basis for examining the various PRA technical elements. If a reviewer identifies a question or Page 92 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT discrepancy, it leads to additional investigation until the issue is resolved or a Fact and Observation (F&O) is written describing the issue and its potential impacts and suggesting possible resolution.
For each technical element (i.e., SHA, SFR, SPR), a team of two peer reviewers were assigned, one having lead responsibility for that area. For each SR reviewed, the responsible reviewers reached consensus regarding which of the Capability Categories defined in the Code Case that the PRA meets for that SR and the assignment of the Capability Category for each SR was ultimately based on the consensus of the full review team. The Code Case also specifies high level requirements (HLR). Consistent with the guidance in the Code Case and the Standard, capability Categories were not assigned to the HLRs, but a qualitative assessment of the applicable HLRs in the context of the PRA technical element summary was made based on the associated SR Capability Categories.
As part of the review teams assessment of capability categories, F&Os are prepared. There are three types of F&Os defined in [6]: Findings, which identify issues that must be addressed in order for an SR (or multiple SRs) to meet Capability Category II; Suggestions, which identify issues that the reviewers have noted as potentially important but not requiring resolution to meet the SRs; and Best Practices, which reflect the reviewers opinion that a particular aspect of the review exceeds normal industry practice. The focus in this Appendix is on Findings and their disposition relative to this submittal as well as Findings from the Focused Peer Review conducted in September 2019.
A.3. Peer Review Team Qualifications The review was conducted by: Kenneth Kiper of Westinghouse, Mr. Jeffrey Kimball of Rizzo International, Inc.; Dr. Arash Zandieh of Lettis Consultants International, Inc.; Dr. Ram Srinivasan, independent consultant; Joe Vasquez of Dominion Energy Company; Benny Ratnagaran of Southern Nuclear Company; Dr. Andrea Maioli of Westinghouse; and Nathan Barber of Pacific Gas & Electric Company and Chris Peckat of American Electric Power participated as a working observer.
The team was assembled by the peer review team lead. The lead and reviewer qualifications are summarized here below and have been reviewed by Duke Energy and have been confirmed to be consistent with requirements in the ASME/ANS PRA Standard and the guidelines of NEI 13.
Mr. Kenneth Kiper, the team lead, has over 35 years of experience at Westinghouse and, previously at Seabrook Station, in the nuclear safety area generally and PRA specifically for both existing and new nuclear power plants. He has lead a number of peer reviews, including reviews of internal events PRAs, internal flood PRAs, fire PRAs, high wind PRAs, and several seismic PRAs.
Mr. Jeff Kimball was the lead for the review of the Seismic Hazard Analysis (SHA) technical element. Mr. Kimball has over 38 years of experience in site characterization; ground motion modeling including site response and probabilistic seismic hazard analysis (PSHA). Mr. Kimball has served as SHA reviewer for a number of recent SPRAs and serves on the Participatory Peer Review Panel for the NGA-East Project. Mr. Kimball was assisted by Dr. Arash Zandieh. Dr.
Zandieh has 8 years of experience in seismic hazard analysis, earthquake engineering, engineering seismology, geotechnical and structural engineering, and statistical analysis. Dr.
Zandieh has participated in a number of SPRA peer reviews.
Dr. Ram Srinivasan led the Seismic Fragility Analysis (SFR) review. Dr. Srinivasan has over 45 years of experience in the nuclear industry, principally in the design, analysis (static and dynamic, including seismic), and construction of nuclear power plant structures. He is actively involved in Page 93 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT the Post-Fukushima Seismic Assessments (NRC NTTF 2.1 and 2.3) and is a member of the NEI Seismic Task Force and the ASME/ANS JCNRM, Part 5 Working Group (Seismic and other External Hazards PRA). He has participated on several previous SPRA peer reviews, either as reviewer or utility consultant. Dr. Srinivasan was assisted by Mr. Joe Vasquez and Mr. Benny Ratnagaran. Mr. Vasquez has 17 years of nuclear engineering experience covering all areas within the Engineering Mechanics including pipe stress analysis, pipe and equipment support analysis, pressure vessel design and analysis, seismic qualification of mechanical and electrical equipment, seismic margins assessment and fragility analyses, and fracture mechanics. He has participated on several previous SPRA peer reviews, either as reviewer or utility defender. Mr.
Ratnagaran has 7 years of experience in developing seismic PRAs, seismic response analysis, and structural fragility analyses. He has also participated on several previous SPRA peer reviews, either as reviewer or utility defender.
Dr. Andrea Maioli was the lead for the review of the Seismic System Response Analysis (SPR) technical element. Dr. Maioli has over 15 years of experience in the nuclear safety area generally and seismic PRA specifically. He has served as lead engineer for a number of seismic PRA and seismic margin studies for existing and new nuclear power plants. He has participated in and led a number of SPRA peer reviews. Dr. Maioli was assisted by Mr. Nathan Barber. Mr.
Barber is a nuclear engineer and mechanical engineer with over 16 years experience working in the nuclear power industry. He is the technical lead for the Seismic PRA update at Diablo Canyon. He has participated in a number of peer reviews, including internal events PRAs and SPRAs. Mr. Chris Peckat from American Electric Power served as working observer for the SPR technical element. Any observations and findings that Mr. Peckat generated were given to the peer review team for their review and ownership. As such Mr. Peckat assisted with the review but was not a formal member of the peer review team.
This peer review report was compiled by the peer review team lead. A draft copy of the peer review report was sent to Duke Energy as well as the other peer review members on Dec 10, 2018.
The Focused Scope Peer Reviewers qualifications are given below:
Mr. LaBarge is a Principal Engineer in the Risk Applications and Methods group Westinghouse and has approximately 14 years of experience with PRA models in the nuclear industry. Mr.
LaBarge has experience developing Level 1 and Level 2 PRA models and methods for a variety of applications. Mr. LaBarge has experience working closely with utilities in order to create PRA models that are consistent with the as-built as-operated plant. Mr. LaBarge is one of the Westinghouse experts in the area of Level 1 and Level 2 PRA model development, MAAP analysis, severe accident analysis and SAMG development. He is also currently the program committee chair for the ANS NISD vice chair of the JCNRM Level 2 PRA Standard Writing Group and a member of the JCNRM Subcommittee on Standards Development. Mr. LaBarge has experience participating in peer reviews representing both the utilities and as a peer reviewer.
Dr. Glenn Rix is a Senior Principal in Kennesaw Georgia with expertise in seismic hazard evaluation, geotechnical earthquake engineering, and performance based and risk based analysis. He had a distinguished career as a faculty member of the school of Civil and Environmental Engineering at Georgia Institute of Technology for 24 years prior to his consulting career. Dr. Rix has participated as a member of several peer reviews as well as a defender on numerous peer reviews.
Page 94 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT A.4. Summary of the Peer Review Conclusions The review teams assessment of the SPRA elements is summarized as follows. Where the review team identified issues, these are captured in peer review findings, for which the dispositions are summarized in the next section of this appendix.
2018 Full Scope Peer Review Seismic Hazard (SHA)
The seismic hazard at RNP was evaluated using a site-specific probabilistic seismic hazard analysis (PSHA). The SPRA Standard requires the inputs to the site-specific PSHA to be based on current geological, seismological, and geophysical data; local site topography; and surficial geologic and geotechnical site properties. The RNP PSHA fully met this requirement. The RNP seismic hazard analysis and PSHA used the seismic source characterization (SSC) and ground motion model (GMM) based on SSHAC studies that have compiled comprehensive earth science datasets. The SSC model was developed as a SSHAC Study Level 3 (CEUS-SSC model; EPRI, 2012) and the GMM was developed as a SSHAC Study Level 2 update of a previous SSHAC Study Level 3 (EPRI, 2013). Both studies involved teams of experts and participatory peer review panels. The Technical Integration teams who completed these models considered the full range of earthquake data (geological, seismological, and geophysical) to develop the SSC model and GMM. The RNP PSHA included an update to the CEUS-SSC earthquake catalog and an assessment of recent seismicity and literature to determine in any changes or updates to the SSC model, seismicity rates, and GMM were needed. It was concluded that no updates to the SSC and GMM were needed.
The SPRA Standard requires the effects of local site response along with the uncertainties in characterizing the local site response analysis to be identified and included. The PSHA includes site characterization efforts to gather new geologic and geotechnical data to aid in the assessment of site response and liquefaction. Site-specific site response analyses were performed to include the effects of local site response. Uncertainties in site response inputs were included in the analyses. Therefore, RNP PSHA met the requirement to include effects of local site response.
However, insufficient basis for establishing the base case site profiles for the Plant Area, representing the epistemic uncertainty, was provided. Therefore, there was a finding that should be addressed to fully satisfy the intent of the Standard.
The Standard requires that a screening analysis be performed to assess whether in addition to the vibratory ground motion other seismic hazards need to be included in the SPRA. For RNP, screening analysis for secondary seismic hazards were performed and documented, including:
fault displacement, landslide, soil liquefaction, liquefaction-induced settlement, soil settlement related to non-liquefying seismic events, potential for earthquake-induced flooding, fracking induced earthquakes, seismic seiches and Tsunami. Two secondary seismic hazards were screened in: soil liquefaction and liquefaction-induced settlement. For those hazards, probabilistic distributions for liquefaction-induced settlement and lateral spreading displacements at locations of important structures, systems and components for different hazard levels were evaluated.
The SPRA Standard requires documentation of the hazard evaluation to be consistent with the applicable supporting requirements. The documentation of the RNP PSHA is a collection of reports that describes the geotechnical studies, site profile development, PSHA methodology, rock hazard results, site response analyses and results, soil hazard results, and the assessment of secondary seismic hazards including liquefaction. The SPRA standard requires the hazard evaluation to be documented in a manner that facilitates PRA applications, upgrades, and peer review. Moreover, the process used to perform the hazard evaluation and the evaluation results is also required to be documented. Overall, the documentation for the PSHA is complete and Page 95 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT meets the intent of the Standard. However, there are a few elements that require additional discussions and justifications. Therefore, there were several findings in the documentation that should be addressed to fully satisfy the intent of the Standard.
Seismic Fragility (SFR)
The SFR assessment of RNP SPRA covered three principal elements of the fragility analysis:
site-specific seismic response analysis, plant walkdown, and fragility analysis calculations. A summary of the three elements are briefly summarized below.
The seismic response analyses of the RNP Structures that feed into the fragility evaluations are based on input response spectra corresponding to the GMRS generated and reported in SHA documents. It is seen that over 50% plant seismic CDF risk contribution would occur at earthquake input levels at or below the GMRS. Thus, the selection of the GMRS as the Reference Earthquake for the RNP SPRA is appropriate. Five sets of time histories of ground motion corresponding to the GMRS were generated. Each set included two horizontal components and one vertical component.
RNP developed new finite element models (FEM) for seismic response analysis. In a few cases, existing lumped mass spring models (LMSM) were enhanced to conform to the current practice.
The peer review team concurs that the structural models are generally realistic.
RNP performed median-centered response analysis for various structures and considered the appropriate variabilities. Soil-structure interaction (SSI) was considered for the Reactor Containment Building (RCB), and Reactor Auxiliary Building (RAB). SSI effects were deemed to be not significant for the lighter Turbine Building (TB). The SSI analysis included the pile-soil spring elements.
RNP performed probabilistic response analysis in developing the fragility of the Turbine Building (Class III). Thirty simulations were used following Latin Hypercube sampling to ensure stability of the analysis. The analyses were performed corresponding to three earthquake levels (GMRS, 70% GMRS, and 85% GMRS).
Seismic walkdowns performed for the Robinson SPRA were generally found to be comprehensive and complete. Walkdown documentation is voluminous and meets expectation.
While a few issues were identified based on peer review team (PRT) walkdown where seismic interactions may not have been noted on walkdown forms and/or bases for seismic review team (SRT) walkdown judgments were not clearly evident, the preponderance of evidence suggests thorough walkdowns were performed and documented by experienced personnel. Per response from RNP, the SPRA model has been reviewed for plant changes up to October 2017. Many ex-control room operator actions were characterized as having multiple pathways therefore investigations were focused only on the equipment that needs to be manipulated and the immediately adjacent areas. Seismic induced fire and flood sources were assessed and documented. With a few exceptions, walkdowns performed adequately identified credible seismic interactions and consequences of potential interactions identified were adequately addressed within fragility documentation developed.
Fragilities were calculated for all the relevant failure modes identified for SSCs (in SFR-E2) that significantly contribute to the seismic CDF or seismic LERF. In addition to the typical functional, structural, and anchorage fragility modes, soil liquefaction and building interactions were identified to be significant for RNP SSCs. As noted in the SHA review, detailed probabilistic assessment of the soil liquefaction effects was performed for affected SSCs. Detailed analysis of the TB and RAB interaction was performed though the peer review team determined to be conservative. RNP fragility evaluation notebook describes the methods used to calculate the seismic fragilities of SSCs that are in the PRA model. The Separation of Variables (SoV)
Page 96 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT method was used for most of the top risk contributors. For non-risk contributors, representative fragilities using EPRI Hybrid approach were calculated. For a few SSCs, that do not contribute significantly to the SCDF or SLERF, judgement based lower bound fragilities were used.
Considering that the fragility of the top risk contributor (TB pounding the RAB) to CDF and LERF was conservatively calculated and representative fragilities were used for a few risk significant components, the peer review team assessed the fragility of top risk contributors to be not realistic.
The peer review team was able to perform the peer review using the documentation received from the project team. Only a few minor items would require correction.
In summary, the fragility analysis generally meets the applicable requirements of the ASME/ANS RA-S Standard CODE CASE 1. However, the peer review team believes that further refinement of the fragility of the top contributor is likely to decrease the SPRA CDF and LERF estimates.
Peer Review Team Interpretation of Supporting Requirement SFR-E3 During a previous SPRA peer review, the peer review team identified an issue with the wording of SR SFR-E3 and concluded that, as published, the wording for CC-I does not match with the intent of the SR due to two typos. Following a dedicated discussion with the authors of the SFR section in the code and with the JCNRM leadership, the typos were confirmed and the team performed the review against a modified version of the SFR-E3 which, for CC-I, reads as follows:
ESTIMATE seismic fragilities for the failure modes of interest identified in SFR-E2 using plant-specific data and ENSURE that they are realistic conservative. JUSTIFY (e.g., through the calculation of seismic CDF and LERF per HLR-SPR-E) the use of generic fragility data (e.g., fragility test data, generic seismic qualification test data, and earthquake experience data) or conservative assumptions for the SSCs as being appropriate for the plant and not significant to the overall results.
This modified wording has been formally approved by the JCNRM. Note that the wording and understanding of SFR-E3 CC-II remain unchanged.
Seismic Plant Response (SPR)
The Robinson seismic plant response (SPR) model integrates the site-specific hazard, fragilities and system-analysis and accident sequence aspects. The starting point for the analysis was the existing internal events PRA model. Limited modifications were made to the underlying model, including the addition of FLEX strategies. These modifications were added to the model in a fashion consistent with the requirements in Part 2 of the Standard.
The RNP SPRA used standard EPRI tools (i.e., CAFTA, FRANX, ACUBE) to incorporate the seismic induced failures within the internal events PRA logic. These tools retained all the underlying random failures and operator actions and then were used to quantify the seismic induced CDF and LERF.
A detailed Seismic Equipment List was generated and the associated fragilities were included in the model consistently with the observations from the walkdowns and correlation considerations made in the fragility analysis. Because of the unique nature of the seismic hazard at the RNP site, the RNP SPRA includes two types of fragilities:
(a) Lognormal fragilities were used to model functional, anchorage and special interaction failures. These fragilities were managed through the FRANX code mapping to the underlying internal events logic. Full correlation was assumed for the modeling of the lognormal fragilities.
Page 97 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT (b) Non-lognormal fragilities were used to model soil failure and lateral spreading. These fragilities were manually added to the internal events logic. For these non-lognormal fragilities, capacities were assumed to be uncorrelated and the fragility groupings were based on the lateral spreading profiles.
The RNP SPRA explicitly models a reasonably complete set of seismic induced initiators, derived from the internal events model. However, the peer review team noted the absence of documentation of a systematic process to derive internal events and external events challenges to the plant and to identify similar seismic-induced versions of such challenges. Since documentation is critical in supporting and reproducing any screening process, the initial technical supporting requirements SPR-A1 and A2, which are being considered not met and related findings were written regarding the need for a documented systematic assessment. Absence of a review of operating experience at the RNP plant associated with seismic events also resulted in SPR-A3 being judged not met.
Seismic performance shaping factors were systematically considered in adapting the human reliability analysis performed for internal events to seismic-induced sequences. The RNP SPRA team went beyond the recently published EPRI method in the level of details applied to the analysis, using a more refined breakdown of each action. The review team noted that operator actions explicitly developed for the SPRA (e.g., operator action associated with FLEX) were not modeled with the same approach and should have similar considerations for different seismic hazard levels.
The RNP SPRA is quantified in a manner that allows an adequate estimation of the risk profile and identification of lead risk contributors in terms of accident sequences, individual components, fragility groups and operator actions. The quantification process is challenged by well-known and understood limitations in the tools used (e.g., truncation and stability challenges due to challenges to the rare event approximation assumption). While such challenges may be slightly overestimating the seismic risk, they are not expected to significantly change the risk insights that can be drawn by the SPRA.
Some limitations were observed in the characterization and documentation of uncertainties associated with key assumptions in the overall SPRA; associated recommendations were given on this topic and supporting requirement SPR-E7 was judged not met.
Finally, because the underlying internal events LERF model meets only capability category I, the seismic LERF model is also judged to be CC-I in SPR-E6.
Page 98 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT 2019 Focused Scope Peer Review Robinson LERF PRA Model The IE-PRA focused peer review included nine SRs from Part 2 of the PRA Standard [5] and 2 SRs from ASME/ANS RA-S CASE [37]. The peer review assessment results show that all applicable SRs were judged to be Met at CC-II or above.
A.5. Summary of the Assessment of Supporting Requirements and Findings Table A-1 presents a summary of the SRs graded as not met or not Capability Category II, and the disposition for each. Table A-2 presents summary of the Finding F&Os that have not been closed through an NRC accepted process, and the disposition for each (included at the end of this Appendix due to size).
Table A-1: Summary of SRs Graded as Not Met or Capability Category I for Supporting Requirements Covered by the Robinson Nuclear Power Plant SPRA Peer Review Assessed Associated SR Capability Finding F&Os Disposition to Achieve Met or Capability Category II Category SHA N/A N/A N/A N/A SFR Page 99 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-1: Summary of SRs Graded as Not Met or Capability Category I for Supporting Requirements Covered by the Robinson Nuclear Power Plant SPRA Peer Review Assessed Associated SR Capability Finding F&Os Disposition to Achieve Met or Capability Category II Category Findings 28-2 and 29-2 were written concerning the use of generic conservative fragilities for SLOCA and SSLOCA and the abbreviated resolution follows: Following the original walkdowns of the items on the seismic equipment list (SEL),
Duke conducted supplemental walkdowns of piping and tubing whose failure could lead to SSLOCA and SLOCA and concluded that all reviewed SSLOCA/SLOCA piping/tubing items would have High seismic capacity. Based on this, the plant specific SSLOCA and SLOCA fragilities were generated for use in the Robinson SPRA.
Finding 28-3 was written concerning a possible improvement in the Turbine Building Class 3 pounding fragility via a refined estimate of building pounding forces at the impact interface; the summarized resolution: The energy dissipation at the impact point does not offer significant protection to the 28-2 vulnerable diaphragm. While model refinement can CCI 28-3 increase the precision of the fragility, the resulting slightly 29-2 modified fragility will not change the risk conclusions.
