ML13043A013

Pilot Submittal and Request for Exemption for a Risk-Informed Approach to Resolve Generic Safety Issue (GSI)-191

ML13043A013
Person / Time
Site:	South Texas
Issue date:	01/31/2013
From:	Crenshaw J South Texas
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NOC-AE-13002954
Download: ML13043A013 (111)

Text

V SMmIFARD Nuclear Operating Company South Texas Pro/ed Ekctric Generating Station PO Bax 289 Wadsworth. Texas 77483 January 31, 2013 NOC-AE-13002954 10 CFR 50.12 10 CFR 50.46 Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555-0001 South Texas Project Units 1 and 2 Docket Nos. STN 50-498 and STN 50-499 STP Pilot Submittal and Request for Exemption for a Risk-Informed Approach to Resolve Generic Safety Issue (GSI)-191

References:

1.

Letter, J. W. Crenshaw, STPNOC, to NRC Document Control Desk, "Status of the South Texas Project Risk-Informed (RI) Approach to Resolve Generic Safety Issue (GSI)-1 91," NOC-AE-1 1002775, dated December 14, 2011 (ML11354A386)

2.

Letter, D. W. Rencurrel to NRC Document Control Desk, "GSI-191 Resolution Path Schedule and Commitment Changes," dated June 4, 2012, NOC-AE-1 2002858

3.

Letter, John C. Butler, NEI, to William H. Ruland, NRC, "GSI-191 - Current Status and Recommended Actions for Closure," dated May 4, 2012 (ML12142A316)

4.

Commission SECY Paper, "Closure Options for Generic Safety Issue - 191, Assessment of Debris Accumulation on Pressurized-Water Reactor Sump Performance," SECY-12-0093, dated July 9, 2012 (ML121320270)

5.

Letter, J. E. Dyer, NRC, to A. W. Harrison, STPNOC, Response to letter requesting an exemption of fees, AE-NOC-11002079, dated April 15, 2011 (ML111050388)

This submittal is the request for NRC review and approval of an exemption to enable the STP Nuclear Operating Company (STPNOC) to apply a piloted risk-informed approach for closure of Generic Safety Issue (GSI)-1 91, "Assessment of Debris Accumulation on Pressurized-Water Reactor Sump Performance." This submittal supports closure of Generic Letter (GL) 2004-02, "Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors," for the South Texas Project (STP) Units 1 and 2.

STPNOC seeks NRC approval based on a determination that the risk associated with the postulated failure mechanisms due to GSI-191 concerns meets the acceptance guidelines in Regulatory Guide (RG) 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis."

STI 33648174

NOC-AE-1 3002954 Page 2 of 5 This submittal includes a request for exemption from part of NRC's regulations and a description of the STP risk-informed approach. Also included for information purposes are changes to the STP Updated Final Safety Analysis Report (UFSAR) to be implemented pursuant to NRC approval of the risk-informed approach and the exemption request.

By Reference 1, STPNOC submitted to the NRC the preliminary results showing that the risks, Core Damage Frequency (CDF) and Large Early Release Frequency (LERF), associated with GSI-191 concerns are in Region III, "Very Small Changes," of RG 1.174 acceptance guidelines, and notified the NRC of the intent to seek exemption from certain requirements of 10 CFR 50.46, "Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors." By Reference 2, STPNOC committed to STP Units 1 and 2 piloting a risk-informed approach and to seek exemption from certain regulatory requirements to support closure for GSI-191.

Additional details of the STP risk-informed approach and schedule are discussed in References 3 and 4, and in Enclosure 2, "Evaluation of Generic Safety Issue-191 Closure Options," and, "Risk-Informed Approach to Address GSI-191, South Texas Project" of Reference

4. The NRC staff plans to use STPNOC as a pilot for other licensees choosing to use this approach (References 4 and 5). The STP piloted risk-informed approach is expected to result in substantial benefit to both the NRC and industry in support of the development and implementation of risk-informed resolution of GSI-191.

The STP piloted risk-informed approach to closure for GSI-1 91 applies the STP Probabilistic Risk Assessment (PRA) model to quantify the risk associated with GSI-191 concerns by calculating the difference in risk for two cases:

the actual plant configuration for STP Units 1 and 2, risk informed to model the failure mechanism associated with the concerns raised by GSI-191, and a hypothetical STP plant, identical to the actual model except for the assumption that it is not subject to the concerns raised by GSI-1 91.

The risk associated with GS1-191 concerns includes the effects on long-term cooling due to debris accumulation on Emergency Core Cooling System (ECCS) and Containment Spray System (CSS) sump strainers in recirculation mode, as well as core flow blockage due to in-vessel effects, following loss of coolant accidents (LOCAs). A full spectrum of postulated LOCAs is analyzed, including double-ended guillotine breaks (DEGBs) for all pipe sizes up to the largest pipe in the reactor coolant system (RCS). To inform the PRA with risk insights, the physical processes are modeled as realistically as possible, using results from industry and plant-specific testing, and applying some conservatism, where appropriate. The changes to CDF and LERF associated with GSI-191 concerns are then compared to RG 1.174 acceptance guidelines. provides the general methodology for the proposed risk-informed approach to closure of GSI-1 91, consistent with RG 1.174 guidance. This enclosure describes the required inputs to the PRA model, the basic structure for appropriately modeling the inputs, and performance criteria used to calculate the risk. provides a request for exemption from parts of certain regulatory requirements in accordance with the provisions of § 50.12, and provides justification for the exemption based on the results of the risk-informed approach demonstrating for STP Units 1 and 2 that the

NOC-AE-13002954 Page 3 of 5 calculated risk associated with GSI-191 concerns is in Region Ill, "Very Small Changes," of RG 1.174 acceptance guidelines. The exemption request addresses regulatory requirements, including Appendix A to 10 CFR Part 50 General Design Criteria (GDC), that concern the ECCS and CSS functions for emergency core cooling, containment heat removal, and containment atmosphere cleanup:

§ 50.46(b)(5), Long-term cooling Criterion 35 - Emergency core cooling Criterion 38 - Containment heat removal Criterion 41 - Containment atmosphere cleanup The exemption request also addresses requirements that concern crediting the CSS with reducing the accident source term:

§ 50.67, Accident source term Criterion 19 - Control room provides the proposed changes to the STP Units 1 and 2 licensing basis, pursuant to NRC approval of the risk-informed approach and exemption request. The current licensing basis for the adequacy of ECCS to meet the criteria of 10 CFR 50.46, including the Appendix K Large-Break Loss-of-Coolant Accident analysis and the associated Chapter 15 accident analysis, remain unchanged. The current licensing basis for demonstrating compliance with the other requirements described above similarly remains unchanged. follows the structure, content and documentation requirements of RG 1.174, and provides references to other supporting documentation. This enclosure provides the details of how the STP piloted approach meets the general guidance and conforms to the risk-informed principles included in RG 1.174:

Meets the current regulations except as provided in the request for partial exemption.

Is consistent with a defense-in-depth philosophy.

Maintains sufficient safety margins.

Shows that for STP Units 1 and 2 the change in risk associated with GSI-1 91 concerns is very small, approximately 1.1E-8/yr (delta CDF) and 8.6E-12/yr (delta LERF).

Includes provisions for monitoring the impact of the change.

To support the completion of work and resolution schedule for closure of GSI-191 as described in Reference 4, STPNOC seeks approval for the risk-informed approach and exemption request by December 2014.

There are no commitments in this letter.

If there are any questions regarding this submittal, please contact Jamie Paul at 361-972-7344, or me at 361-972-7074.

NOC-AE-1 3002954 Page 4 of 5 I declare under penalty of perjury that the foregoing is true and correct.

Executed on if ?171-John W. Crenshaw Vice President Projects, Outages & IT ccc

Enclosures:

1.

STP Piloted Risk-Informed Approach to Closure for GS1-191

2.

Request for Exemption for STP Piloted Risk-Informed Approach to Closure for GSI-191

3.

Changes to the South Texas Project Units 1 and 2 Licensing Basis (Information Only)

4.