C-SFR-E3 Not Met for 30-2 Finding 30-2 was written concerning the issue of a possible the FSPR correlation between the Turbine Building Class 3 pounding 2-1 (FSPR) and shaking fragilities; a summary of the resolution: While realistically these two failure modes should be at least partially correlated, any partial correlation curves are bounded by these two cases; and therefore, partial correlation possibilities are not an important consideration.
In summary, all the Findings identified have been appropriately resolved and meet Capability Category II of the Standard.
Page 100 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-1: Summary of SRs Graded as Not Met or Capability Category I for Supporting Requirements Covered by the Robinson Nuclear Power Plant SPRA Peer Review Assessed Associated SR Capability Finding F&Os Disposition to Achieve Met or Capability Category II Category SPR Finding 25-4 documents the lack of a systematic disposition of internal initiating events. The disposition has been C-SPR-A1 Not Met 25-4 completed and the documentation updated such that this SR is met. The resolution to this Finding meets Capability Category II of the Standard.
Finding 24-4 documents a lack of a complete identification of seismically induced consequential events. The review 24-4 has subsequently been completed, documented and no additional hazards identified.
25-2 C-SPR-A2 Not Met Finding 25-2, 25-3, and 25-10 are addressing potential 25-3 issues with seismically induced fires. All 3 potential issues 25-10 have been addressed.
The resolutions to these Findings meet Capability Category II of the Standard.
Finding 25-1 identifies the lack of documentation for any site specific events or review of industry events that could be applicable to the SPRA. A review has been documented C-SPR-A3 Not Met 25-1 and no changes were required.
The resolution to this Finding meets Capability Category II of the Standard.
Finding 24-2 was written to identify the lack of re-binning hazard intervals for optimization. Re-binning was performed and documented.
Finding 24-8 addresses uncertainty surrounding the Turbine 24-2 Building failing the SDAFW pump. Documentation of this C-SPR-E7 Not Met 24-8 uncertainty has been updated.
24-20 Finding 24-20 was written to identify the lack of a systematic review of key assumptions. This review has been complete and is documented.
The resolutions to these Findings meet Capability Category II of the Standard.
Page 101 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-1: Summary of SRs Graded as Not Met or Capability Category I for Supporting Requirements Covered by the Robinson Nuclear Power Plant SPRA Peer Review Assessed Associated SR Capability Finding F&Os Disposition to Achieve Met or Capability Category II Category Finding 24-21 relates to a lack of LERF timing change consideration. The binning of CDF to LERF sequences were re-visited and no LERF model changes were required.
Findings 24-7 and 24-22 are concerned with documentation 24-7 updates only. Based on this, the resolutions of these Findings meet Capability Category II of the Standard.
24-21 C-SPR-E6 CCI Finding 24-23 relates to the use of the underlying internal 24- 22 events LERF model Met at CC-I. Although only Met at CC-I, 24-23 the LERF methodology used by the site is an acceptable means of calculating LERF for use in risk-informed applications, including changes to its licensing basis as well as its response to NTTF 2.1 Thus, this Finding is resolved for the purposes of NTTF 2.1 seismic.
A.6. Summary of Technical Adequacy of the SPRA for the 50.54(f) Response The set of supporting requirements from the ASME/ANS PRA Standard [5] that are identified in Tables 6-4 through 6-6 of the SPID [2] define the technical attributes of a PRA model required for a SPRA used to respond to implement the 50.54(f) letter. The conclusions of the peer review discussed above and summarized in this submittal demonstrates that the RNP SPRA model meets the expectations for PRA scope and technical adequacy as presented in RG 1.200, Revision 2 [12] as clarified in the SPID [2].
The main body of this report provides a description of the SPRA methodology, including:
o Summary of the seismic hazard analysis (Section 3) o Summary of the structures and fragilities analysis (Section 4) o Summary of the seismic walkdowns performed (Section 4) o Summary of the internal events at power PRA model on which the SPRA is based, for CDF and LERF (Section 5) o Summary of adaptations made in the internal events PRA model to produce the seismic PRA model and bases for the adaptations (Section 5)
Detailed archival information for the SPRA consistent with the listing in Section 4.1 of RG 1.200 Rev. 2 is available if required to facilitate the NRC staffs review of this submittal.
The RNP SPRA reflects the as-built and as-operated plant as of the freeze date for the SPRA, June 2015. There are no permanent plant changes that have not been reflected in the SPRA model, except for those discussed further in Section A.9.
A.7. Summary of SPRA Capability Relative to SPID Tables 6-4 through 6-6 The PWR Owners Group performed peer reviews of the RNP internal events PRA and internal flooding PRA that form the basis for the SPRA to determine compliance with ANS/ASME PRA Standard RA-Sa-2009 [5] along with the NRC clarifications provided in Regulatory Guide 1.200, Revision 2 [12]. The full scope internal events peer review was performed in October 2009 and Page 102 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT the focused scope internal flooding peer review was performed in June 2015. These reviews documented findings for all supporting requirements (SRs) which failed to meet at least Capability Category II. All the internal events and internal flooding PRA peer review findings that may affect the SPRA model have been addressed.
The PWR Owners Group performed a peer review of the RNP SPRA in November 2018. The results of this peer review are discussed above, including resolution of SRs not assessed by the peer review as meeting Capability Category II, and resolution of peer review findings pertinent to this submittal. The peer review team expressed the opinion that the RNP seismic PRA model is of good quality and integrates the seismic hazard, the seismic fragilities, and the systems-analysis aspects appropriately to quantify core damage frequency and large early release frequency. The general conclusion of the peer review was that the RNP SPRA is judged to be suitable for use for risk-informed applications.
- Table A-1 provides a summary of the disposition of SRs judged by the peer review to be not met, or not meeting Capability Category II.
- Table A-2 provides a summary of the disposition of the open SPRA peer review findings (included at the end of this Appendix due to size).
- Table A-3 provides an assessment of the expected impact on the results of the RNP SPRA of those SRs and peer review Findings that have not been fully addressed.
Page 103 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-3 Summary of Impact of Not Met SRs and Open Peer Review Findings Summary of Issue Not Fully SR # F&O # Impact on SPRA Results Resolved Documentation update only and Systematic disposition of SPR-A1 25-4 no impact on the model (See internal events Table A-2 for more details).
Finding 24-4 documents a lack of a complete identification of seismically induced consequential events. The 24-4 review has subsequently been Documentation update only and SPR-A2 25-2 completed, documented and no no impact on the model (See 25-3 additional hazards identified. Table A-2 for more details).
25-10 Finding 25-2, 25-3, and 25-10 are addressing potential issues with seismically induced fires.
All 3 potential issues have been addressed this SR is now met.
Finding 25-1 identifies the lack of documentation for any site specific events or review of Documentation update only and industry events that could be SPR-A3 25-1 no impact on the model (See applicable to the SPRA. A Table A-2 for more details).
review has been documented and no changes were required; this SR is now met.
Finding 24-2 was written to identify the lack of re-binning Per FO 24-2, the model was hazard intervals for revised to place more bins at the optimization. Re-binning was lower accelerations and to 24-2 performed and documented. condense the higher SPR-E7 24-8 Finding 24-8 addresses accelerations into fewer bins.
24-20 uncertainty surrounding the Turbine Building failing the Documentation update only for SDAFW pump. Documentation FOs 24-8 and 24-20 and no of this uncertainty has been impact on the model.
updated.
Page 104 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-3 Summary of Impact of Not Met SRs and Open Peer Review Findings Summary of Issue Not Fully SR # F&O # Impact on SPRA Results Resolved Finding 24-20 was written to identify the lack of a systematic review of key assumptions.
This review has been complete and is documented.
This SR is met.
A.8. Identification of Key Assumptions and Uncertainties Relevant to the SPRA Results.
The PRA Standard includes a number of requirements related to identification and evaluation of the impact of assumptions and sources of uncertainty on the PRA results. NUREG-1855 [14] and EPRI 1016737 [15] provide guidance on assessment of uncertainty for applications of a PRA. As described in NUREG-1855, sources of uncertainty include parametric uncertainties, modeling uncertainties, and completeness (or scope and level of detail) uncertainties.
- Parametric uncertainty was addressed as part of the RNP SPRA model quantification (see Section 5 of this submittal).
- Modeling uncertainties are considered in both the base internal events PRA and the SPRA. Assumptions are made during the PRA development as a way to address a particular modeling uncertainty because there is not a single definitive approach.
Plant-specific assumptions made for each of the Robinson Nuclear Power Plant SPRA technical elements are noted in the SPRA documentation that was subject to peer review, and a summary of important modeling assumptions is included in Section 5.
- Completeness uncertainty addresses scope and level of detail. Uncertainties associated with scope and level of detail are documented in the PRA but are only considered for their impact on a specific application. No specific issues of PRA completeness were identified in the SPRA peer review.
A summary of potentially important sources of uncertainty in the Robinson Nuclear Power Plant SPRA is listed in Table A-4.
Table A-4 Summary of Potentially Important Sources of Uncertainty Summary of Treatment of Sources Potential Impact on SPRA PRA Element of Uncertainty per Peer Review Results The seismic hazard reasonably Site-specific site response analyses reflects sources of uncertainty.
were performed to include the effects of local site response. Uncertainties in Regarding the peer review teams site response inputs were included in comment on the Robinson Dam the analyses. Therefore, RNP PSHA fragility, Duke provided the met the requirement to include effects explanation that was acceptable Seismic Hazard of local site response.
to the peer review team during the peer review week and the The focused peer review team dam fragility calculation was commented that a more thorough updated to include a more discussion of uncertainty in estimating through discussion of uncertainty.
fragility of the Robinson dam should be Thus, no changes were made to discussed.
the model.
Page 105 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-4 Summary of Potentially Important Sources of Uncertainty Summary of Treatment of Sources Potential Impact on SPRA PRA Element of Uncertainty per Peer Review Results Many sensitivity studies described in Section 5.7 of this report evaluate the impact of No specific peer review team changes to fragilities on the Seismic comments on sources of uncertainty in SPRA results as one means of Fragilities fragilities. assessing the impact of fragilities uncertainties on the SPRA results. No changes to the model were recommended based on these results.
A follow-up walkdown was performed to determine if the SDAFW Pump can be damaged if the TB Class III collapses away from the TB Class I. The The plant model assigns a 50 %
walkdown results showed that probability of SDAFW pump failure due there was a sufficient spatial to failure of the TB Class III, assuming separation from the closest point that the TB Class III collapses toward of the TB Class III to the SDAFW the TB Class I causing damage to the Pump skid and the turbine speed SDAFW pump.
Seismic PRA governor, steam piping and pump Model discharge line are well shielded The review team commented that one from the impact of the TB Class important assumption (i.e., SDAFW III failure. In addition, there were pump surviving the failure of the Class no soft targets whose failure III TB failure) is not fully addressed for could prevent pump function.
the associated uncertainty.
Based on this, the use of 50 %
probability of SDAFW pump failure due to failure of the TB Class III is justified. Thus, no changes were made to the model.
Page 106 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT A.9. Identification of Plant Changes Not Reflected in the SPRA The RNP SPRA reflects the plant as of the cutoff date for the SPRA, which was June 2015.
Table A-5 lists significant plant changes subsequent to this date and provides a qualitative assessment of the likely impact of those changes on the SPRA results and insights.
Table A-5 Summary of Significant Plant Changes Since SPRA Cutoff Date Description of Plant Change Impact on SPRA Results As all of these are the changes to the existing Transmission Upgrade Project installed switchyard that provides offsite power from the grid, new 115/230 kV SUTs, 4 kV the industry generic seismic fragility associated with Switchgear and associated equipment a Loss of Offsite Power (LOOP) that has been used for Bldg. 469 etc. in the Robinson SPRA still applies to them. Thus, no beneficial impact on the Robinson SPRA.
Page 107 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-2: Summary of Finding F&Os and Disposition Status Suggested SR F&O Description Basis Disposition Resolution Section 6.3 of the It was initially assumed that the station calculation for FLEX actions were independent of the Robinson Seismic the SSCs that drive the bin Probabilistic Risk adjustments. However, it was Assessment Human judged that this approach may be Reliability Analysis non-conservative for FLEX actions Notebook [65]
at higher bin levels in which an addresses the addition extended time window (i.e., greater of operator actions than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />) was not available.
related to use of FLEX SPR-D3 requires that the HRA Therefore, FLEX actions were equipment. FLEX account for relevant seismic-related For FLEX actions with a reviewed to determine which actions were evaluated effects on operator actions. The short time window (e.g., actions may require additional bin-with detailed HRA timing provided in the FLEX less than 2 hrs), consider specific modeling. First, a review of approach, except that a validation may be appropriate for quantifying separate SR C-SPR- the time window for the FLEX 19-1 bin context adjustment low seismic hazards but should be actions based on actions was performed. Those D3 was not used. Instead, adjusted to account for challenges different seismic hazard FLEX actions with a Tsw 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> FLEX actions were from increasing seismic hazard levels. were excluded from further considered to apply levels. consideration, specifically OPER-across all seismic bins.
64, -67, -68 and -69. Only OPER-The rationale for not 61 meets the criteria with a time using seismic bins is window of less than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> since it that the SSCs that drive has a Tsw of 61 minutes. To the bin adjustments address the finding, seismic HRA become irrelevant for bin-specific variations of OPER-61 FLEX.
were developed, based on the bin adjustments detailed in Table 8.4 of However, this fails to the station calculation for the account for the Robinson Seismic Probabilistic increasingly challenging Risk Assessment Human Reliability plant context created by Analysis Notebook [65]
failures at higher Page 108 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-2: Summary of Finding F&Os and Disposition Status Suggested SR F&O Description Basis Disposition Resolution seismic accelerations. It The HRA Calculator now contains may be possible to detailed evaluations for OPER argue that actions such S1, S2, S3 and S5. The resulting as OPER-69 with long HEPs are included in the Table time windows are 10.1 of the station calculation for independent of seismic the Robinson Seismic hazard level. However, Probabilistic Risk Assessment for actions such as Human Reliability Analysis OPER-61 (SG makeup)
Notebook [65].
with a relatively short time window, variability of operator reliability by Therefore, this does not represent hazard level should be a change in methodology, scope, or considered. capability as defined in Appendix 1-A of the ASME/ANS PRA Standard and is not considered an upgrade.
The response to this Finding meets the requirements of NTTF 2.1 seismic and Capability Category II of the Standard with NRC Clarification. This finding is considered resolved for the purposes of NTTF 2.1.
The station calculation 1.Table 8.4 in the station 1. To address the finding, Bins 4 for the Robinson calculation for the Robinson Address the and 6 have been retained in Table SR C-SPR- Seismic Probabilistic Seismic Probabilistic Risk documentation issues. 8.4 for completeness but have 19-2 Risk Assessment Assessment Human Reliability been shaded to reflect that they are F1 Human Reliability Analysis Notebook [65] documents screened from the analysis. HFEs Analysis Notebook [65] the PSF Adjustments for Seismic for these bin values have been and the station HFEs for the six HRA bins. deleted from the HRA Calculator calculation for the However, Bin 4 is defined for only and the Table 10.1 of the station Page 109 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-2: Summary of Finding F&Os and Disposition Status Suggested SR F&O Description Basis Disposition Resolution Robinson Seismic one acceleration value (0.75g) and calculation for the Robinson Probabilistic Risk this bin is apparently combined with Seismic Probabilistic Risk Assessment Bin 5 (defined for 0.75g or higher). Assessment Human Reliability Quantification Notebook In addition, Bin 6 is defined as Analysis Notebook [65]
[13] 0.40g or higher, with overlaps with include some specific Bins 1 to 5. The Note 1 explains 2. The HFE OPER-99 has been issues that do not that Bin 6 is screened due to the deleted from the station calculation facilitate peer reviews building failures at this same for the Robinson Seismic and future uses and hazard level. It appears that Bin 4 Probabilistic Risk Assessment upgrades. and Bin 6 should be removed from Quantification Notebook [13].
the HRA bin definitions.
- 2. Action OPER-99 is in the station 3. The station calculation for the calculation for the Robinson Containment Analysis [76] was Seismic Probabilistic Risk reviewed to identify the following Assessment Quantification HFEs that are embedded in the Notebook [13] but not the station Level 2 quantification: OPER-ILI, calculation for the Seismic OPER-ISOL, OPER-IV, and OP-Equipment List [64] or the station H2REC. OPER-ISOL is already set calculation for the Robinson to 1.0 and was screened from Seismic Probabilistic Risk seismic evaluation. The other Assessment Human Reliability events had Tsws less than 4 Analysis Notebook [65] hours, so seismic HRA bin values Based on discussions with the were calculated for them. The HRA Duke team, the PR team Calculator now contains understands that OPER-99 was a quantifications of -S1, S2, S3 and placeholder human failure event S5 values for OPER-ILI, OPER-IV that is not needed and should be and OP-H2REC. The resulting deleted from the Quantification HEPs are included in the station Notebook [13]. calculation for the Robinson Seismic Probabilistic Risk
- 3. Based on discussions with the Assessment Human Reliability Duke team, the PR team Analysis Notebook [65] Table 10.1.
Page 110 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-2: Summary of Finding F&Os and Disposition Status Suggested SR F&O Description Basis Disposition Resolution understands that the level 2 model includes some embedded operator Therefore, this does not represent actions that are not included in the a change in methodology, scope, or station calculation for the capability as defined in Appendix 1-Robinson Seismic Probabilistic A of the ASME/ANS PRA Standard Risk Assessment Human Reliability and is not considered an upgrade.
Analysis Notebook [65]. These are The response to this Finding meets longer term actions compared to the seismic event and, thus, the the requirements of NTTF 2.1 impact from the seismic event is seismic and Capability Category II not expected to be significant. of the Standard. This finding is However, this screening analysis considered resolved for the has not been documented. purposes of NTTF 2.1.
The entire list of seismic HFEs was Several operator actions reviewed in HRA Calculator against that were reviewed For all operator actions, the Table 8.4 of the station against the PSF revisit the application of calculation for the Robinson Adjustments for Seismic PSF Adjustment rules to Seismic Probabilistic Risk HFEs defined in Tables The rules for PSF Adjustments for verify they have been Seismic HFEs should be Assessment Human Reliability 8.3 and 8.4 of the applied consistently. If completely defined and used, with Analysis Notebook [65] detailed SR C-SPR- station calculation for additional rules are 19-3 the Robinson Seismic any exceptions clearly seismic HFE adjustment rules.
D3 needed for exceptions to During this review, the rules were Probabilistic Risk documented. the current rules, either applied as-is or diversions Assessment Human document those rules from the rules were specifically Reliability Analysis and provide a basis to noted on the first (BE Data) screen Notebook [65]. Some support them. of the HFE. An internal review was inconsistent were also conducted to ensure identified with the correctness and concurrence with application of these the seismic HRA bin rule rules.
application and documentation.
Page 111 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-2: Summary of Finding F&Os and Disposition Status Suggested SR F&O Description Basis Disposition Resolution OPER-03 (implement Some instances were also feed and bleed) was identified during the seismic HFE modeled using the five review where the rules themselves seismic HRA bins. were refined; these changes were Several minor made to Table 8.4 in the station deviations were calculation for the Robinson identified from the PSF Seismic Probabilistic Risk adjustment rules. The Assessment Human Reliability S0 version (OPER Analysis Notebook [65].
S0) uses moderate Updated HEPs resulting from this Stress, while S1, S2 and review are shown in Table 10.1.
S3 use high Stress, inconsistent with the Therefore, this does not represent rules. For the S1 a change in methodology, scope, or version, the Tdelay is capability as defined in Appendix 1-the same as S0, but A of the ASME/ANS PRA Standard should be 2 min longer. and is not considered an upgrade.
For the S4 and S5 The response to this Finding meets versions, the cause the requirements of NTTF 2.1 decision tree Pca use seismic and Capability Category II branch d rather then e. of the Standard with NRC Clarification. This finding is OPER-28 (provide considered resolved for the alternate cooling to purposes of NTTF 2.1.
CCW) was modeled using the five seismic HRA bins for ex-MCR actions. Several minor deviations were identified from the PSF adjustment rules. The S0 version (OPER S0) uses moderate Page 112 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-2: Summary of Finding F&Os and Disposition Status Suggested SR F&O Description Basis Disposition Resolution Stress, while S1, S2 and S3 use high Stress, inconsistent with the rules. For the S0 version, the Tcog is 15 min, but Tcog values for the S1 to S5 versions seem to be based on an S0 value of 16 min (then modified according to Table 8.3 of the station calculation for the Robinson Seismic Probabilistic Risk Assessment Human Reliability Analysis Notebook [65]
OPER-14 (start deepwell pumps for AFW source) was modeled using the five seismic HRA bins (e.g.,
OPER-14-S1).
However, this action has a large time margin (330 min) and Table 8.4 of the station calculation for the Robinson Seismic Probabilistic Risk Assessment Page 113 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-2: Summary of Finding F&Os and Disposition Status Suggested SR F&O Description Basis Disposition Resolution Human Reliability Analysis Notebook [65]
includes the rule: If Time Margin is > 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, no seismic bin values are calculated. Based on the long time margin, it is not clear why seismic-impacted HEPs were calculated.
OPER-49 (manually initiate SI) was modeled using the five seismic HRA bins. However, For this action, only minor modifications were made to the HEPs (i.e.,
adding 2 minutes to Tdelay). It is not clear why the standard PSF adjustments have not been used. Perhaps because this is a memorized action with very limited time window (10 mins), unique rules have been used. If so, they have not been documented with a basis.
Page 114 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-2: Summary of Finding F&Os and Disposition Status Suggested SR F&O Description Basis Disposition Resolution This SR requires the This SR requires the assessment of assessment of Revise Section C in the accessibility as impacted by accessibility as station calculation for the seismic hazard level. The PSF impacted by seismic Seismic Equipment List adjustments in Table 8.3 of the hazard level. The PSF [64] to include paths station calculation for the Adjustments in Table taken for each local Robinson Seismic Probabilistic 8.3 of the station operator action, including Risk Assessment Human Reliability calculation for the FLEX actions. In Analysis Notebook [65] provide an Robinson Seismic addition, provide a increase in Texe to account for Probabilistic Risk conclusion for each local generic delays in access. However, Assessment Human action with regard to no detailed assessments of Reliability Analysis Assessment of accessibility for ex- feasibility (based operator pathways for ex-MCR Notebook [65] MCR actions is critical to determine accessibility) and actions were documented. Table SR C-SPR- feasibility of these actions at C-1 of the station calculation for the 19-4 provide an increase in whether the timing-D3 Texe to account for different seismic hazard levels. adjustment factors Seismic Equipment List [64]
generic delays in provided in the station provides a description of the access. However, no calculation for the location that ex-MCR actions are detailed assessments of Robinson Seismic performed. However, this table operator pathways for Probabilistic Risk does not provide a description of ex-MCR actions were Assessment Human the pathways to the action location.
documented. Reliability Analysis Also, it does not provide a Notebook [65] conclusion regarding whether the Table C-1 in the station are sufficient to account action is feasible based on calculation for the for delays in operator accessibility and whether the Seismic Equipment List transit. timing-adjustment factors provided
[64] provides a in the station calculation for the description of the Robinson Seismic Probabilistic location that ex-MCR Risk Assessment Human Reliability actions are performed. Analysis Notebook [65] are Page 115 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-2: Summary of Finding F&Os and Disposition Status Suggested SR F&O Description Basis Disposition Resolution However, this table sufficient to account for delays in does not provide a operator transit. In addition, Table description of the C-1of the station calculation for the pathways to the action Seismic Equipment List [64] does location. Also, it does not address the locations and not provide a conclusion pathways for performing FLEX regarding whether the actions (e.g., OPER-61).
action is feasible based Resolution: Telephone interviews on accessibility and were conducted with Operations whether the timing-personnel from RNP. Operator adjustment factors action pathways were discussed, provided in the station and documentation has been calculation for the updated in the station calculation Robinson Seismic for the Robinson Seismic Probabilistic Risk Probabilistic Risk Assessment Assessment Human Human Reliability Analysis Reliability Analysis Notebook [65] to describe these Notebook [65] are pathways in detail. Further, a sufficient to account for conclusion regarding the feasibility delays in operator of these actions has been added to transit.
the aforementioned notebook.
In addition, Table C-1 of the station calculation Therefore, this does not represent for the Seismic a change in methodology, scope, Equipment List [64]
or capability as defined in does not address the Appendix 1-A of the ASME/ANS locations and pathways PRA Standard and is not for performing FLEX considered an upgrade. The actions (e.g., OPER-61).
response to this Finding meets the requirements of NTTF 2.1 seismic and Capability Category II of the Standard with NRC Clarification.
Page 116 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-2: Summary of Finding F&Os and Disposition Status Suggested SR F&O Description Basis Disposition Resolution This finding is considered resolved for the purposes of NTTF 2.1.
The second documented A review of the internal events assumption in Section 4.0 of the model and supporting station calculation for the H.B.
Robinson Seismic Probabilistic documentation was performed in Risk Assessment Model Notebook order to ensure applicability of the
[77] says: internal events modeling assumptions to the SPRA, as "The Internal Events PRA is used discussed in Section 4.0 of the as the technical basis for both CDF station calculation for the H.B.
and LERF. All assumptions and Robinson Seismic Probabilistic Documentation of the success criteria in the Internal Generate adequate Risk Assessment Model Notebook review of internal events Events PRA are retained in the documentation to confirm [77]. However, the peer reviewer (SR C-SPR- assumptions SPRA for the portions of the applicability of the deemed this documentation B1/C-SPR- 24-1 applicability to the sequence models that apply [This underlying internal inadequate per the Basis of F2 SPRA is missing. assumption provides continuity events for seismic. Significance.
between the Internal Events PRA and the SPRA. Any future changes to the Internal Events PRA success One approach to resolve this F&O criteria would be addressed as part is to disposition every assumption of the maintenance and update pertaining to the internal events process of the integrated PRA." model and any supporting calculations. This approach was Based on the answers to Peer taken and a list of the assumptions Review questions, this assumption with dispositions has been added to was meant to imply that a review of Section 5.3.8 of the station the assumptions associated with calculation for the H.B. Robinson the internal events model was Seismic Probabilistic Risk performed, and that all Page 117 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-2: Summary of Finding F&Os and Disposition Status Suggested SR F&O Description Basis Disposition Resolution assumptions related to the portions Assessment Model Notebook [77].
of the internal events model that This documentation is considered apply to the SPRA were retained. more than adequate.
The only documented evidence of No changes to the SPRA model the review of the assumptions is were required based on the the quoted assumption in the reviewed assumptions.
station calculation for the H.B.
Robinson Seismic Probabilistic Therefore, this does not represent Risk Assessment Model Notebook a change in methodology, scope,
[77]. The review was or capability as defined in verified/internally reviewed as part Appendix 1-A of the ASME/ANS of model development and during PRA Standard and is not cutset reviews of the model results. considered an upgrade. The response to this Finding meets the This high level statement does not requirements of NTTF 2.1 seismic provide adequate documentation of and Capability Category II of the the applicability of the internal Standard with NRC Clarification.
events model to SPRA. The review This finding is considered resolved of the assumptions associated with for the purposes of NTTF 2.1.
the underlying model is a critical step and absence of its documentation does not allow for appropriate review or verification even from the internal reviewer to confirm that the appropriate modeling changes have been made (see part b of the SR).
Note also that an assumption that, for example, resulted in not modeling specific components, Page 118 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-2: Summary of Finding F&Os and Disposition Status Suggested SR F&O Description Basis Disposition Resolution would be difficult to capture in a cutset review.
This F&O is linked to SPR-B1 because any assumption that is not considered valid needs to be explained (or the model appropriately modified) to fully meet SPR-B1. Absent appropriate documentation it is not possible to fully confirm this, although there is no evidence of anything inappropriately modeled.
The effects of refining the hazard Section C.6 of the station bins were investigated and calculation for the Robinson discussion has been added to Seismic Probabilistic Risk Investigate the effects of Section C.6 of the station Assessment Quantification a more refined binning of calculation for the Robinson Notebook [13]. says: the hazard curve at lower Seismic Probabilistic Risk g levels.
There is no optimization Assessment Quantification SR C-SPR- "Ten (10) hazard intervals provide of the hazard bins used Notebook [13] It was determined E3/C-SPR- 24-2 quantification. a very reasonable compromise If the current binning is that increasing the number of total E7 between quantification precision, retained confirm that this bins has little impact on the model maintenance, and is not overestimating CDF/LERF and importance quantification processing significantly CDF at lower measures results, while leading to challenges. If more hazard intervals g levels.
much longer quantification and were built into the model, the importance measure calculation overall calculated SCDF and times. Therefore, ten hazard bins SLERF may reduce by a few are still used for quantification.
percent, depending upon the Page 119 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-2: Summary of Finding F&Os and Disposition Status Suggested SR F&O Description Basis Disposition Resolution interval slicing and number of Similarly, it was determined that intervals". rebinning the hazard curve so that the lower acceleration levels are In response to a Peer Review separated into more bins has little question, it was concluded that a impact on the CDF/LERF and sensitivity to show stability of the importance measures results; SPRA results with respect to the however, the model was revised to number and size of hazard place more bins at the lower intervals was not performed. A accelerations and to condense the sensitivity (Case IE-1c) to show the higher accelerations into fewer potential change in SCDF and bins. This rebinning was performed SLERF that would result from in order to provide additional adding more intervals was intervals where CCDP/CLERP is performed and documented in the less than 1.0 and to be more station calculation for the consistent with common practice.
Robinson Seismic Probabilistic Risk Assessment Uncertainty and Therefore, this does not represent Sensitivity [70]. From figure 4-3 of a change in methodology, scope, or the station calculation for the capability as defined in Appendix 1-Robinson Seismic Probabilistic A of the ASME/ANS PRA Standard Risk Assessment Quantification and is not considered an upgrade.
Notebook [13]. it is evident that the The response to this Finding meets CCDP of the plant goes to 1.0 the requirements of NTTF 2.1 (except for the plant availability seismic and Capability Category II factor) after the 3rd hazard interval. of the Standard with NRC After that, the CDF contribution is Clarification. This finding is essentially the hazard curve. There considered resolved for the are therefore only 3 intervals that purposes of NTTF 2.1.
meaningfully describe the plant response to the seismic event. This is not consistent with common practice, which is to generate a more refined binning of the hazard Page 120 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-2: Summary of Finding F&Os and Disposition Status Suggested SR F&O Description Basis Disposition Resolution in the interval that is more meaningful for the quantification.
This can also overestimate the current risk profile.
This F&O is written against back-referenced SRs QU-B1, B2 and B3 because this limitation (B1) is not appropriately addressed and the truncation (B2 and B3), which in seismic PRA is to be intended as a more generic stability of the results, including bin numbers and size, is not fully investigated. The F&O is also applicable to SPR-E7 for back-referenced SR QU-E4 because this uncertainty in the quantification process is not assessed (note that adding a single hazard bin at the end is meaningless if CCDP is already 1.0).
Perform a systematic Section 5.3 of the SPRAIG [11].
Section 5.3.7.2 of the station review of other external was used as guidance to perform a calculation for the H.B. Robinson hazards (beyond the systematic review of other external Incomplete identification Seismic Probabilistic Risk secondary hazards that hazards that may have a seismic-of seismically-induced Assessment Model Notebook [77]
SR C-SPR- have been addressed) 24-4 consequential events. discusses Seismic Hazards Other equivalent. Each identified external A2 that may have a seismic-Than Vibratory Ground Motion but hazard was then dispositioned for equivalent and address there is no apparent discussion of inclusion in the SPRA.
any events that may not the potential for seismic-induced be screened out.
equivalent of other external Section 5.3.7.3 was added to the Page 121 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-2: Summary of Finding F&Os and Disposition Status Suggested SR F&O Description Basis Disposition Resolution initiators (e.g., seismic failure of Section 5.3 of the station calculation for the H.B.
nearby major gas lines). SPRAIG [11] can be Robinson Seismic Probabilistic used as guidance to Risk Assessment Model Notebook Based on Duke's answer to a Peer perform this task [77] in order to document this Review question, the IPEEE review. For each identified hazard, analysis is available for Robinson.
a determination was made whether The information was reviewed to to include the hazard in the SPRA support the SPRA. Formal review was performed as part of the and a disposition is provided.
station calculation for the Seismic Equipment List [64], but not As stated in the Basis for specifically to screen out the Significance, multiple hazards were potential for other seismic-induced included in the SPRA. No additional external initiating events. hazards were modeled in the SPRA A screening analysis of other as a result of this review.
seismic hazards was performed specifically for the SPRA. Section Therefore, this does not represent 11 of Attachment 1 to the station a change in methodology, scope, or calculation for the Final Seismic capability as defined in Appendix 1-Analysis Report [16] provides the A of the ASME/ANS PRA Standard screening of other seismic hazards and is not considered an upgrade.
including, fault displacement, The response to this Finding meets landslides, soil liquefaction, soil the requirements of NTTF 2.1 settlement, earthquake-induced seismic and Capability Category II flooding, fracking-induced of the Standard with NRC earthquakes, and seismic seiches Clarification. This finding is (this addresses hazards identified in SHA-I2). All other seismic considered resolved for the hazards were screened from the purposes of NTTF 2.1.
SPRA except soil liquefaction and soil settlement.
Page 122 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-2: Summary of Finding F&Os and Disposition Status Suggested SR F&O Description Basis Disposition Resolution Section 5.3.7.2 of the station calculation for the H.B. Robinson Seismic Probabilistic Risk Assessment Model Notebook [77]
discusses only lateral spreading and liquefaction-induced settlement failures since these were not screened for RNP. The station calculation for the Robinson Nuclear Power Plant Seismic Induced Flood and Fire Assessment [69] analyzes seismic-induced internal fires and floods as a result of seismically failed SSCs.
The station calculation for the H.B. Robinson Seismic Appendix B of the station Probabilistic Risk Assessment calculation for the H.B. Robinson Ensure that there is Seismic Probabilistic Risk Model Notebook [77] Appendix B is adequate documentation Assessment Model Notebook [77]
a report from the FRANX fragility to of the mapping of the documents the FRANX Incomplete Component table. This report of the seismic failures (i.e., Fragility_To_Comp table, which documentation of the mapping reflects the mapping used fragility groups) to basic provides the BEs mapped to each SR C-SPR- mapping of seismic in the model but per se does not events included (or fragility group. The mapping is B3/C-SPR- 24-6 failures to existing basic document the mapping (i.e., it does appropriately added) in straightforward (e.g., the RWST F2 events. not provide reason for the the model when the fragility group is mapped to the mapping).
mapping is not failure of the RWST BE in the straightforward. model). Some items were identified While it is recognized that a large portion of the mapping is in the Basis for Significance that straightforward (i.e., a fragility made it appear the mapping was group for a single component or not as straightforward; these items two unequivocally identical are dispositioned below.
Page 123 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-2: Summary of Finding F&Os and Disposition Status Suggested SR F&O Description Basis Disposition Resolution components is mapped only to the Additionally, all group to BE basic events for those components) mappings were reviewed and no some other mapping is less other issues were identified.
straightforward and needs to be explained. This is particularly 1. Fragility group SF-PM-CHG-evident for 2 over 1 failures PMP is no longer mapped to mapping. initiator %S1. This was an original conservatism in the model that An example of this is the mapping should have been removed of the fragility group for the following the addition of the charging pump to the %S1 initiator. VSLOCA logic in the model.
Based on Duke answer to the peer review question, '"the mapping of
- 2. Fragility group SF-PA-HAGAN is the charging pump fragility group to now only mapped to Dummy_CD.
initiating event %S1 (the fragility This group was originally mapped group is not mapped to %S2) was to a specific BE in the model but an original conservatism that was then updated to be mapped to should have been removed Dummy_CD; however, the mapping following the addition of the very to BE SF_PA_HAGAN was small LOCA (VSLOCA) logic in the inadvertantly not removed from model. Originally it was assumed FRANX. Additionally, BE that a loss of all charging pumps SF_PA_HAGAN has been removed would result in VSLOCA conditions from the model as it is no longer requiring long-term RCS makeup.
used. Hagan Racks have been Due to the high capacity of the added to the list in Section 5.3.6 of charging pumps, this conservatism the station calculation for the H.B.
has negligible impact on the Robinson Seismic Probabilistic results."
Risk Assessment Model Notebook Another example is the SF-PA- [77]
HAGAN fragility group, that is modeled both to the DUMMY_CD 3. Fragility group SF-MV-RCP-and to other specific basic events, CLNG-MOV is no longer mapped to Page 124 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-2: Summary of Finding F&Os and Disposition Status Suggested SR F&O Description Basis Disposition Resolution without explicitly showing up in Dummy_CD. This group was section 5.3.6 of the station updated to be mapped to specific calculation for the H.B. Robinson BEs instead of Dummy_CD; Seismic Probabilistic Risk however, the mapping to Assessment Model Notebook [77] Dummy_CD was inadvertantly not nor anywhere else in the notebook removed from FRANX.
where a description of the reason of this double mapping is provided. Therefore, this does not represent a change in methodology, scope, or Absence of this documentation capability as defined in Appendix 1-makes the internal review and A of the ASME/ANS PRA Standard verification of the appropriate and is not considered an upgrade.
modeling (covered in SPR-B3)
The response to this Finding meets challenging.
the requirements of NTTF 2.1 seismic and Capability Category II of the Standard with NRC Clarification. This finding is considered resolved for the purposes of NTTF 2.1.
One direct to CD accident Document the rationale Discussion of the mapping to basic sequence was added to the logic for the mapping to event DUMMY_PDS has been (via mapping to the DUMMY_CD, DUMMY_PDS with a added as the second bullet in Documentation of the thus the relevance of this F&O to rationale for the selection Section 5.3.6. of the station modeling associated SPR-B8), and not all the of equipment.
SR C-SPR- calculation for the H.B. Robinson with DUMMY_PDS is contributions are appropriately E6/C-SPR- 24-7 missing. transferred to LERF. Justify not appropriately Seismic Probabilistic Risk F2 Assessment Model Notebook [77]
transferring CDF Duke answer to the peer review sequences to LERF a subset of those fragility groups question explained which subset of (e.g., a sensitivity can be mapped straight to core damage fragilities (among those mapped to provided where either are also mapped straight to large DUMMY_CD) is also mapped to DUMMY_PDS is early release by mapping the Page 125 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-2: Summary of Finding F&Os and Disposition Status Suggested SR F&O Description Basis Disposition Resolution DUMMY_PDS). There is no removed from the logic groups to BE DUMMY_PDS. While documentation of the rationale for or all the fragilities the mapping of these fragility this mapping beyond the mere mapped to DUMMY_CD groups directly to core damage is a reporting of the mapping. This are also mapped to known conservatism with little modeling is not straightforward and DUMMY_PDS). impact on the results, allowing the needs to be documented and failure of these SSCs to go to LERF clarified.
is overly conservative. Therefore, the DUMMY_PDS basic event is introduced to limit which fragility groups impact LERF.
No model changes were made in resolving this finding.
Therefore, this does not represent a change in methodology, scope, or capability as defined in Appendix 1-A of the ASME/ANS PRA Standard and is not considered an upgrade.