Risk-Informed Closure of GSI-191, Volume 1.0, Project Summary

NOC-AE-13002954 Page 5 of 5 cc: (paper copy)

(electronic copy)

Regional Administrator, Region IV U. S. Nuclear Regulatory Commission 1600 East Lamar Boulevard Arlington, TX 76011-4511 Balwant K. Singal Senior Project Manager U.S. Nuclear Regulatory Commission One White Flint North (MS 8 B13) 11555 Rockville Pike Rockville, MD 20852 NRC Resident Inspector U. S. Nuclear Regulatory Commission P. 0. Box 289, Mail Code: MN1 16 Wadsworth, TX 77483 C. M. Canady City of Austin Electric Utility Department 721 Barton Springs Road Austin, TX 78704 A. H. Gutterman, Esquire Morgan, Lewis & Bockius LLP Balwant K. Singal Stewart Bailey Jack Davis Robert Elliott Michael Markley John Stang U. S. Nuclear Regulatory Commission John Ragan Chris O-Hara Jim von Suskil NRG South Texas LP Kevin Polio Richard Pefia City Public Service Peter Nemeth Crain Caton & James, P.C.

C. Mele City of Austin Richard A. Ratliff Alice Rogers Texas Department of State Health Services

NOC-AE-1 3002954 ENCLOSURE 1 STP Piloted Risk-Informed Approach to Closure for GSI-191 NOC-AE-1 3002954 Page 1 of 6 STP Piloted Risk-Informed Approach to Closure for GSI-191 Introduction This enclosure provides the general methodology for the proposed risk-informed approach to closure of GSI-1 91, consistent with RG 1.174 guidance. The required inputs to the plant-specific probabilistic risk assessment (PRA) model, the basic structure for modeling the inputs, and performance criteria used to calculate the risk are discussed below, and in more detail in Enclosure 4.

Backqround Generic Safety Issue (GSI)-191, "Assessment of Debris Accumulation on PWR Sump Performance," concluded that debris could clog the containment sump strainers in PWRs, leading to the loss of net positive suction head for the Emergency Core Cooling System (ECCS) and Containment Spray System (CSS) pumps. The NRC issued Generic Letter (GL) 2004-02, "Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors" requesting that licensees address the issues raised by GSI-1 91. GL 2004-02 was focused on demonstrating compliance with 10 CFR 50.46.

In response, the industry implemented plant modifications, such as installing larger sump strainers and removing debris generating (fibrous) insulation from containment, and other compensatory actions to reduce the risk of strainer clogging. Considerable effort has also been made to reduce the uncertainties and conservatisms in the standard models used to assess GSI-191 concerns.

Summary of the STP approach The STP piloted risk-informed approach to closure for GSI-191 applies the plant-specific Probabilistic Risk Assessment (PRA) model to calculate the difference in risk (delta risk) between the actual plant configuration subject to the concerns raised by GSI-191 and a hypothetical plant configuration not subject to GSI-191, but otherwise identical. The difference in risk is a quantification of the risk associated with GSI-191 concerns. This risk includes the effects on long-term cooling due to debris accumulation on the ECCS and CSS containment sump strainers and the in-vessel effects following LOCAs that require recirculation flow from the containment sump to mitigate the event. The quantification of the risk associated with GSI-1 91 concerns conservatively defines the change to be evaluated, as discussed in Regulatory Guide (RG) 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis."

Section II provides the methodology for calculating the risk associated with GSI-1 91 concerns using the probabilistic risk assessment (PRA) model. A full spectrum of postulated LOCAs is analyzed, including double-ended guillotine breaks (DEGBs) for all pipe sizes up to and including the design basis accident (DBA) LOCA. The required inputs to the PRA, the basic structure for modeling the inputs, and performance criteria used to calculate the risk, are described. The physical processes are modeled as NOC-AE-13002954 Page 2 of 6 realistically as possible, using results from industry and plant-specific testing, and applying some conservatism, where appropriate. The risk due to GSI-191 concerns is then shown to meet RG 1.174 acceptance guidelines for changes to Core Damage Frequency (CDF) and Large Early Release Frequency (LERF).

This risk-informed approach is expected to be applicable to plants with substantial fibrous insulation, and also may be beneficial to plants with low to medium fibrous insulation. If the results show high risk, they may be used to assess and prioritize those plant modifications with the highest risk benefit.