The response to this Finding meets the requirements of NTTF 2.1 seismic and Capability Category II of the Standard with NRC Clarification. This finding is considered resolved for the purposes of NTTF 2.1.
One important The catastrophic failure of the The survivability of the The plant model assigns a 50%
SR C-SPR- 24-8 assumption (i.e., Class III TB is modeled with a 50% SDAFW following any probability of SDAFW pump failure E7 SDAFW pump surviving chance of impacting the Class I TB failure of the Class III TB due to failure of the Class III the failure of the Class (apparently based on the prevalent is a potential uncertainty Turbine Building. In this scenario, Page 126 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-2: Summary of Finding F&Os and Disposition Status Suggested SR F&O Description Basis Disposition Resolution III TB failure) is not direction of the collapse). When the that should be the Class III Turbine Building addressed for the Class III TB falls towards the Class addressed. A collapses towards the Class I associated uncertainty. I TB, everything in the TB is conservative way would Turbine Building causing damage assumed to be failed (including the be to fail specifically to the SDAFW pump located in the SDAFW pump). When the Class III SDAFW every time the latter. A follow-up walkdown was TB falls away from the Class I TB, TB Class III fails. A performed to determine if the then the CST is impacted but the dedicated sensitivity SDAFW pump can be damaged if SDAFW pump is considered could be performed to the Class III Turbine Building unaffected. address this eventuality. collapses in some other direction; Additional justification e.g., away from the Class I Turbine It is noted that the SDAFW is a should probably be Building. The walkdown team couple of feet away from the provided to support the accessed the SDAFW pump skid invisible line that separates the current base case in term and reviewed and measured the Class I and Class III portion of the of the specific soft distances, envisioning how a TB. During the peer review targets observed during collapse might occur and its effect walkdown the team observed the peer review on the pump. The distance significant equipment of relatively walkdown. measured from the closest point of large dimension crossing the the Class III Turbine Building to the boundary between the two TB SDAFW pump skid was a minimum sections. The peer review team of eight feet. Equipment mounted also observed elements on the on the pump exposed to the Class SDAFW skid that could be judged III Turbine Building includes the as soft target (in disagreement with pump suction piping, various the original SEWS). It is at least associated process gauges and unsure that a collapse of the Class transmitters, some lube oil piping, III TB would leave the SDAWF and a control valve and tubing unaffected. associated with temperature control. The turbine speed In answering to the peer review governor, steam piping, and pump question, Duke pointed to a discharge line are located on the sensitivity performed on the split opposite side of the skid and were fraction (current 50-50 for the base judged to be shielded from case) which only partially elements of the collapsed Class III Page 127 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-2: Summary of Finding F&Os and Disposition Status Suggested SR F&O Description Basis Disposition Resolution addresses the concern of Turbine Building that might be damaging the SDAFW even if the present.
Class III TB does not interact significantly with the Class I TB Piping in the vicinity of the pump (i.e., damage to the SDAWF would passes across the boundary be due to ancillary failures and between the Class I and Class III interaction). Turbine Buildings. The larger piping includes a 20-in. diameter Duke also pointed to a December feedwater line and a 16-in.
2017 memo from the Fragility diameter line from the heater drain Vendor. Such memo, again, tank. This piping is separated from supports the 50% chance of the SDAFW pump by distances of interaction, but does not answer several feet. In addition, one of the the original question as page 2 of Class I Turbine Building columns the memo says: "The document located between the pump and the requests that the Frag, Show that feedwater line shields the former the SDAFW pump and all support from impact by the latter. The SSCs will survive seismic failure of walkdown team judges that the the Class I Turbine Building, available spatial separation and including any above ground shielding is sufficient to preclude suction, steam inlet, turbine damage to the SDAFW Pump even cooling, and pump discharge If the Class III Turbine Building piping, as well as DC power cables, were to fail in some direction other given seismic failure of the Class III than towards the Class I Turbine Turbine Building and subsequent Building.
interaction with the Class I Turbine Building. Based on the station During the walkdown, the AFW calculation for the Turbine Building system engineer was interviewed to Class III fragility) for the Class 3 better understand how damage to Turbine Building, we judge the the various components might Steam Driven Auxiliary Feedwater affect pump operability. He indicated that the pump/turbine assembly was modified so that it is Page 128 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-2: Summary of Finding F&Os and Disposition Status Suggested SR F&O Description Basis Disposition Resolution pump unlikely to survive failure of now self-cooling, and the control the Class 1 Turbine Building. valve and tubing that provided cooling previously have been abandoned in place. Therefore damage to these items would not affect the pumps operability. The gauges and transmitters on the suction pipe could very well be damaged and, given a breach in the line some water would be lost, but the overall volume of process fluid passing through the pumps would be minimally affected. This is consistent with the resolution to F&O 29-5, Item 3 that there are no soft targets whose failure could prevent pump function. Based on the above, the walkdown team judged that the probability of failure of the SDAFW pump given failure of the Class III Turbine Building is 50%. The SEWS for the SDAFW pump has been updated for the above and is in the station calculation for the Seismic Capacity Walkdown Report [53]. Additional information may be obtained from this SEWS.
Therefore, this does not represent a change in methodology, scope, or capability as defined in Appendix 1-A of the ASME/ANS Page 129 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-2: Summary of Finding F&Os and Disposition Status Suggested SR F&O Description Basis Disposition Resolution PRA Standard and is not considered an upgrade. The response to this Finding meets the requirements of NTTF 2.1 seismic and Capability Category II of the Standard with NRC Clarification.
This finding is considered resolved for the purposes of NTTF 2.1.
Some inconsistencies have been Resolve the Inconsistencies were identified observed between the station inconsistencies between between the SPRA model and the calculation for the H.B. Robinson the model and the change log provided as Table A-1 Seismic Probabilistic Risk change log. in the station calculation for the Assessment Model Notebook [77] H.B. Robinson Seismic Table A-1, that discusses the It is observed that issues Probabilistic Risk Assessment model log changes, and the actual have been observed on Model Notebook [77]. Furthermore, final model. the first two randomly selected entries of Table it was suggested that additional An example is gate CLASS-3-TB, A-1 of the station detail be provided for model Discrepancies between which is documented in entries calculation for the H.B. changes, as needed.
the model notebook log SR C-SPR- #34, 35, 36 and 37 of Table A-1 of Robinson Seismic change and the CAFTA B3/C-SPR- 24-13 the station calculation for the H.B. Probabilistic Risk The specific items provided in the model.
F2 Robinson Seismic Probabilistic Assessment Model Basis of Significance pertaining to Risk Assessment Model Notebook Notebook [77] . An gate CLASS-3-TB and the lateral
[77] , from which it seems justified extend of condition is spreading failure of the RWST, under the following gates in the recommended. have been corrected in Table A-1 model: of the station calculation for the Consider adding enough H.B. Robinson Seismic FMMSEGAMFN, FMMSEGBMFN, details in Table A-1 of Probabilistic Risk Assessment FMMSEGCMFN, FFLEX002. the station calculation for the H.B. Robinson Model Notebook [77] . An extent of Gate CLASS-3-TB is apparently Seismic Probabilistic condition review was performed, present in the model in multiple Risk Assessment Model and Table A-1 of the station other places (e.g., under gates Notebook [77] (either in calculation for the H.B. Robinson Page 130 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-2: Summary of Finding F&Os and Disposition Status Suggested SR F&O Description Basis Disposition Resolution FDGAFW-CFN-TB3, the notes or in the Seismic Probabilistic Risk ATK%%PACFF-TB3, QAVV1- description of the Assessment Model Notebook [77]
3CFF-TB3, etc.). change) to describe why was corrected as necessary to the change is done and match the SPRA model.
In response to the peer review not only that is done, and Additionally, the comments for question, the RNP S-PRA indicated if applicable point to a some modeling changes were that the other gates should also be section in the notebook enhanced in Table A-1 of the added in the change log. where the rationale is station calculation for the H.B.
further discussed. This Another example is entry #31 of would make easier to Robinson Seismic Probabilistic Table A-1 the station calculation for review and confirm than Risk Assessment Model Notebook the H.B. Robinson Seismic the modeling and the log [77] .
Probabilistic Risk Assessment are complete and Model Notebook [77] that does Therefore, this does not represent comprehensive.
not match with the model (the a change in methodology, scope, or model shows that gate capability as defined in Appendix 1-SEISMIC_SPREAD_DC2-3 is A of the ASME/ANS PRA Standard under some additional/different and is not considered an upgrade.
gates not reported in the notebook The response to this Finding meets (e.g., RWST-1). In response to the the requirements of NTTF 2.1 peer review question, the RNP S- seismic and Capability Category II PRA confirmed that the model of the Standard with NRC seems to be correct and the issue Clarification. This finding is remains in the documentation. considered resolved for the purposes of NTTF 2.1.
To meet back-reference SR QU- Present the quantification The station calculation for the Limited quantification A2, the quantification needs to be results in a more refined Robinson Seismic Probabilistic SR C-SPR- results provided.
24-15 performed at the sequence level as fashion (e.g., as Risk Assessment Quantification E3 applicable at least well as the top level. It is Notebook [13] has been revised to recognized that in a seismic PRA, classes of plant initiators provide the risk contributions from when multiple initiators are in (e.g., LOCAs, LOOPs, the top accident sequences for Page 131 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-2: Summary of Finding F&Os and Disposition Status Suggested SR F&O Description Basis Disposition Resolution theory possible, a quantification on others)) to meet the CDF and LERF. The accident a sequence by sequence basis requirement/intent of QU- sequence contributions are now may not be practical. Nevertheless, A2. provided in Section 4.1.3 of the the station calculation for the station calculation for the Robinson Seismic Probabilistic Robinson Seismic Probabilistic Risk Assessment Quantification Risk Assessment Quantification Notebook [13] has been revised to Notebook [13].
provide the risk contributions from Therefore, this does not represent the top accident sequences for a change in methodology, scope, or CDF and LERF. The accident capability as defined in Appendix 1-sequence contributions are now A of the ASME/ANS PRA Standard provided in Section 4.1.3.
and is not considered an upgrade.
Therefore, this does not represent The response to this Finding meets a change in methodology, scope, the requirements of NTTF 2.1 or capability as defined in Appendix seismic and Capability Category II 1-A of the ASME/ANS PRA of the Standard. This finding is Standard and is not considered an considered resolved for the upgrade. The response to this purposes of NTTF 2.1.
Finding meets the requirements of NTTF 2.1 seismic and Capability Category II of the Standard with NRC Clarification. does not discusses this limitation nor presents any additional break down of the results beyond the break-down based on g levels.
Page 132 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-2: Summary of Finding F&Os and Disposition Status Suggested SR F&O Description Basis Disposition Resolution To demonstrate convergence the quantification was modified. The details can be found in the station calculation for the Robinson Seismic Probabilistic Risk Assessment Quantification Notebook [13]. The results show that the percent change between the baseline and lower truncation The suggested rationale for the 5%
limit cases is approximately 6.0%. If difference between the base case the CLERP is assumed to be the LERF and the best achievable ACUBE code lower bound limit for truncation is based on a Use the deepest each of the hazard intervals for speculation and not supported by truncation scheme as which ACUBE could only process a Base case LERF is not any additional considerations. It is base case LERF or subset of the cutsets, then the converged per Standard also noted that the top cutsets are SR C-SPR- provide additional percent change would be 24-16 requirements. normally the ones that have much E3 justification that LERF is approximately 4.6%, which is less more of an impact in CDF/LERF converged. than 5%, indicating the results reduction when post-processed via would be close to the 5%
ACUBE, so some of the LERF convergence value if more cutsets increase may be real and not only could be evaluated using the related to ACUBE limitations.
ACUBE code.
Using the truncation values from the lower truncation case requires a much longer calculation time and does not meaningfully impact the risk results, the importance measures, or the top contributors.
Therefore, the model is considered to show convergence at the chosen truncation levels and is documented in the station Page 133 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-2: Summary of Finding F&Os and Disposition Status Suggested SR F&O Description Basis Disposition Resolution calculation for the Robinson Seismic Probabilistic Risk Assessment Quantification Notebook [13].
Therefore, this does not represent a change in methodology, scope, or capability as defined in Appendix 1-A of the ASME/ANS PRA Standard and is not considered an upgrade.
The response to this Finding meets the requirements of NTTF 2.1 seismic and Capability Category II of the Standard. This finding is considered resolved for the purposes of NTTF 2.1.
No specific model changes were QU-B5 requires to address circular made in order to address circular No discussion on logic. While CAFTA will logic, as circular logic was not circular logic automatically stop the introduced due to the incorporating Confirm that no model check/correction. quantification if circular logic is of the SPRA logic.
changes were made to detected, there is no SR C-SPR- address circular logic 24-17 documentation of whether any This statement has been added to E3 issues.
specific model change was Section 5.3.3 of the station performed to resolve any circular calculation for the H.B. Robinson logic issue. Seismic Probabilistic Risk Assessment Model Notebook [77]:
"No circular logic was introduced by the addition of the SPRA logic; Page 134 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-2: Summary of Finding F&Os and Disposition Status Suggested SR F&O Description Basis Disposition Resolution therefore, no model changes were made to address circular logic."
Therefore, this does not represent a change in methodology, scope, or capability as defined in Appendix 1-A of the ASME/ANS PRA Standard and is not considered an upgrade.
The response to this Finding meets the requirements of NTTF 2.1 seismic and Capability Category II of the Standard. This finding is considered resolved for the purposes of NTTF 2.1.
Back-referenced SR QU-B7 and QU-D3 requires to address The RNP mutually exclusive logic mutually exclusive logic and flags was reviewed for applicability to the potentially impacting the results. It SPRA and was determined to be is understood that any mutually applicable. Furthermore, the Confirm that the mutually mutually exclusive logic does not exclusive logic retained from the exclusive logic present in hinder failure propagation of the No evidence of mutually internal events model have been the S-PRA is applicable fragility groups.
exclusive logic check for carried through the in seismic PRA.
SR C-SPR- and that no fragility 24-18 seismic relevance.
E3 groups are prevented This statement has been added to There is no documented evidence from propagation. Section 5.3.3 of the station of any review performed to confirm that:1. The mutually exclusive logic calculation for the H.B. Robinson is appropriate also in case of Seismic Probabilistic Risk seismic, where correlation can Assessment Model Notebook [77]:
challenge the original logic;2. No "The mutually exclusive logic fragility group (especially those that contained in the Internal Events are mapped to multiple basic model has been reviewed and is Page 135 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-2: Summary of Finding F&Os and Disposition Status Suggested SR F&O Description Basis Disposition Resolution events) are prevented from applicable to the SPRA. The propagating through the logic from mutually exclusive logic does not the existing mutually exclusive prevent the propagation of seismic logic. failures for any fragility groups. No mutually exclusive logic was added Based on discussion with the RNP specifically for the SPRA."
S-PRA team, no mutual exclusive logic has been added explicitly for Therefore, this does not represent the S-PRA. a change in methodology, scope, or capability as defined in Appendix 1-A of the ASME/ANS PRA Standard and is not considered an upgrade.
The response to this Finding meets the requirements of NTTF 2.1 seismic and Capability Category II of the Standard. This finding is considered resolved for the purposes of NTTF 2.1.
There is no evidence of A formal cutset review was QU-D5 explicitly requires to review Provide evidence of a a review of non- performed which included a review non-significant cutsets. sufficient review of non-significant cutsets of cutsets from each truncation significant cutsets to consistent with current decade for both CDF and LERF; Review of non-significant cutsets is assess model expectation from the thus, a review of non-significant mentioned in the minutes of interim inconsistencies or SR C-SPR- NEI guidance [6] (i.e., a cutsets was performed. All cutsets 24-19 cutsets review but there is no simplifications.
E3 minimal number of were determined to be reasonable.
discussion on how many and how randomly selected This cutset review is documented deep in the model. Good practice would be cutsets for each as Section H.3 of the station to also document the decade). NEI guidance suggested the calculation for the Robinson non-significant cutsets minimal number of randomly that are reviewed, Seismic Probabilistic Risk Assessment Quantification Page 136 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-2: Summary of Finding F&Os and Disposition Status Suggested SR F&O Description Basis Disposition Resolution selected cutsets for each decade to although it is understood Notebook [13] and is discussed in be addressed. not to be a requirement. Section 4.1.4.2 of the station calculation for the Robinson Seismic Probabilistic Risk Assessment Quantification Notebook [13].
Therefore, this does not represent a change in methodology, scope, or capability as defined in Appendix 1-A of the ASME/ANS PRA Standard and is not considered an upgrade.
The response to this Finding meets the requirements of NTTF 2.1 seismic and Capability Category II of the Standard.
Systematically process A systematic review of key Provide a systematic review and all the assumptions assumptions supporting the SPRA uncertainty estimate of the supporting the S-PRA was performed. All aspects of the assumptions supporting the S-through the process SPRA were reviewed including the PRA. This should include:
No systematic review of discussed in QU-E4. seismic hazard assessment, the specific assumptions for seismic fragility assessment, the
- 1. Assumptions from the SHA SR C-SPR- their impact on the risk It is suggested that what seismic logic model development, assessment (because the SSHAC E7/C-SPR- 24-20 profile and the Duke judges are the key and the seismic HRA. The process already converts epistemic F3 associated insights. assumptions supporting uncertainty associated with all uncertainties in the fractiles used the different aspect of the identified key assumptions is for the quantification, it is not RNP S-PRA (see points characterized; sensitivity analyses expected that all the assumptions 1 through 6) are initially were performed to evaluate the are considered, but only those that collected in the subject uncertainty as applicable. This can be addressed with a potential summary reports, and review and uncertainty evaluation is sensitivity);
that the uncertainty documented in Section 3.1 and report addresses the associated sensitivity analyses are Page 137 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-2: Summary of Finding F&Os and Disposition Status Suggested SR F&O Description Basis Disposition Resolution
- 2. Assumptions from the SFR uncertainty associated detailed in Section 4 of the station evaluation. F&O 24-8 is an with them in a systematic calculation for the Robinson example of this. way. Seismic Probabilistic Risk Assessment Uncertainty and
- 3. Assumption from the underlying Consistent with the Sensitivity [70]
internal events model (at least the expectations form back-Therefore, this does not represent assumption that were recognized to referenced SR QU-E4, a change in methodology, scope, or have model uncertainties in the IE the uncertainty capability as defined in Appendix 1-uncertainty assessment should be assessment for each A of the ASME/ANS PRA Standard re-visited in the S-PRA. assumption may have and is not considered an upgrade.
multiple elements, and
- 4. Assumption from the S-PRA The response to this Finding meets there is no expectation or logic model development; the requirements of NTTF 2.1 requirement that a seismic and Capability Category II dedicated sensitivity
- 5. Assumptions from the seismic of the Standard with NRC analysis is performed for HRA Clarification. This finding is each assumption.
considered resolved for the
- 6. Assumption from the It is suggested that, for purposes of NTTF 2.1.
quantification process (the hazard each sensitivity that is binning addresses this in some indeed performed for this extent). F&O 24-2 is an example of purpose, not only the this. overall CDF/LERF results are reported, but any indication of changes in insights (e.g.,
unexpected fluctuation of importance measures, etc)
It is also suggested that the range of CDF/LERF calculated based on sensitivities are Page 138 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-2: Summary of Finding F&Os and Disposition Status Suggested SR F&O Description Basis Disposition Resolution overlapped with the parametric uncertainty range to assess how CDF/LERF may change in complex. This would be valuable in a risk-aggregation perspective.
Although the Late time is not explicitly defined with a single evacuation time value, the MAAP Justify that the binning of results show the non-LERF/late CDF sequences into release categories have a range of LERF PDS is still 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> to 25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> between vessel applicable for higher g failure and containment failure.
levels where evaluation Other non-LERF may be early or There is no evidence that the RNP time can be challenged late categories but have small No timing change S-PRA team addressed the need to (i.e., the definition of release magnitude and thus are not considerations for SR C-SPR- revisit the binning of CDF to LERF Early should be LERF. Thus, all late release 24-21 LERF. categories are also medium or E6 sequences. confirmed applicable).
small in magnitude, except for one Re-bin any non-LERF release category which is large in sequences into LERF if magnitude but has ~20 hours the timing can be between vessel failure and impacted. containment failure, which is sufficient for evacuation, so no late release categories need to be re-binned to LERF for the RNP SPRA.
Further details on this F&O can be found in the station calculation for Page 139 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-2: Summary of Finding F&Os and Disposition Status Suggested SR F&O Description Basis Disposition Resolution the RNP SPRA Peer Review Resolution [71]
Therefore, this does not represent a change in methodology, scope, or capability as defined in Appendix 1-A of the ASME/ANS PRA Standard and is not considered an upgrade.
The response to this Finding meets the requirements of NTTF 2.1 seismic and Capability Category II of the Standard. This finding is considered resolved for the purposes of NTTF 2.1.
The station calculation for the Robinson Seismic Probabilistic Risk Assessment Quantification Notebook [13] has been revised to provide the risk contributions from the top plant damage states for Provide a LERF result LERF. The plant damage state LE-F1 CC-II requires a break-down contributions are now provided in No LERF breakdown breakdown based on SR C-SPR- of the LERF results based on PDS. Section 4.1.3 of the station 24-22 per PDS PDS.
E6 calculation for the Robinson Seismic Probabilistic Risk Assessment Quantification Notebook [13].
Therefore, this does not represent a change in methodology, scope, or capability as defined in Appendix 1-A of the ASME/ANS PRA Standard and is not considered an upgrade.
Page 140 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-2: Summary of Finding F&Os and Disposition Status Suggested SR F&O Description Basis Disposition Resolution The response to this Finding meets the requirements of NTTF 2.1 seismic and Capability Category II of the Standard. This finding is considered resolved for the purposes of NTTF 2.1.
Performing a LERF analysis per NUREG/CR-6595 [91] is an acceptable methodology as referenced in Reg. Guide 1.174 Rev 3 [92]. The advantage of this approach is that it allows LERF to be calculated quickly, though approximately, without the need for performing a detailed Level-2 PRA The seismic PRA uses the internal and the NRC has explicitly events LERF as basis. A number of accepted this approach as being Underlying internal back-referenced SRs are only met Resolve findings in the sufficient for the determination of events LERF SRs met SR C-SPR- at CC-I in internal events LERF and underlying LERF study. LERF.
24-23 at CC-I.
E6 no seismic specific changes are The Robinson LERF analysis made.
employs a methodology fully endorsed by the NRC as an acceptable means of calculating LERF for use in risk-informed applications, including changes to its licensing basis as well as its response to NTTF 2.1 seismic.
This does not represent a change in methodology, scope, or capability as defined in Appendix 1-A of the ASME/ANS PRA Standard Page 141 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-2: Summary of Finding F&Os and Disposition Status Suggested SR F&O Description Basis Disposition Resolution
[5 and 37] and is not considered an upgrade. The response to this finding meets the requirements of NTTF 2.1 seismic and Capability Category I of the Standard. This finding is considered resolved for the purposes of NTTF 2.1.
No plant-specific seismic events have occurred.
Review of relevant seismic risk SPR-A3 requires a The self assessment states that evaluations from other plants was review of plant specific Industry documents such as the performed. A review of the Surry response to any past SPRAIG [11] and SPID [2] were and Oconee SPRAs was performed seismic event. The RNP reviewed to ensure that the list of in order to ensure all initiating SPRA documentation initiating events included in the events are accounted for in the does not include a Robinson SPRA accounts for Perform a review of plant RNP SPRA. These evaluations discussion of any industry experience. specific events (if considered events, such as, applicable site specific applicable) and relevant seismic-induced flooding, seismic-SR C-SPR- events (or relevant 25-1 Beyond reference to the industry seismic risk evaluations induced fires, and seismic-induced A3 industry events).
documents in the self assessment, from other plants. failures of structures and dams, Although mention is no specific review of other SPRAs which are similarly evaluated in the given to the SPID [2]and was included. No documentation of RNP SPRA. No new initiating SPRAIG [11] in the self the findings of the industry events or accident sequences were assessment, no document review with respect to included in the RNP SPRA based discussion was found in this requirement were found. on the review of the Surry and the SPRA notebooks.
Oconee SPRAs.
Discussion on this review has been added to Section 5.2.1 of the plant station calculation for the H.B.
Robinson Seismic Probabilistic Page 142 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-2: Summary of Finding F&Os and Disposition Status Suggested SR F&O Description Basis Disposition Resolution Risk Assessment Model Notebook
[77]
Therefore, this does not represent a change in methodology, scope, or capability as defined in Appendix 1-A of the ASME/ANS PRA Standard and is not considered an upgrade. The response to this Finding meets the requirements of NTTF 2.1 seismic and Capability Category II of the Standard.
Because of the uncertainty in the Based on the capacity of the anchorage for the information in the DS-BUS and the potential for FRANX database, component DS-BUS has SPRA-A2 requires that a Review the potential for a interaction between adjacent systematic process be performed seismically induced bus panels, the DS-BUS was ranked a representative HCLPF of 0.10, well below the to identify initiators caused by duct fire associated with Low (L) regardless of seismic secondary hazards. Not assessing DS-BUS and identify the deficiencies identified during the screening HCLPF of SR C-SPR- the potential for fire due to arc flash consequences of such a seismic capacity walkdown [53].
25-2 0.75. The potential for A2 from an electrical bus failure may fire if it can't be screened This resulted in a judgment-based and consequences of result in omission of a valid based on frequency lower bound fragility with a HCLPF seismically induced bus initiating event. alone. capacity of 0.1 g PGA.
arcing fire was not assessed for this This fragility information was first component.
used for defining the DS DG System Fragility Group, SF-DG-DS-SPRT (DSDG Supporting Equipment). Then, this fragility Page 143 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-2: Summary of Finding F&Os and Disposition Status Suggested SR F&O Description Basis Disposition Resolution group was mapped to various DSDG-related items [77].
The fire compartment where the DS Bus is located is FC250 and the fire ignition source ID assigned to the DS Bus is 1448 [81]. The conditional core damage probability associated with this scenario is 4.39E-04, setting severity factor and non-suppression probability all conservatively to one due to seismic damage anticipated in the area [81]. This CCDP is used in the bounding seismic CDF analysis for the purpose of screening the DS Bus duct as a seismic-induced fire ignition source. The resulting total seismic CDF is 5.07E-8, which is less than the threshold value of 5E-07 for screening. Based on this result, the seismically induced bus duct fire associated with DS-BUS is screened out of the RNP SPRA.
This is documented in the station calculation for the Robinson Nuclear Power Plant Seismic Induced Flood and Fire Assessment [69].
Page 144 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-2: Summary of Finding F&Os and Disposition Status Suggested SR F&O Description Basis Disposition Resolution Therefore, this does not represent a change in methodology, scope, or capability as defined in Appendix 1-A of the ASME/ANS PRA Standard and is not considered an upgrade.
The response to this Finding meets the requirements of NTTF 2.1 seismic and Capability Category II of the Standard with NRC Clarification. This finding is considered resolved for the purposes of NTTF 2.1.
The SIFF Notebook statement that The Lube Oil (LO) Storage Tanks the class III turbine building will are located in the southwest portion collapse before any LO tank fire of the Class III portion of the SPRA-A2 requires that can occur is misleading. There will Turbine Building. The finding points a systematic process be be two scenarios or cutsets: out Sections 1.2.1 and 4 of the The scenario with the LO station calculation for the performed to identify
- 1. One in which the LO tank fails tank failure and ignition Robinson Nuclear Power Plant initiators caused by may be insignificant Seismic Induced Flood and Fire secondary hazards. The and TBIII does not fail with a subsequent LO tank fire. compared to TBIII failure, Assessment [69] as an SR C-SPR- assessment used to 25-3 but the scenario would inappropriate basis for screening A2 screen out the Lube Oil
- 2. A separate scenario will occur occur independent of the LO tanks as seismic-induced (LO) storage tank fire TBIII failure and should fire ignition source.
scenario is not properly when the LO tank does not fail and then the TBIII fails. be screened in this characterized. The LO storage tanks have the low context probability of fire damage due to In all likelihood, the probability of the high lube oil flash point (432° F)
LO tank seismic failure combined and the availability of fire detection with a low ignition probability (due and suppression system to limit the to high flash point, potential spread of the fire. The high flash suppression etc.), will have an point would make it difficult for a Page 145 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-2: Summary of Finding F&Os and Disposition Status Suggested SR F&O Description Basis Disposition Resolution insignificant contribution when fire to occur even if a seismic event compared to the TBIII failure. was to rupture the LO storage tanks and a damaging hot gas layer is unlikely to form due to the Turbine Building being large and open to the outdoors. Based on this consideration, the two quoted sentences have been deleted to clear up the inconsistency noted in the Finding in the station calculation for the Robinson Nuclear Power Plant Seismic Induced Flood and Fire Assessment [69].
This is a documentation issue only and this does not represent a change in methodology, scope, or capability as defined in Appendix 1-A of the ASME/ANS PRA Standard and is not considered an upgrade.
The response to this Finding meets the requirements of NTTF 2.1 seismic and Capability Category II of the Standard with NRC Clarification. This finding is considered resolved for the purposes of NTTF 2.1.
Page 146 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-2: Summary of Finding F&Os and Disposition Status Suggested SR F&O Description Basis Disposition Resolution SPRA-A1 requires that a A systematic review of all internal systematic process be performed events and internal flooding to identify initiators caused by the initiating events were reviewed for seismic event. No evidence of a Document a systematic inclusion in the SPRA. Table 5-1 systematic process to identify review of each internal was added to Section 5.3.1 of the seismic imitators was identified. initiating event and station calculation for the H.B.
identify whether or not Robinson Seismic Probabilistic Table A-4 of the station calculation that event should be included as an initiating Risk Assessment Model Notebook for the Seismic Equipment List
[64] lists the internal events event in the seismic [77], which provides a list of the initiators that were screened out as PRA. initiating events and whether to There is no description non-applicable and Table A-22 in include each initiating or not in the in the modeling the station calculation for the During the peer review, SPRA, both initial and final notebook of a Seismic Equipment List [64] the Robinson SPRA assessments. A final disposition is systematic disposition of team described a provided for each initiating event, SR C-SPR-25-4 internal initiating events documents the disposition of all process that would which either states why an initiating A1 that were included in the basic events and initiating events in address this issue. This event is not included or how it is SPRA. the model for the purposes of SEL process involved a development however, they are utilized in the SPRA.
comprehensive review of reviewed only for the purposes of internal and flooding Therefore, this does not represent SEL development and not for initiating events and a identification of initiating events. a change in methodology, scope, or subsequent disposition capability as defined in Appendix 1-Based on Duke's response to a for each with regard to Peer Review question, there are A of the ASME/ANS PRA Standard system impacts or several other initiators modeled in mapping to seismic and is not considered an upgrade.
the internal events PRA that are The response to this Finding meets initiating events.
not listed in Section 5.3.1 of the the requirements of NTTF 2.1 station calculation for the H.B. seismic and Capability Category II Robinson Seismic Probabilistic of the Standard. This finding is Risk Assessment Model Notebook considered resolved for the
[77] purposes of NTTF 2.1.
Page 147 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-2: Summary of Finding F&Os and Disposition Status Suggested SR F&O Description Basis Disposition Resolution A review of all initiators in the internal events model was performed and dispositioned for the SPRA; however, a table of each initiator with its disposition is not formally documented.
Appropriate documentation of the rationale for dispositioning seismic-equivalent of internal events initiators needs to be provided to support reproducibility of the analysis and future maintenance of the SPRA.
The description in Internal Events F&O AS-A5-1 was Section 7.0 of the dispositioned in Section 7 of the station calculation for station calculation for the H.B.
The response to a Peer Review the H.B. Robinson Include the discussion in Robinson Seismic Probabilistic question indicates that impact of Seismic Probabilistic the response to the peer Risk Assessment Model Notebook TB failure on steam lines was Risk Assessment Model review question [77]; however, the discussion did considered in the disposition to the SR C-SPR- Notebook [77] regarding regarding assumed MSIV not include the assessment of the open F&O (AS-A5-1). This B2/C-SPR- 25-6 disposition of open F&O closure given steam line seismic modeling for the steam line discussion is important to F2 AS-A5-1 does not failure due to turbine understanding how this F&O does break. The assessment concluded include discussion of the building seismic failure.
not impact the results of the SPRA. that the MSIVs would remain seismic modeling assumption for steam available to provide function to line break. isolate the main steam line given a Class 3 TB failure, including failure which results in interaction with the Class 1 TB.
Page 148 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-2: Summary of Finding F&Os and Disposition Status Suggested SR F&O Description Basis Disposition Resolution This discussion has been added to the disposition of AS-A5-01 in Section 7 of the station calculation for the H.B. Robinson Seismic Probabilistic Risk Assessment Model Notebook [77].
Therefore, this does not represent a change in methodology, scope, or capability as defined in Appendix 1-A of the ASME/ANS PRA Standard and is not considered an upgrade.
The response to this Finding meets the requirements of NTTF 2.1 seismic and Capability Category II of the Standard with NRC Clarification. This finding is considered resolved for the purposes of NTTF 2.1.
FLEX AFW pumps in Section D.5.6 of the station Recommend either use Nominal AFW pump reliability data the RNP SPRA use the calculation for the H.B. Robinson of available industry data was originally used as a surrogate internal event AFW Seismic Probabilistic Risk for FLEX equipment or to for the FLEX pump reliability data.
pump reliability. Use of Assessment Model Notebook [77] adjust the Robinson plant The FLEX pump data has been SR C-SPR- data for safety related discusses the decision to use data to account for the updated based on Duke fleet-25-8 installed plant nominal AFW pump reliability data expected difference in B8 specific FLEX data.
equipment for diesel for the diesel driven AFW pumps. reliability between Because of the distinctly different Duke PRA collected FLEX pump driven mobile FLEX installed plant equipment equipment may not be characteristics between the mobile and mobile FLEX data from all six of its fleet sites.
appropriate. FLEX AFW pumps and the equipment. This data was used to determine installed, motor driven AFW FLEX pump failure probabilities.
Page 149 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-2: Summary of Finding F&Os and Disposition Status Suggested SR F&O Description Basis Disposition Resolution pumps, use of this data is not Section D.5.6 of the station appropriate. (refer to DA-D2). calculation for the H.B. Robinson Seismic Probabilistic Risk Assessment Model Notebook [77]
has been revised with this discussion and the testing data.
Sections D.3.4 and D.3.6 of the station calculation for the H.B.
Robinson Seismic Probabilistic Risk Assessment Model Notebook
[77] were also revised.
Therefore, this does not represent a change in methodology, scope, or capability as defined in Appendix 1-A of the ASME/ANS PRA Standard and is not considered an upgrade.
The response to this Finding meets the requirements of NTTF 2.1 seismic and Capability Category II of the Standard. This finding is considered resolved for the purposes of NTTF 2.1.
Documentation of the In the seismically This IFSN HLR requires an The concern about the eyewash eyewash station induced fire and flooding assessment of flooding propagation station which is supplied by the SR C-SPR- seismically induced document, provide either 25-9 flood does not address path and identification of potential Potable Water System and its B9 a reference to the downstream impacts. Discussion of potential breach and subsequent the potential for relevant section of the the eyewash station flooding consequences was conservatively propagation into other RNP internal flooding scenario in the seismically induced evaluated. With a LOSP, the areas, features that PRA or include a Page 150 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-2: Summary of Finding F&Os and Disposition Status Suggested SR F&O Description Basis Disposition Resolution could terminate flooding PRA does not address the discussion of the Potable Water System ceases to propagation or operator potential for propagation. eyewash station flooding be a flood source.
actions to terminate modeling considerations flow. A response from the RNP PRA (e.g. assessment of In other words, the loss of offsite modeling consultant indicates that propagation). power prevents or terminates a the RNP internal flooding PRA has seismic-induced flooding event. To assessed propagation for this better understand the risk impact of scenario but this information is not this scenario, a bounding CDF contained in the seismic fire and assessment was performed using flooding documentation. the fragilities for the eyewash station and the LOSP.
Using this seismic-induced flooding fragility, a bounding risk assessment was performed to gauge the impact of this flooding event on the overall seismic CDF.
The results demonstrate the insignificance of the seismic-induced flooding event initiated by the eyewash station failure and provides the justification that additional flooding scenario for flood propagation does not need to be retained for the RNP SPRA model.
Further details on this F&O can be found in the station calculation for Page 151 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-2: Summary of Finding F&Os and Disposition Status Suggested SR F&O Description Basis Disposition Resolution the RNP PRA Model Peer Review Resolution [71]
This does not represent a change in methodology, scope, or capability as defined in Appendix 1-A of the ASME/ANS PRA Standard and is not considered an upgrade. The response to this finding meets the requirements of NTTF 2.1 seismic and Capability Category II of the Standard. This finding is considered resolved for the purposes of NTTF 2.1.
The station calculation for the Robinson Fire PRA identified a total Robinson Nuclear Power Plant Seismic Induced Flood and Fire of 2120 fire ignition sources with Disposition the screening Assessment [69] references the of all ignition sources different fire scenarios [81]. These EPRI Report 3002012980 listed as Retain in EPRI sources are processed using a Some components that seismically induced fire screening 3002012980 [93] Table combination of qualitative and should be retained for methodology [93]. Table 2-2 of this 2-2 or include them in the quantitative screening filters.
further assessment EPRI report lists ignition sources to SPRA. Details are provided in the station SR C-SPR- based on EPRI Report 25-10 be considered for the SPRA. Air calculation for the Robinson A2 3002012980 [93] are not compressors, pumps, diesel Consider the need to Nuclear Power Plant Seismic retained.
generators, and bus ducts are conduct walkdowns of Induced Flood and Fire listed as 'Retain' in Table 2-2 but equipment that should be Assessment [69].
are not dispositioned in the station retained per SFR-D6.
calculation for seismic induced Based on the results of these flood and fire assessment [69]. screening steps, there are no potential seismic-induced fire Page 152 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-2: Summary of Finding F&Os and Disposition Status Suggested SR F&O Description Basis Disposition Resolution ignition sources that need to be retained in the Robinson SPRA.
This does not represent a change in methodology, scope, or capability as defined in Appendix 1-A of the ASME/ANS PRA Standard and is not considered an upgrade.
The response to this finding meets the requirements of NTTF 2.1 seismic and Capability Category II of the Standard with NRC Clarification. This finding is considered resolved for the purposes of NTTF 2.1.
The failure mode listed in the A bounding SCDF was computed station calculation for Seismic using the failure of the Robinson Fragility Evaluation Notebook [82] Dam to represent the Service for heat exchangers is anchorage Assess whether an Water and inducing a flood when Potential flooding failure. In the RNP SPRA model, anchorage failure for the CCW Hx anchorage fails and consequences of heat the seismic failure is mapped to a heat exchangers would taking the consequence directly to exchanger anchorage plugging failure of the heat also result in a core damage.
SR C-SPR-25-11 failure not assessed. exchanger and seismically induced seismically induced B9 The results show that overall SCDF flooding impacts are not assumed. flood.