Methodology Define the Proposed Change Although this approach does not necessarily result in physical changes to the facility, it is expected to result in changes to the plant's licensing basis. The STP risk-informed approach to closure of GSI-191 evaluates a full spectrum of pipe breaks, including DEGBs up to and including the DBA LOCA. The approach uses a RG 1.200 compliant PRA model, using inputs as described below, to calculate the risk associated with concerns raised by GSI-1 91 based on a comparison of the as-built, as-operated plant to an identical plant with the exception that it is not subject to the phenomena associated with GSI-191 concerns. This difference in risk defines the risk associated with GSI-191 concerns, and the change to be evaluated.

The plant licensing basis considers the requirement for ECCS to satisfy the criterion for long-term core cooling following a LOCA as provided in 10 CFR 50.46(b)(5), and requires ECCS to operate with high probability following a LOCA. Using a risk-informed approach to address the concerns of GSI-1 91, the probability and uncertainty associated with the operation of the ECCS to maintain long-term cooling following a LOCA is quantified. Based on the risk associated with GSI-191 concerns meeting the acceptance guidelines in RG 1.174, with appropriate supporting engineering analysis, justification is provided for a change to the licensing basis which, in addition to the plant's existing licensing basis, provides reasonable assurance for compliance with certain regulatory requirements, and closure for GSI-1 91.

The regulatory requirements include those associated with ECCS and CSS functions for emergency core cooling, containment heat removal, and containment atmosphere cleanup, as provided in § 50.46(b)(5), Appendix A to 10 CFR Part 50 General Design Criteria (GDC) 35, GDC 38, and GDC 41. Also included are the regulatory requirements that concern crediting the CSS with reducing the accident source term, as provided in

§ 50.67 and GDC 19. Application of this methodology is accompanied by a request for exemption that addresses these regulatory requirements.

Engineering Analysis and PRA modeling The method of analysis for the risk-informed approach uses an integrative approach to explicitly provide the probabilities for post-LOCA events. This is accomplished by NOC-AE-1 3002954 Page 3 of 6 modeling the underlying physical phenomena of the basic events and by propagating uncertainties in the physical models.

To determine the risk associated with GSI-191 concerns, under the framework of RG 1.174, the STP piloted risk-informed approach to closure for GSI-191 applies the plant-specific PRA model to calculate the difference in risk (delta risk) for two cases:

Case 1: the actual plant configuration, risk informed to model the failure mechanism associated with the concerns raised by GSI-191, and Case 2: a hypothetical plant assuming no failure mechanisms associated with the concerns raised by GSI-191, otherwise identical to the actual plant.

The risk associated with GSI-191 concerns is determined for the as-built, as-operated plant (Case 1) and compared with the risk of a plant not subject to the concerns raised by GSI-191 (Case 2), as described in Table 1. The plant-specific PRA model is informed with risk insights to address the risk associated with failure modes associated with GSI-191 concerns, as described in Table 2.

To apply the inputs, the demand recirculation failure probability in the plant-specific PRA model is replaced with basic events (strainer failures, core flow blockage with chemical effects, and boron precipitation in the core), and failure modes leading to core damage are explicitly modeled, excluding those that were previously addressed for the plant using deterministic evaluations.

Failure probabilities and associated uncertainties determined in the supporting engineering analysis provide inputs to the three new top events added to the PRA to accommodate composite GS1-191 failure processes (sump strainer failure, core flow blockage, and boron precipitation in the core). The outcome of a full spectrum of LOCA events is tested against appropriate performance thresholds for the top events, as shown in Table 2.

Using the inputs noted above, the PRA uses risk insights for the failure modes resulting from GSI-191 concerns. The approach defines the change and performs engineering analysis and PRA assessments, which are principle elements of risk-informed, plant-specific decision-making as discussed in RG 1.174.

Ill.

Conclusions The risk-informed approach results in the calculation of the risk associated with GSI-1 91 concerns, based on the difference in risk, CDF and LERF, between the as-built, as-operated plant with debris generating insulation, and the hypothetical plant with the insulation removed and therefore not subject to the failure mechanisms associated with GSI-191 related phenomena.

The PRA analysis yields results that are compared to the acceptance guidelines defined by Regulatory Guide 1.174 to show that long-term cooling is ensured with high probability, and provide a basis for NRC approval of the requested licensing actions for closure of GSI-191.

NOC-AE-1 3002954 Page 4 of 6 IV.