These flooding impacts may be remains essentially the same limited to loss of the system relative the baseline SCDF. This function (e.g. CCW) or could demonstrates the insignificance of involve flooding induced failure of the seismic-induced flooding event nearby components. initiated by the CCW Hx anchorage failure and provides the justification Page 153 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-2: Summary of Finding F&Os and Disposition Status Suggested SR F&O Description Basis Disposition Resolution that the postulated flooding scenario does not need to be retained for the RNP SPRA model.
Further details on this F&O can be found in the station calculation for the RNP PRA Model Peer Review Resolution [71].
This does not represent a change in methodology, scope, or capability as defined in Appendix 1-A of the ASME/ANS PRA Standard and is not considered an upgrade.
The response to this finding meets the requirements of NTTF 2.1 seismic and Capability Category II of the Standard. This finding is considered resolved for the purposes of NTTF 2.1.
The SF-TK-DG-FO-STRG-TNK As discussed in the Basis for fragility group is associated with 8 Significance, failure of this group is different components including associated with multiple No documented basis entries for the tank itself along with Document the rationale components due to interactions.
for the SF-TK-DG-FO- others for the DFO XFER pumps, behind grouping in the Discussion on interaction items SR C-SPR- STRG-TNK fragility piping etc. In this case, per SF-TK-DG-FO-STRG-25-12 using the DFOST fragility as an B4 group. response from the Robinson SPRA TNK fragility group.
example has been added to the team, failure of the tank is assumed to impact the other components in bullet list in Section 5.3.5 of the the group. This rationale for station calculation for the H.B.
grouping is different than what is Robinson Seismic Probabilistic described in section 5.3.5 of the Risk Assessment Model Notebook Page 154 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-2: Summary of Finding F&Os and Disposition Status Suggested SR F&O Description Basis Disposition Resolution station calculation for the H.B. [77].
Robinson Seismic Probabilistic Risk Assessment Model Notebook This F&O is related to F&O 24-6.
[77].
Therefore, this does not represent a change in methodology, scope, or capability as defined in Appendix 1-A of the ASME/ANS PRA Standard and is not considered an upgrade.
The response to this Finding meets the requirements of NTTF 2.1 seismic and Capability Category II of the Standard. This finding is considered resolved for the purposes of NTTF 2.1.
The site response Section 6.2.3 of the station Revise Section 6.2.3 of Basis for Establishing Three analysis is based on calculation for the Geotechnical the station calculation for Shear Wave Velocity Profiles three site shear wave Analysis Report [83] and Section the Geotechnical velocity profiles which 5.1.1 of the station calculation for Analysis Report [83] and As discussed in Section 5.2.1 of the are discussed in Section Seismic Plant Hazard Analysis [16] Section 5.1.1 of the station calculation for Seismic Plant 6.2.3 of the station The provide the information and station calculation for Hazard Analysis [16]
calculation for the basis for establishing the three site Seismic Plant Hazard a site stratigraphic framework was SR C-SHA- identified that included an upper 26-1 Geotechnical Analysis profiles used to perform the site Analysis [16]
E2 Report [83] and Section response analysis. New shear to improve the layer with sand and clay/silt layers 5.1.1 of the station wave velocity data was gathered at explanation and basis for of variable consistency, a clay layer calculation for Seismic the site to aid in defining the site establishing the three with typically hard consistency, and Plant Hazard Analysis profile. base case shear wave a lower layer of interbedded sand
[16], the two reports do velocity profiles and the and clay layers of variable not document a The station calculation for the weights assigned to each thickness and variable sufficient basis for Geotechnical Analysis Report [83] one. consistencies. Rock was identified establishing the three notes that it is normal practice to Page 155 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-2: Summary of Finding F&Os and Disposition Status Suggested SR F&O Description Basis Disposition Resolution base case profiles and develop three site profiles with a by a deep boring as being the weights assigned to reference to EPRI (2012, the SPID) approximately 400 feet below the each one. [2], however the station calculation ground surface.
for the Geotechnical Analysis Report [83 ] does not sufficiently Nine Spectral Analysis of Surface explain the role of the new data Wave (SASW) lines were and why it leads to the three performed at locations on all sides profiles selected. For example, is of the Plant Area, and Suspension the new data reflective of profile PS seismic velocity logging to epistemic uncertainty or lateral measure the shear wave velocity variability, particularly for profile A (Vs) was performed in Borings B-1A and B? The station calculation for (upper 120 feet) and B-1B (full Seismic Plant Hazard Analysis [16]
depth, to rock). All the Vs data indicates that the relative weights assigned to each profile are showed a distinct increase in Vs in primarily based on the available the hard clay layer, and the Vs in data; however the PSHA does not material above the hard clay was sufficiently explain how the position generally similar at all points. The of the data was considered in significant difference was the establishing the profile weights. indicated presence of a low Vs layer under the hard clay (velocity inversion) in 4 of the data sets, but not in others. Thus, a single site profile with variation represented by upper and lower bound values to represent uncertainty was not reasonable. Three separate Vs profiles were developed, one to represent areas without the velocity inversion (Profile C) and two to Page 156 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-2: Summary of Finding F&Os and Disposition Status Suggested SR F&O Description Basis Disposition Resolution represent areas with the velocity inversion (Profiles A and B).
The separate Profiles A and B were based on the observation of differences in the velocities observed between the P-S suspension log data from Boring B-1B (basis for Profile A) compared to those obtained for SASW Arrays 6, 7, and 8 (basis for Profile B). Figure GAR-28A compares the shear wave velocity profiles interpreted for the three SASW arrays used to develop B to the Boring B-1B P-S suspension log shear wave velocity profile of Profile A. All four profiles show the same general characteristics of the hard clay layer underlain by a layer of lower velocity. However, the variability in velocity among the SASW profiles is small and none of the three captures the strength of the velocity inversion seen in the P-S suspension log data.
The assessment was made that the P-S suspension log provides a more detailed picture of the variation of velocity with depth than Page 157 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-2: Summary of Finding F&Os and Disposition Status Suggested SR F&O Description Basis Disposition Resolution does the deeper portion of the SASW lines. This is indicated by the decrease in detail in the interpreted SASW profiles with increasing depth. Discussions with technical reviewers also indicated a preference for the use of P-S suspension log data compared to SASW to define a detailed velocity profile. On this basis it was judged appropriate to capture epistemic uncertainty in the strength of the velocity inversion beneath the hard clay with two profiles, one based on the detailed P-S suspension log data that show a strong inversion and one based on the SASW data that show a weaker inversion.
Basis for Selecting Weights for the Three Site Velocity Profiles Section 5.1.1 of the station calculation for the Final Seismic Analysis Report [16]
notes that the three velocity profiles were developed from geophysical investigations conducted around the periphery of the Category 1 structures (plant protected area).
Information on Vs within the plant Page 158 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-2: Summary of Finding F&Os and Disposition Status Suggested SR F&O Description Basis Disposition Resolution protected area was limited to downhole velocity testing in three borings in the area of the ISFSI.
Those borings did not extend through the hard clay layer.
Previous borings in 1965 by Dames and Moore generally stopped within the hard clay layer and did not include testing for Vs. Because of the absence of definitive Vs data in the plant protected area, the presence or absence of a velocity inversion in the plant protected area could not be determined.
Equal weight was assigned to the condition of a velocity inversion existing or not existing.
Review of potentially useful data within the plant protected area has identified four pieces of information relevant to the velocity inversion presence or absence that had not been previously evaluated. These are the records from drilling and installation of four deep wells for plant water supply. Three deep wells were installed within the protected area in 1968 (Deep Wells Page 159 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-2: Summary of Finding F&Os and Disposition Status Suggested SR F&O Description Basis Disposition Resolution A, B and C). A fourth well (Deep Well D), was installed in 2004.
Records for the deep well installation from plant files provided by Duke contain, among other information, the Formation Logs of the wells. The Formation Log lists the well drillers description of different soil layers penetrated, and, in most cases, notes about drilling difficulty if greater or less than normal drilling. Normal well drilling practice is to observe wash water coloration and to examine sediments captured by inserting a strainer into the wash water return.
In addition to the Formation Logs, natural gamma and resistivity logs and rates of drilling penetration were available for Deep Well D.
The stratigraphic and drilling information on the Formation Logs for Deep Wells A, B, C and D has been transcribed to the boring log format used in reporting results of the geotechnical borings. The transcribed formation logs, along with the upper 250+/- of the Boring B-1B log are summarized on the Page 160 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-2: Summary of Finding F&Os and Disposition Status Suggested SR F&O Description Basis Disposition Resolution station calculation for the Geotechnical Analysis Report [83].
Boring B-1B formed the basis for developing Profile A and identified the velocity inversion below Layer 2 of the Profile.
Boring B-1B has a natural gamma log as does Deep Well D. Natural gamma logs provide information on the variations in soil type and are useful in stratification. In the station calculation for the Geotechnical Analysis Report [83],
those two gamma logs were compared and a close similarity in the indications of clay and sand was noted. The purpose of that comparison was to include the Deep Well D information on depth to rock in the overall analysis. A further comparison is made showing the gamma log plotted on a portion of the stratigraphic boring log for boring B-1B.
The gamma logs in Boring B-1B and Deep Well D are further compared for their behavior in the zone where Profile A (based on Boring B-1B) indicated the Page 161 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-2: Summary of Finding F&Os and Disposition Status Suggested SR F&O Description Basis Disposition Resolution presence of the velocity inversion.
The Vs log shows similar increases and decreases to the gamma logs.
The stiff clay layer (Layer 2) is consistent with the depth range of higher gamma values shown on the gamma logs. The velocity inversion (Layer 3A) also begins in reasonable proximity to the start of the lower gamma readings, indicating a change in soil from a more clayey to a more sandy composition. The Formation Log for Deep Well D notes that below this change in soil type, loose sandy soils, rapid drilling advance and losses of circulation were encountered for approximately 100 feet. Those notes indicate presence of weak soils, consistent with a low Vs. From these observations, the conditions in Deep Well D are consistent with those in Boring B-1B and the possibility of the velocity inversion being present at Deep Well D is interpreted as likely.
Similarly, the Formation Logs for Deep Wells A, B and C consistently indicate the presence Page 162 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-2: Summary of Finding F&Os and Disposition Status Suggested SR F&O Description Basis Disposition Resolution of the stiff/hard clay zone described as Layer 2 in the velocity profiles.
There is also notation in the logs for Deep Wells A and C that soils below the hard clay exhibit soft drilling over the remainder of the well bore holes; however, the log for Deep Well B does not clearly indicate such conditions. Based on the match between Deep Well D and Boring B-1B information, the formation logs for Deep Wells A, C and D show indications that the velocity inversion could be present in these wells.
Because three of the four deep wells, which are on all sides of the reactor containment building, show potential for a velocity inversion, the use of an equal weighting for Profiles A and B which include a velocity inversion and Profile C which does not include a velocity inversion is reasonable.
Section 5.2.1 of the station calculation for the Geotechnical Analysis Report [83] is edited to add information on the Deep Wells A, B and C. Section 6.2.3 of the Page 163 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-2: Summary of Finding F&Os and Disposition Status Suggested SR F&O Description Basis Disposition Resolution Geotechnical Analysis Report [83]
is edited to provide more information on the potential for having (or not having) a velocity inversion within the area of the Category 1 structures. Section 5.1.1 of the Final Seismic Analysis Report [16] is edited to further explain the selection of equal weighting for Profile C relative to Profiles A and B.
Therefore, this does not represent a change in methodology, scope, or capability as defined in Appendix 1-A of the ASME/ANS PRA Standard and is not considered an upgrade.
The response to this Finding meets the requirements of NTTF 2.1 seismic and Capability Category II of the Standard. This finding is considered resolved for the purposes of NTTF 2.1.
The approach used to The SRA is based on using seed Revise Sections 4.2.3, The response to F&O 26-3 is based perform the SRA is time histories which are loosely 5.2 and 5.3 of the Final on work developed in the station different than that matched to the target CMS for Seismic Analysis Report calculation for the Final Seismic SR C-SHA-26-3 described in the EPRI either high frequency or low [16] to enhance the Analysis Report [16]. The J1 Guidance Document frequency. Insufficient detail is justification for calculation describes two sensitivity (SPID) [2] Sections provided to understand what is performing the SRA analyses that demonstrate that the 4.2.3, 5.2 and 5.3 of the meant by loosely matched and using single time approach of using a single time Page 164 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-2: Summary of Finding F&Os and Disposition Status Suggested SR F&O Description Basis Disposition Resolution Seismic Analysis Report more importantly whether this histories per randomized history selected from the set of 30 (PSHA) [16] describe impacts the resulting SRA soil profile, and that time histories for each site how time histories are amplification factors at any given loosely conditioning the response analysis with one of the selected and input loading level. The text states time history to the CMS 60 randomized dynamic properties conditioned to the target that this was done to account for is appropriate. The profiles produces both consistent conditional mean the natural variability in the response to the peer mean results and consistent spectra (CMS). While 30 frequency content of ground review questions related variability results.
time histories are motions; however since each SRA to the adequacy of the Therefore, this does not represent selected for use, these run (one randomized profile, one time history approach for a change in methodology, scope, or are used one-by-one randomized G/Gmax and damping) SRA adequately capability as defined in Appendix 1-with each of the uses only one loosely matched addressed this issue; A of the ASME/ANS PRA Standard randomized soil profiles time history, the approach taken appropriate text from and is not considered an upgrade.
to perform the SRA. would appear to introduce ground these responses forms The response to this Finding meets Insufficient detail is motion amplitude (or strain) the input for the revision the requirements of NTTF 2.1 provided to understand variability into the SRA process to Section 5.3.1of the seismic and Capability Category II whether the conditioning beyond what is intended by the station calculation for of the Standard. This finding is process or the use of EPRI guidance (SPID) [2]. Seismic Plant Hazard considered resolved for the one time history per Analysis [16]
purposes of NTTF 2.1.
randomized soil profile impacts the assessment of site response amplification factors.
In Section 5.3.1 of the Figures 5-52 to 5-57 of the station Revise Section 5.3.1 of Station calculation for the Final station calculation for calculation for Seismic Plant the station calculation for Seismic Analysis Report [16] was Seismic Plant Hazard Hazard Analysis [16] show the Seismic Plant Hazard prepared to review recent work SR C-SHA- Analysis [16]: shear strain computed in the SRA. Analysis [16] regarding when results of 26-4 J1 There are shear strain values to justify the level of equivalent linear (EL) 1-D site response may be biased compared The computed effective exceeding 1% especially for shear strain that is to non-linear (NL) analyses. The shear strains in the soil profiles A and B and at higher considered acceptable loading levels. Using the equivalent when using the EL results of the calculation show that layers were examined Page 165 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-2: Summary of Finding F&Os and Disposition Status Suggested SR F&O Description Basis Disposition Resolution and found to be linear (EL) approach out of the approach. The response the EL site response analyses for generally less than 1 range applicability at any point in to the peer review the RNP provide an acceptable percent at the higher the soil profile can impact the question related to shear representation of site amplification loading levels. reliability of the site amplifications strain limits adequately for the purposes of risk Therefore, the use of and therefore the seismic hazard addressed this issue; quantification at the RNP.
the equivalent linear calculated for the control point. appropriate text from this Information from the calculation is approach is considered response forms the input incorporated into the station appropriate. It should be for the revision to Section calculation for the Final Seismic noted that at the highest 5.3.1. Analysis Report [16].
loading levels, the imposition of a minimum Therefore, this does not represent site amplification of 0.5 a change in methodology, scope, or means that the capability as defined in Appendix 1-computed site A of the ASME/ANS PRA Standard amplification values at and is not considered an upgrade.
high strains are often The response to this Finding meets not used in developing the requirements of NTTF 2.1 the soil hazard. seismic and Capability Category II of the Standard. This finding is The level of shear strain considered resolved for the that is considered purposes of NTTF 2.1.
acceptable when using equivalent linear approach should be justified. Using the results beyond the acceptable strain level to develop the median and standard deviation site amplification needs to be justified as well.
Page 166 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-2: Summary of Finding F&Os and Disposition Status Suggested SR F&O Description Basis Disposition Resolution Figure 5-29 of the station calculation for Seismic Plant Hazard Analysis [16] Revise Section 5.5.1 of shows weighted mean the station calculation for amplification functions Seismic Plant Hazard for the full column site Analysis [16]
profile for elevation 226 to provide figures Mean amplification functions for the ft at different hazard showing the mean site six site response analysis cases levels. These functions amplification factor for were computed and plotted in the represent the To better understand and examine each of the six station calculation for the Final combinations of mean the effect of different sets of alternative Seismic Analysis Report [16].
amplifications for the six dynamic properties used in the site characterizations of the Therefore, this does not represent sets of dynamic response analysis on the seismic RNP site profile (3 base a change in methodology, scope, or SR C-SHA- hazard, site amplification functions case velocity profiles and capability as defined in Appendix 1-26-5 properties. It will be 2 alternative dynamic J2 useful to provide figures for each of the alternative showing the mean and characterizations could be material properties) for A of the ASME/ANS PRA Standard standard deviation site illustrated. selected ground motion and is not considered an upgrade.
amplification factor for levels. The response to The response to this Finding meets each of the six the peer review question the requirements of NTTF 2.1 related to site profiles seismic and Capability Category II alternative of the Standard. This finding is characterizations of the adequately addressed this issue; appropriate considered resolved for the RNP site profile (3 base case velocity profiles text from this response purposes of NTTF 2.1.
and 2 alternative forms the input for the dynamic material revision to Section 5.5.1.
properties) for selected ground motion levels.
Page 167 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-2: Summary of Finding F&Os and Disposition Status Suggested SR F&O Description Basis Disposition Resolution Section 5.8.2 of the The text of the third paragraph of station calculation for Section 5.8.2 of the station Seismic Plant Hazard calculation for Seismic Plant Analysis [16] Hazard Analysis [16] is modified as indicates that the follows:
vertical to horizontal The soil hazard curves are then (V/H) response spectral used to compute a horizontal The base of reactor piles located in ratios described in SCOR FIRS for the base of the the hard clay layer which is stiffer Section 5.6 and used for than the soil condition at elevations Revise Section 5.8.2 of reactor building piles, which is GMRS and FIRS at elevations 226 ft and 226 ft and 216 ft. Therefore, the the station calculation for smoothed and extended to cover V/H ratios developed and use for Seismic Plant Hazard the frequency range of 0.1 to 100 216 ft, respectively were Hz using the approach described soil condition may not be applicable Analysis [16]
also used to develop a vertical FIRS for the to the stiffer clay. The horizontal to indicate that different above in Section 5.5. Because the and vertical FIRS provided in table V/H ratios were applied piles are founded in the hard clay, SR C-SHA- base of reactor piles. which is stiffer than the average 26-6 However, review of 5-38 of the Seismic Analysis for the vertical SCOR J2 tables 5-19 and 5-20 Report (PSHA) [16] indicate that FIRS at the base of the soil, the vertical to horizontal (V/H) different V/H ratios were used for piles. The response to response spectral ratios described which provide the the base of reactor piles. The use the peer review question in Section 5.6 may be too low at GMRS and FIRS at elevations 226 ft and of V/H ratios other than those provides appropriate text low frequency. Therefore, the V/H developed in Section 5.6 should be for revising Section 5.8.2. ratios for the base of the piles are 216 ft, respectively and calculated using both the envelope discussed in Section 5.8.2 of the table 5-38 which spectral ratios from Figure 5-34 and station calculation for Seismic Plant provides the FIRS at the the ratios developed by McGuire et Hazard Analysis [16]
base of reactor piles al. (2001) for CEUS hard rock suggests that the V/H conditions of PGA in the 0.2 to 0.5 ratios developed in g range interpolated to the 3 Section 5.6 were not frequency values at which used for the base of horizontal FIRS are computed.
reactor piles. These values and the envelope from Figure 5-34 are then enveloped and used to develop a vertical FIRS for the elevation Page 168 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-2: Summary of Finding F&Os and Disposition Status Suggested SR F&O Description Basis Disposition Resolution 159.2 ft. control point. The SCOR FIRS are tabulated in Table 5-38 and illustrated on Figure 5-44.
Therefore, this does not represent a change in methodology, scope, or capability as defined in Appendix 1-A of the ASME/ANS PRA Standard and is not considered an upgrade.
The response to this Finding meets the requirements of NTTF 2.1 seismic and Capability Category II of the Standard. This finding is considered resolved for the purposes of NTTF 2.1.
The damping value (2% A damping value of 2% critical is critical) used in deriving used in deriving the Rayleigh The Fragility vendor performed a Perform a sensitivity the Rayleigh damping damping coefficients for use in the sensitivity analysis using the analysis to determine if coefficients for use in nonlinear analysis of the station proposed 4% damping to see if it use of a more realistic the nonlinear analysis of calculation for the Seismic would significantly lower the plant damping (e.g., 4%)
the Class III TB in the Fragility of Pounding between the risk (CDF and LERF). The changes would significantly lower station calculation for Class III Turbine Bldg. and the to the fragility were less than 5% as the plant risk (CDF and SR C-SFR- the Seismic Fragility of Reactor Aux. Bldg. by Nonlinear a median capacity and were 28-1 LERF). Alternatively, B3 Pounding between the Analysis [68] While this value may relatively minor. The impact on justify the use of 2%
Class III Turbine Bldg. be appropriate for light steel braced CDF and LERF would be non damping as being and the Reactor Aux. structures, it may not be realistic for consequential.
realistic for this response Bldg. by Nonlinear a structure such as the Class III analysis. Further details on this F&O can be Analysis [68] may be too TB. The peer review team is aware conservative. that damping values corresponding found in the station calculation for to Response Level 1 as defined in the RNP PRA Model Peer Review ASCE 4 [59] are to be used in Resolution [71].
performing nonlinear dynamic Page 169 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-2: Summary of Finding F&Os and Disposition Status Suggested SR F&O Description Basis Disposition Resolution analysis. However, for the case of Therefore, this does not represent the TB Class III, additional factors a change in methodology, scope, or need to be considered in selecting capability as defined in Appendix 1-an appropriate damping value for A of the ASME/ANS PRA Standard the structure. Though the steel and is not considered an upgrade.
framing was designed as friction- The response to this Finding meets type connections, it is not clear the requirements of NTTF 2.1 they would remain as such during seismic and Capability Category II the GMRS. The forces in the steel of the Standard. This finding is members need to be evaluated to considered resolved for the determine the connections would purposes of NTTF 2.1.
slip resulting in a bearing-type connection. For the latter, 4%
damping would be appropriate for Response Level 1. In addition, consideration should be given to the deformations of the non-structural elements and cracking of the mezzanine slabs caused by the pounding between TB Class III and RAB.
Page 170 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-2: Summary of Finding F&Os and Disposition Status Suggested SR F&O Description Basis Disposition Resolution Following the original walkdowns of the items on the seismic equipment list (SEL), Duke conducted SSLOCA has been assigned a supplemental walkdowns of piping representative fragility of 0.1g and tubing whose failure could lead HCLPF corresponding to the SSE to Small Small Loss of Coolant level. Given that SSLOCA appears Accident (SSLOCA) in order to as one of the top contributors to generate a more plant-specific SCDF as reported in the station fragility for SSLOCA. These calculation for the Robinson walkdowns are summarized in the Seismic Probabilistic Risk station calculation for the Seismic A refined fragility Assessment Quantification Capacity Walkdown Report [53].
evaluation was not Update the fragility Notebook [13] (FV of 5.05%) and evaluation for SSLOCA performed for SSLOCA SLERF (FV of 9.1%), a refined All of the piping/tubing was judged though this item is one fragility would be required for this incorporating the results to have High seismic capacity. This SR C-SFR- of the on-going risk capacity ranking translates into the 28-2 of the top risk component.
E3 reduction program at same fragility information contributors to CDF and In response to a question, RNP LERF. RNP. recommended for the RNP safety-stated that a separate risk related piping determined to have reduction project has been initiated high seismic capacities by the EPRI for the site, and as part of that, SPRAIG [11]and considered to be walkdowns were performed for appropriate as a more refined SSLOCA during the October 2018 fragility for the SSLOCA-related refueling outage. That work should piping and tubing in the station result in an improved HCLPF for calculation for the Robinson SSLOCA, and along with potential Representative Fragilities Overview modifications.
[61].
This does not represent a change in methodology, scope, or capability as defined in Appendix 1-A of the ASME/ANS PRA Standard and is not considered an upgrade.
Page 171 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-2: Summary of Finding F&Os and Disposition Status Suggested SR F&O Description Basis Disposition Resolution The response to this finding meets the requirements of NTTF 2.1 seismic and Capability Category II of the Standard. This finding is considered resolved for the purposes of NTTF 2.1.
The fragility of the Class Class III TB and RAB pounding Though the Class III TB Fragility Perform a more realistic III TB was evaluated for interactions were assessed to have (Pounding effects) appears as the evaluation of the TB III the pounding between a relatively low seismic fragility for top CDF/LERF contributor (FV > and RAB pounding TB and RAB in the the TB Class III. The effects of this 40%) per the station calculation for effects. For example, the station calculation for pounding affected several SSCs the Robinson Seismic impact forces on TB III the Seismic Fragility of and resulted in being a significant Probabilistic Risk Assessment resulting from the RAB-Pounding between the risk contributor for the Robinson Quantification Notebook [13], no TB III pounding could be Class III Turbine Bldg. SPRA. The peer reviewers noted sensitivity analysis could be found determined from a and the Reactor Aux. that there was a potential this SR C-SFR- in the notebook. The fragility simplified response 28-3 Bldg. by Nonlinear evaluation of the Class III TB needs analysis model of the fragility might be conservatively E3 Analysis [68] Per biased and further noted that to be based on realistic evaluation RAB and TB III and Section 6 refinements may be possible based of the pounding effects of the incorporation of impact (Methodology), the on their past experiences with other Reactor Auxiliary Building and TB elements and energy pounding effects were SPRAs. However, the specifics of Class III. It is seen, however, that dissipators. The impact determined based on the impact mechanism for the the impact between the TB forces could then be imposing the relative Robinson Class III TB and RAB are Pedestal and the TB Structure was applied to a static model displacement between different from other more typical modeled using impact elements of the TB III to evaluate the RAB and Class III building impacts observed in other and energy dissipators. the fragility.
TB. Such an approach SPRAs and Duke has solid could potentially technical reasons for concluding Page 172 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-2: Summary of Finding F&Os and Disposition Status Suggested SR F&O Description Basis Disposition Resolution overestimate the impact that more detailed analyses along forces (and thus the lines suggested by the peer underestimate the review team would result in minimal capacity of the changes to the fragility and the structure) as it does not resulting risk.
take credit for the energy dissipation of the The energy dissipation at the impacting surfaces. The impact point does not offer industry practice for significant protection to the evaluation of pounding vulnerable diaphragm. While between adjacent model refinement can increase the structures during a precision of the fragility, the seismic event is based resulting slightly modified fragility on modeling the impact will not change the risk surfaces with conclusions.
appropriate stiffness (Hertz Contact Law) and Further details on this F&O can be non-linear damping found in the station calculation for properties. the RNP PRA Model Peer Review Resolution [71]
Therefore, this does not represent a change in methodology, scope, or capability as defined in Appendix 1-A of the ASME/ANS PRA Standard and is not considered an upgrade.
The response to this Finding meets the requirements of NTTF 2.1 seismic and Capability Category II of the Standard. This finding is considered resolved for the purposes of NTTF 2.1.
Page 173 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-2: Summary of Finding F&Os and Disposition Status Suggested SR F&O Description Basis Disposition Resolution The following items require correction to the documents as 1. Descriptions of the other two indicated: dam failure modes have been added to Section 7 of the station
- 1. Section 7.0 of the station calculation for Seismic Fragility calculation for Seismic Fragility Evaluation Notebook [82]
Evaluation Notebook [82] should be Emphasis on the controlling expanded to describe all three failure mode being liquefaction-failure modes, since they are induced failure has been added included in the logic model, and to .
highlight that the failure mode for 2. Section 5.1 of the station liquefaction-induced settlement calculation for the Seismic controls over the other failure Fragility Evaluation Notebook Several areas were modes. [82] has been updated to The various items listed indicate that the Turbine identified throughout the in the 'Basis' column Building Class III vertical ISRS fragility analysis 2. Section 5.1 (5th bullet) of the SR C-SFR- require corrections as have peaks at frequencies up to 28-4 documentation which station calculation for Seismic F2 noted. 20 Hz.
required corrections. Fragility Evaluation Notebook [82]
should be corrected to reflect that the vertical ISRS have peaks up to 3. Section 8.1.1.2 and section 11, 20 Hz. references 2 to 4 of the station calculation for the Robinson
- 3. Revise the station calculation for Nuclear Power Plant Relay the Robinson Nuclear Power Contact Chatter Analysis [72]
Plant Relay Contact Chatter has been updated to document Analysis [72] to provide additional the vibration isolators and justification for why the vibration updated SQURTs reports.
isolators at the EDG control panels are ineffective. Also, revise the document to update references to 4. Section 4.9.1.1 of the station SQURTS reports rather than just calculation for the Robinson the test data summary sheets. Representative Fragilities Overview [61] has been updated Page 174 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-2: Summary of Finding F&Os and Disposition Status Suggested SR F&O Description Basis Disposition Resolution
- 4. Revise the station calculation for to document the basis for the Robinson Representative variabilities included in the relay Fragilities Overview [61] to reflect representative fragilities.
justification for the use of 0.4 composite variability for relays
- 5. SEWS for Fans HVE-17 and
- 5. Update the station calculation for HVE-18 and Tanks DG-A-EXP-the Seismic Capacity Walkdown TK and DG-B-EXP-TK have Report [53] for EDG room exhaust been updated to document the fans and coolant expansion tanks basis for adequate anchorage as discussed in response to the and to re-rank the components peer review questions. as High in the station calculation for the Seismic Capacity
- 6. Update walkdown report for SI- Walkdown Report [53].
870B to supplement walkdown judgment that interaction between the valve and the handrail of the 6. A supplemental review of BIT platform is not a credible interaction room was performed and the concern -Also, confirm that the SEWs forms were updated to platform is not a SI concern for any enhance the documentation for other equipment in the BIT room. the observations noted in the F&O. The station calculation
- 7. Walkdown SEWS [53] for the for the Seismic Capacity MCR ceiling and the fragility Walkdown Report [53] was analysis within the station updated accordingly.
calculation for Representative Fragilities should be updated to confirm the open item with respect 7. The SEWs for the MCR ceiling to anchorage of light fixtures. in the station calculation for the Seismic Capacity Walkdown
- 8. Track completion of WO Report [53] was updated 13316743 for the CST in order to accordingly as was the clear the unverified assumption in corresponding fragility Page 175 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-2: Summary of Finding F&Os and Disposition Status Suggested SR F&O Description Basis Disposition Resolution the station calculation for the contained in the station Seismic Fragility Evaluation of the calculation for the Robinson CST [85]. Representative Fragilities Overview [61].
- 9. The basis for screening out the displacement-based failure mode 8. Work Order 13316743 for the between the RAB and RCB should CST was completed and closed be documented. on 4 June 2019. The station calculation for the Seismic
- 10. The basis for screening out FLEX haul path should be Fragility Evaluation of the CST documented. [85] has been revised accordingly to reflect the
biased and additional documentation should be provided
- 9. The station calculation for the to justify its impact.
Response Analysis Notebook
[55] has been updated to justify
- 12. The GMRS is used as the input the screening out of failure motion for SSCs housed within the modes associated with the intake structure and for the vertical impact between the RCB and direction, FIRS is used as the RAB.
vertical input motion. Since this embedded structure is expected to move with the soil, additional
- 10. Section 5.7 of the station clarification should be provided on why the soil amplification in the calculation of the vertical direction would not affect Representative Fragilities [61]
the vertical ISRS. has been added to document screening of the FLEX haul
- 13. Preliminary studies were paths.
conducted to determine the effect of pile foundation modeling on the Page 176 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-2: Summary of Finding F&Os and Disposition Status Suggested SR F&O Description Basis Disposition Resolution response of the RCB and RAB, but 11. Section 5.1 of the station it is not documented. Additional calculation for the Turbine basis should be provided to better Building Class III seismic characterize the final SSI modeling response analysis [75] has approach adopted.
been updated to discuss the impact of disregarding horizontal pile flexibility. As indicated, inclusion of pile flexibility could affect gantry crane response at certain frequency ranges but not at others. Disregarding pile horizontal flexibility is considered to be sufficiently accurate.
- 12. Section 3 of the station calculation for seismic response analysis [87] has been updated to provide justification for using the soil column outcrop vertical Foundation Input Response Spectrum as vertical input to SSCs in the Intake Structure. As noted, vertical FIRS are available at Elevation 159 ft.
Vertical response of the Intake Structure occurs at very high frequencies. At these frequencies, the vertical FIRS at Page 177 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-2: Summary of Finding F&Os and Disposition Status Suggested SR F&O Description Basis Disposition Resolution Elevation 159 ft exceeds the vertical GMRS at Elevation 226 ft. Significant amplification of motion between Elevation 159 ft and the Intake Structure Foundation at Elevation 172 ft is considered to be very unlikely.
Use of the vertical FIRS is considered to be appropriate.
- 13. 5.7 Section 4.1.6 of the station calculation for the Response Analysis Notebook [55] is added to identify conclusions obtained by sensitivity studies investigating modeling of the pile-founded structures.
Therefore, this does not represent a change in methodology, scope, or capability as defined in Appendix 1-A of the ASME/ANS PRA Standard and is not considered an upgrade.
The response to this Finding meets the requirements of NTTF 2.1 seismic and Capability Category II of the Standard. This finding is considered resolved for the purposes of NTTF 2.1.
Page 178 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-2: Summary of Finding F&Os and Disposition Status Suggested SR F&O Description Basis Disposition Resolution The model does not assume The 2nd bullet of Section 5.3.5 of correlation between similar the station calculation for the H.B. equipment with similar orientation Robinson Seismic Probabilistic on different elevations of a Risk Assessment Model Notebook structure. As discussed in the
[77] states: finding, this may be non-conservative as the ISRS are Located on the same elevation in Modify the correlation similar at different elevations within the building: In-structure response criteria specified in a given structure in the horizontal spectrum at the given elevation in Section 5.3.5 of the direction due to the pile-soil-The assumption that the building is expected to be station calculation for the structure failure. However, equipment in different essentially the same for all H.B. Robinson Seismic assuming complete correlation elevations of the same equipment located at that Probabilistic Risk between equipment on separate building as having elevation. Assessment Model elevations would be overly different seismic Notebook [77] to reflect conservative.
SR C-SFR- demand, used in Per this criterion, similar equipment the seismic response of 28-5 with similar orientation at different The impact this assumption has on A2 determining fragility RNP structures.
group correlation, may elevations of the same building Document the results of the model was analyzed by not be appropriate for would be assumed to be not the assessment made for reviewing the top contributors. All RNP. correlated. Given that the RNP the top contributors SSCs in the top 25 contributors for Structures are dominated by the would not be adversely CDF and/or LERF based on pile-soil-structure mode, the ISRS impacted by the Fussel-Vesely (FV) or with an FV appear to be similar at different correlation assumption. greater than 5.0E-03 were elevations within a given structure reviewed for potential correlations in the horizontal direction. with SSCs at different elevations in multi-story buildings expected that In response to a peer review team are expected to undergo similar question, RNP provided the results seismic responses on the different of an assessment of the top 25 elevations.
contributors to CDF and LERF. The assessment concluded that none of This correlation assumption in not the 25 components evaluated applicable to SSCs such as structures, tanks, cranes, and Page 179 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-2: Summary of Finding F&Os and Disposition Status Suggested SR F&O Description Basis Disposition Resolution would be impacted by the multi- piping. Additionally, liquefaction-elevation correlation criterion. induced settlement and lateral spreading failures are treated separately, so this assumption is not applicable to these hazards.
The remaining SSCs include conduit in the reactor auxiliary building (RAB), conduit in the containment building (CB), vacuum relief valves, air handlers and coolers for the AFW pump room, and Barksdale and Dwyer Instruments relays; these SSCs and their FVs were reviewed and are included in the F&O Resolution Notebook [71]. The impact of the correlation assumption on the remaining top contributors is dispositioned for RAB Conduit, CB Conduit, Vacuum Relief Valves, AFW Room Air Handlers and Coolers, Barksdale Relays and Dwyer Instruments Relays.
Therefore, this does not represent a change in methodology, scope, or capability as defined in Appendix 1-A of the ASME/ANS PRA Standard and is not considered an upgrade.
The response to this Finding meets the requirements of NTTF 2.1 seismic and Capability Category II of the Standard with NRC Page 180 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-2: Summary of Finding F&Os and Disposition Status Suggested SR F&O Description Basis Disposition Resolution Clarification. This finding is considered resolved for the purposes of NTTF 2.1.
For the safety-related valves, Duke has confirmed that the majority of Valves ranked as high capacity by them were purchased and installed walkdown were assigned a fragility to meet the generic capacities of using a capacity extracted from a 3.0 g horizontal and 2.0 g at the Robinson site standard document Robinson site, with the exception of for Pipe Stress Analysis Procedure FCV-6416 (SDAFW Pump which requires the following: Discharge Flow Control Valve).
Insufficient justification 'Valves must meet allowable valve Therefore, it is appropriate to use provided to support accelerations of 2 g vertical and 3 g Perform a review of the high generic capacities as an aggressive capacity for each horizontal direction for all valves ranked high to input to development of the safety-used for fragility dynamic conditions unless other confirm that all have a related valve fragility for all practical SR C-SFR-evaluation of high values are approved by the capacity of at least 3g in purposes. For FCV-6416, the C1/C-SFR- 29-1 ranked valves based on qualification.' The peer review team each horizontal direction walkdown report notes that the E1 Robinson site standard expressed some concern that not and 2g vertical. operator height of valve is 47 document. all plant valves may meet these inches, exceeding the GIP seismic requirements. It is possible guidelines [50] and the yoke is that exceptions were taken for laterally supported by a structural certain valves and lower capacities angle frame bolted to both sides of were adequate for design basis (for the operator. Because of this example, valves with cast iron unique support configuration, the yokes). valve fragility analysis identified yoke failure as the controlling failure mode in the station calculation for the Robinson Representative Fragilities Overview
[61] and developed the Page 181 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-2: Summary of Finding F&Os and Disposition Status Suggested SR F&O Description Basis Disposition Resolution corresponding HCLPF capacity of 0.67g, NOT based on the generic valve capacities. For the non-safety valves, Systems, Structures, and Components (SSCs) included in the station calculation for the Seismic Equipment List [64]
have been first reviewed to identify relevant valves. This review has found that there are a total of 40 non-safety valves. Each valve is re-assessed with respect to the appropriateness of its fragility basis. The objective of this assessment is to identify any non-safety valve case where the generic high capacity only applicable to the safety-related or Seismic Category I valve was inappropriately used as a capacity input for fragility development and to make appropriate adjustments or corrections to the existing fragility parameters of the valve. These values can be found in the station documentation for the station calculation for the RNP PRA Model Peer Review Resolution
[71].
This does not represent a change in methodology, scope, or capability as defined in Appendix 1-Page 182 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-2: Summary of Finding F&Os and Disposition Status Suggested SR F&O Description Basis Disposition Resolution A of the ASME/ANS PRA Standard and is not considered an upgrade.
The response to this finding meets the requirements of NTTF 2.1 seismic and Capability Category II of the Standard. This finding is considered resolved for the purposes of NTTF 2.1.
Duke conducted supplemental walkdowns of piping and tubing whose failure could lead to SLOCA in order to generate a more plant-In Robinson the station calculation specific fragility for SLOCA. The for the Robinson Representative Provide technical basis scope of the SLOCA walkdown Fragilities Overview [61] limited to support the conclusion included piping and tubing with basis is provided for the use of that generic fragilities inside diameters of 0.35 in. to 1.5 generic fragility values obtained from the EPRI SPRAIG EPRI SPRAIG [11] in. The walkdown results showed from the EPRI SPRAIG [11] for [11] are appropriate for values used for small that all of the piping/tubing was Robinson. Evaluate SR C-SFR- and medium LOCA with small and medium LOCA (SLOCA judged to have High seismic and MLOCA, respectively). The appropriateness of E1/C-SFR- 29-2 minimal technical capacity. Because no seismic E3 justification document concludes that the use is SPRAIG values concerns were identified by the representative and 'possibly wherever else they are walkdown, the SPRAIG fragility for conservative'. The Peer Review credited in the RNP piping (rather than the SPRAIG Team notes that based on current SPRA logic model (e.g., fragility for SLOCA) is now SPRA quantification results, distributed systems).
considered to be appropriate. With SLOCA shows up as a risk this fragility, SLOCA should not be contributor. risk-significant so further cost to develop a plant-specific fragility is considered to be unwarranted.
Section 4.4.2 of the station calculation for the Robinson Page 183 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-2: Summary of Finding F&Os and Disposition Status Suggested SR F&O Description Basis Disposition Resolution Representative Fragilities Overview
[61] has been updated to provide additional justification for the use of the SPRAIG fragility for medium loss of coolant accident (MLOCA).
During the area walkdowns for Robinson, the walkdown teams did not note any specific issues for the piping that could cause a MLOCA which would invalidate the use of the EPRI SPRAIG value. The contribution to SCDF and LERF from MLOCA is negligible. As such, the use of the EPRI SPRAIG fragility for MLOCA is judged to be appropriate and the costs to develop a plant-specific fragility is considered to be unwarranted.
Therefore, this does not represent a change in methodology, scope, or capability as defined in Appendix 1-A of the ASME/ANS PRA Standard and is not considered an upgrade.
The response to this Finding meets the requirements of NTTF 2.1 seismic and Capability Category II of the Standard. This finding is considered resolved for the purposes of NTTF 2.1.
Page 184 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-2: Summary of Finding F&Os and Disposition Status Suggested SR F&O Description Basis Disposition Resolution Excel files containing representative fragility calculations Backup calculations prepared for have been attached to the station the station calculation for the calculation for the Robinson Robinson Representative Fragilities In order to satisfy the Representative Fragilities Overview Overview [61] are not documented. intent of this supporting
[61] These files are identified in Methods are described and sample requirement, backup Section 2.3 of the station Seismic fragility backup calculations are provided; however, calculations prepared for calculation for the representative calculations not readily detailed calculations and / or the station calculation for fragility report [61].
accessible for future spreadsheets were not provided the Robinson SR C-SFR-updates and the peer review team for review. Representative Fragilities E4/C-SFR- 29-3 This does not represent a change maintenance of the For example, anchorage Overview [61] should be F1 in methodology, scope, or Robinson SPRA. calculations supporting SRT documented in order to capability as defined in Appendix 1-walkdown judgments, CDFM relay facilitate future PRA A of the ASME/ANS PRA Standard fragility calculations, and other applications, upgrades, and is not considered an upgrade.
calculation results for SEL and peer reviews.
The response to this finding meets equipment (e.g., valves) in the the requirements of NTTF 2.1 table in Appendix A.
Seismic and Capability Category II of the Standard. This finding is considered resolved for the purposes of NTTF 2.1.
Walkdown SEWS for fire protection Revise the walkdown piping located in the second floor of SEWS for the fire Fire protection piping Subsequent review determined that the RAB correctly identified the protection piping on the with vulnerable piping all fire protection piping on the presence of Victaulic couplings. second floor of the RAB SR C-SFR- joints dispositioned by second floor of the RAB is dry and Based on PRT walkdown, however, (Walkdown SEWS RAB-D5/C-SFR- 29-4 walkdown SRT with the SRT did not document other FLOOR 2-ALL-FIRE therefore flooding is not a concern.
F1 limited technical basis.
issues existing in the field including PIPING) to indicate that The SEWS for RAB-FLOOR 2-ALL-what appeared to be loose pipe the piping is dry and not FIRE PIPING has been revised straps and hard contact of this a flooding concern. accordingly (Attachment 2, Station piping with a primary water line. Similarly, update fire and Further, the SRT dispositioned the flood assessment Page 185 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-2: Summary of Finding F&Os and Disposition Status Suggested SR F&O Description Basis Disposition Resolution presence of Victaulic couplings as documentation as calculation for the Seismic
'OK' by judgment and that the line appropriate. Capacity Walkdown Report [53]
could leak a little. More technical basis should have been provided to This does not represent a change support SRT judgment. in methodology, scope, or capability as defined in Appendix 1-It is noted that subsequent input A of the ASME/ANS PRA Standard from the RNP fragility team and is not considered an upgrade.
provided during the peer review The response to this finding meets indicates that the subject fire line is the requirements of NTTF 2.1 dry; therefore, flooding concerns Seismic and Capability Category II are not credible. of the Standard. This finding is considered resolved for the purposes of NTTF 2.1.
During Peer Review Since Peer Review Team All of the specifics identified in the During Peer Review Team Team walkdown of the F&O were re-verified and the walkdown of a sample of SEL walkdown only reviewed Robinson site, a few SEWs were updated accordingly.
components at the Robinson site, a limited sample of SEL There were no issues found from instances were the following seismic interactions components, walkdown the verification walkdown that identified where the were either not documented or walkdown SEWS either SEWS documented would have invalidated the SPRA adequately dispositioned as non-did not appear to within the station model or any of its conclusions.
credible concerns on the walkdown capture all potential calculation for the SR C-SFR- SEWS:
29-5 seismic interactions or Seismic Capacity Based on the sampling conducted D7 by the extended condition judgment of the (1) Disposition of potential seismic Walkdown Report [53] walkdown and a cross check of the walkdown SRT to interaction between MS line PORV document disposition of (RV1-1) and structural steel as not should be reviewed on SEWS that document the original potentially credible an expanded sampling walkdowns, it is concluded that the a credible concern. The walkdown seismic interactions was SEWS includes a photo showing basis to confirm whether walkdowns were conducted not provided. there are any other properly with respect to evaluating proximity of the valve to the steel, but no discussion provided. instances where credible and documenting potential interactions.
Earthquake experience indicates seismic interactions were Page 186 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-2: Summary of Finding F&Os and Disposition Status Suggested SR F&O Description Basis Disposition Resolution that valves have failed due to either not noted or Further details on this F&O can be impact with adjacent structural properly dispositioned. found in the station calculation for steel. the RNP PRA Model Peer Review Additionally, walkdown Resolution [71].
(2) Flexibility of piping attached to documentation should be This does not represent a change CST for uplift displacements at tank updated as appropriate failure and disposition that potential for items identified in the in methodology, scope, or interaction of the gasifier tank is not basis portion of this F&O. capability as defined in Appendix 1-credible A of the ASME/ANS PRA Standard and is not considered an upgrade.
(3) No basis provided on the SEWS The response to this finding meets for SRT conclusion that SDAFW-the requirements of NTTF 2.1 PMP contains no credible soft targets. Seismic and Capability Category II of the Standard with NRC (4) CO2 tanks not identified as a clarification. This finding is possible seismic interaction with considered resolved for the MCC-5 on the walkdown SEWS purposes of NTTF 2.1.
(5) It does not appear as if walkdown SEWS for all DS DG equipment (e.g., DS DG Batteries) indicate the potential interaction with the gantry crane The piping in the BIT room area of Provide and document An area walkdown of the BIT room Potential differential the RAB was observed during PRT disposition that piping in was performed to determine if displacement of piping walkdown of the Robinson plant the BIT room area is piping crossing between the SR C-SFR- in the BIT room area of 29-6 the RAB was not and it appeared as if piping bounded by the review of Reactor Containment Building F2 traversing between the differential displacements (RCB) and RAB has sufficient explicitly evaluated by containment and the RAB had of piping systems exiting flexibility to accommodate relative walkdown inspection.
somewhat limited flexibility. Based containment documented building displacements. The on responses from the RNP fragility within the station walkdown found two instances of Page 187 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-2: Summary of Finding F&Os and Disposition Status Suggested SR F&O Description Basis Disposition Resolution (This F&O originated team, walkdown SRTs did not calculation for the piping crossing between structures, from SR C-SFR-D7) access the BIT room for purposes Seismic Fragility several instances of tubing of performing area walkdown Evaluation of the RCB spanning the gap, and flexible inspections for distributed systems. [94]. conduit powering the motors to Valves SI-870 A and B. The containment purge line was previously reviewed in Seismic Fragility Evaluation of the RCB [94].
The walkdown confirmed that this line should have a displacement capacity greater than that of the RCB piles. The hydrogen purge line was observed to have several bends with no attachment points near the RCB-RAB interface, and should also have a displacement capacity greater than that of the RCB piles. The tubing and conduit were observed to have sufficient flexibility providing displacement capacities greater than that of the RCB piles. A new area piping SEWS, RAB-BIT ROOM-PIPING, has been developed to address the effect of relative building displacements on piping in the BIT room (Attachment 2, Station calculation of the Seismic Capacity Walkdown Report [53]). Additional details on the BIT room piping walkdown may be obtained from this SEWS.
Page 188 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-2: Summary of Finding F&Os and Disposition Status Suggested SR F&O Description Basis Disposition Resolution Therefore, this does not represent a change in methodology, scope, or capability as defined in Appendix 1-A of the ASME/ANS PRA Standard and is not considered an upgrade.
The response to this Finding meets the requirements of NTTF 2.1 seismic and Capability Category II of the Standard with NRC Clarification Résumés for Jensen Hughes personnel participating in the seismic walkdowns (John Seismic Review Team resumes are Reddington and Rick Anoba) have provided as appropriate for SEL been added to Appendix C of the equipment walkdowns documented station calculation for the Seismic Provide team Walkdown team within Appendix C of the station Capacity Walkdown Report [53]
composition and composition and calculation for the Seismic Capacity qualifications for qualifications not Walkdown Report [53] ; however, Therefore, this does not represent SR C-SFR- participants other than 29-7 specified for Operator team composition and a change in methodology, scope, or F2 those listed in the station Pathway Walkdowns qualifications for Operator Pathway capability as defined in Appendix 1-calculation for the walkdowns documented within A of the ASME/ANS PRA Standard Seismic Capacity Calculation SEL Notebook [64], and is not considered an upgrade.
Walkdown Report [53]
Appendix C are not provided. The response to this Finding meets the requirements of NTTF 2.1 seismic and Capability Category II of the Standard. This finding is considered resolved for the purposes of NTTF 2.1.
Page 189 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-2: Summary of Finding F&Os and Disposition Status Suggested SR F&O Description Basis Disposition Resolution Station calculations RAB SSI Analysis [95] and RCB SSI In the vertical response analysis, Analysis [96] have been updated to the skin resistance offered by the include justification for using the sand layers is neglected compared vertical in-layer motion at Elevation to the tip resistance offered by the 159.2 ft as vertical input to the underlying clay layer and the piles seismic response analyses of the are interacting with the soil only at RCB and RAB. As noted, the SSI the bottom. Since the approach models represent the piles as does not credit any soil resistance connected vertical to the soil within along the length except at the the clay layer. The piles are not The technical basis for The additional bottom, using in-layer motion at connected to the soil vertically at using in-layer motions discussions with the depth an input to the vertical any other point because very little (defined at depth SPRA team during the response analysis seemed not side friction resistance during the corresponding to the on-site review as such appropriate for this configuration.
bottom of piles) in the explained in the basis of installation process was observed SR C-SFR- vertical direction is this finding should be and similar behavior would be 30-1 Based on the discussions with the B5 limited to justify the included in the expected in response to seismic SPRA team during the onsite compatibility between corresponding SSI vertical motion. Since the vertical review, it was judged that the in-the SSI modeling response analysis force transfer is primarily or layer motion at depth was the approach adopted. calculation. completely through the piles, the appropriate motion for this vertical foundation level is deemed configuration. The piles are to be the bottom of the pile model anchored into the clay layer for where the force enters the about 11' which in turn provides the structure. The control point for fixity such that the pile bottom vertical analyses is set at the same experiences the same input motion elevation at the bottom of pile. The as the soil column at depth bottom of the piles will move in irrespective of the pile soil response to the movement of the interaction above the clay layer.
hard clay layer and the movement of the hard clay layer will be affected by the response of the soil above regardless of whether or not the piles are connected to the soil Page 190 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-2: Summary of Finding F&Os and Disposition Status Suggested SR F&O Description Basis Disposition Resolution above this point in the model.
Therefore an in-layer (or in-column) motion at Elevation 159.2 ft. is proper for use. Additional details may be obtained from station calculations RAB SSI Analysis [95]
and RCB SSI Analysis [96].
Therefore, this does not represent a change in methodology, scope, or capability as defined in Appendix 1-A of the ASME/ANS PRA Standard and is not considered an upgrade.
The response to this Finding meets the requirements of NTTF 2.1 seismic and Capability Category II of the Standard. This finding is considered resolved for the purposes of NTTF 2.1.
Assess the degree of The fragility group SF-TB-CLASS-3 The collapse of the TB- correlation between As stated in the Finding, some represents the collapse due to Class III building has these two failure modes degree of correlation exists excessive first story drifts and SF-been included in the S- and calculate the between the Class 3 Turbine TB-CLASS-3-POUND represents PRA model as two combined fragility. Building shaking fragility and the collapse due to pounding with SR C-SFR- different fragility groups pounding fragility. The composite 30-2 the adjacent RAB.
E3 (SF-TB-CLASS If the failure modes are fragilities for two bounding cases in POUND and SF-TB- assessed to be terms were compared for The consequence of both these CLASS-3). completely correlated, correlation: zero correlation fragility groups is the collapse of then model the fragility (independent) and perfect the turbine building but their group with the lowest correlation governing failure modes (excessive fragility.
drift and pounding due to relative Page 191 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-2: Summary of Finding F&Os and Disposition Status Suggested SR F&O Description Basis Disposition Resolution displacement) are not completely As the pounding fragility is much independent of each other. larger than the shaking fragility over the hazard range of interest, there is virtually no difference between the pounding fragility curve and the perfect correlation composite fragility curve.
The composite fragility curve for the zero correlation case is slightly more conservative than the perfect correlation case, but the degree of conservatism is so small that they are equal for all practical purposes.
While realistically these two failure modes should be at least partially correlated, any partial correlation curves are bounded by these two cases; and therefore, partial correlation possibilities are not an important consideration. This justifies the current modeling assumption of independence between the pounding and shaking fragilities in the as-built and as-operated model.
This does not represent a change in methodology, scope, or capability as defined in Appendix 1-A of the ASME/ANS PRA Standard Page 192 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-2: Summary of Finding F&Os and Disposition Status Suggested SR F&O Description Basis Disposition Resolution and is not considered an upgrade.
The response to this finding meets the requirements of NTTF 2.1 seismic and Capability Category II of the Standard. This finding is considered resolved for the purposes of NTTF 2.1.
Focused Scope Peer Review Findings and Resolutions SR C-SFR-E3 Page 193 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-2: Summary of Finding F&Os and Disposition Status Suggested SR F&O Description Basis Disposition Resolution Page 194 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-2: Summary of Finding F&Os and Disposition Status Suggested SR F&O Description Basis Disposition Resolution Duke agrees with the statement As noted in Section 6.4.1 of the that the dam fragility included the station calculation for seismic aleatory uncertainty associated with fragility evaluation of the Robinson the dam/foundation soil properties Dam [67], the confidence levels in Figure 13 capture the variability in and the earthquake ground SFR-F3 requires that A more thorough motions, but neglected calculating sources of model the dam soil model properties and discussion of sources of the epistemic uncertainties such as uncertainty and related ground motion included within the uncertainty in estimating those associated with the dam soil assumptions associated LHS conducted. These are judged the fragility of Robinson with the fragility analysis to be the dominant sources of Dam should be model, the relationship between be appropriately uncertainties in the fragility. There developed consistent crest deformation and crack depth SR C-SFR-2-2 documented. There are are other sources of uncertainty with the verbal or in the relationship between crack F3 that may affect the confidence several sources of explanation provided by exposure and probability of failure.
uncertainty that should levels, including uncertainty the fragility vendor during There are several reasons that be acknowledged and associated with the estimated a conference call with the justify Dukes judgement that the discussed further. depth of transverse cracking peer review team. aleatory uncertainties are sufficient conditioned on crest displacement for this fragility assessment of the (Figure 17 of Appendix B [67]) and Robinson Dam:
the estimated probability of failure conditioned on crack exposure 1. The aleatory uncertainty (Figure 5). Additional discussion calculated from the Latin and evaluation of these sources is Hypercube Simulation are very needed before concluding that the large, they incorporate a large variability in soil properties and variability in the soil parameters Page 195 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-2: Summary of Finding F&Os and Disposition Status Suggested SR F&O Description Basis Disposition Resolution ground motions are the dominant and in the earthquake ground sources. motions.
- 2. Epistemic uncertainly certainly exists, but the calculation of those uncertainties would entail a very large effort and a much longer schedule than is available for this Robinson SPRA. Our collective judgment of the SPRA team is that additional large effort is not warranted. Based on our experience the epistemic uncertainty will be much smaller than the calculated aleatory uncertainty for the dam. And once this epistemic uncertainty is calculated, it will be SRSSed with the larger aleatory uncertainty which will result in a composite uncertainty close to the existing aleatory uncertainty.
- 3. Duke performed a completely independent method of developing a seismic fragility using earthquake experience data. This independent fragility was developed at the lower end of the ground motions but the fragility overlapped the more Page 196 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-2: Summary of Finding F&Os and Disposition Status Suggested SR F&O Description Basis Disposition Resolution detailed FLAC analysis fragility in the 0.3 to 0.35 g part of the spectrum. These two completely independent fragility derivations resulted in nearly identical best estimate values for the fragility which provides a very strong case that the epistemic uncertainties that were not evaluated as part of the fragility derived from the FLAC analyses will not bias the best estimate results. In addition the uncertainties associated with the earthquake experience derived fragilities (these uncertainties include both aleatory and epistemic variabilities based on the use of a large number of dam performance data from a large number of large earthquakes) are smaller than the aleatory uncertainties derived from the Latin Hypercube Simulations. Which provides additional justification that these large aleatory uncertainties are sufficient for this fragility.
Finally, it is Dukes judgement that even if we were to estimate some Page 197 of 198
H. B. ROBINSON SEISMIC PROBABILISTIC RISK ASSESSMENT
SUMMARY
REPORT Table A-2: Summary of Finding F&Os and Disposition Status Suggested SR F&O Description Basis Disposition Resolution additional epistemic uncertainty to add into the existing fragility, the overall results would not be expected to change appreciably.
Therefore, this does not represent a change in methodology, scope, or capability as defined in Appendix 1-A of the ASME/ANS PRA Standard and is not considered an upgrade.
The response to this Finding meets the requirements of NTTF 2.1 seismic and Capability Category II of the Standard. This finding is considered resolved for the purposes of NTTF 2.1.
Page 198 of 198