References

1)

Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis" (ML100910006)

2)

NEI 04-07, "Pressurized Water Reactor Sump Performance Evaluation Methodology," Volume 1 "Pressurized Water Reactor Sump Performance Evaluation Methodology," Revision 0, dated December 2004 (ML050550138)

3)

NEI 04-07, "Pressurized Water Reactor Sump Performance Evaluation Methodology," Volume 2 "Safety Evaluation by the Office of Nuclear Reactor Regulation Related to NRC Generic Letter 2004-02, Revision 0, December 6, 2004," Revision 0, dated December 2004 (ML050550156)

4)

Regulatory Guide 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities" V.

List of Tables Table 1: General Methodology for Determining Risk Associated with GSI-1 91 Concerns Table 2: Modeling Basic Events, Failure Modes, and Top Events with Performance Thresholds NOC-AE-13002954 Page 5 of 6 Table 1: General Methodology for Determining Risk Associated with GSI-191 Concerns Case 1:

Evaluate the risk associated with the concerns raised in GS-1 91 for the as-built, as-operated facility as described in the current licensing basis using a plant-specific, RG 1.200 compliant probabilistic risk assessment (PRA). The PRA is based on realistic assessments to the extent practical and contains conservative assumptions where appropriate. Modeling of basic events, failure modes, and new top events to accommodate composite GSI-191 failure processes with appropriate performance thresholds is described in Table 2.

The inputs to the risk model encompass the concerns raised in GSI-191, including the major topical areas discussed in NEI 04-07 (Reference 2), as appropriate:

o pipe break characterization o

debris generation/zone of influence (ZOI), including latent debris o

debris transport o

chemical effects o

strainer head loss, including structural margin o

air intrusion o

debris penetration o

ex-vessel downstream effects o

in-vessel downstream effects o

boron precipitation For each input to the risk model, any differences between the methods to be used in the model and NRC-approved methods are defined (refer to Reference 3 for an example).

For each input to the risk model, an uncertainty quantification process is used to add detail (basic events, refer to Table 2) to the PRA model for the LOCA initiating sequences. Examples of appropriate sources of information include, but are not limited to:

o applicable risk assessments o

results obtained from generic industry and/or plant-specific testing o

expert elicitation o

assumptions, realistic or conservative o

qualitative insights based on engineering judgment For each input to the risk model, interdependencies between the inputs to the model are considered and appropriately described in the risk model.

The risk is determined using a PRA that meets the necessary requirements identified in RG 1.200 (Reference 4), including the capability to model a full spectrum of LOCA events, and the capability for Level 1 and Level 2 risk assessments, including internal and external events.

Case 2:

Evaluate the risk assuming no long term cooling failure contributors associated with GSI-191 concerns, and assuming no additional failures. Other than the basic events details associated with GSI-191 concerns, the Case 2 assessment model is identical to model used for Case 1.

Calculate the risk associated with GSI-191 concerns:

The risk associated with GSI-191 concerns is the difference in risk, CDF and LERF, between Case 1 and Case 2 for comparison with the acceptance guidelines in RG 1.174, Section 2.4.

NOC-AE-13002954 Page 6 of 6 Table 2: Modeling Basic Events, Failure Modes, and Top Events with Performance Thresholds Using the inputs noted below, applied within the framework described in Table 1, the PRA uses risk insights to address the risk associated with failure modes resulting from GSI-191 concerns.

Basic Events In the plant-specific PRA model, the demand recirculation failure probability is replaced with the following:

Pressure drop due to debris build-up on the sump strainers with chemical effects resulting in loss of net positive suction head (NPSH) margin for pumps Strainer mechanical collapse Air ingestion through the sump strainers Core blockage with chemical effects Boron precipitation in the core Failure Modes For input into the plant-specific PRA, accident sequences from a full spectrum of LOCAs are analyzed in a realistic time-dependent manner with uncertainty propagation to determine the probabilities of various failures potentially leading to core damage.

The failure modes shall be explicitly modeled in the PRA analysis, except for failure modes that were addressed with no issues of concern as part of previous deterministic evaluations for the plant.

Top Events and Performance Thresholds Failure probabilities and associated uncertainties determined in the supporting engineering analysis are passed to the plant-wide PRA, which determines the incremental risk associated with GSI-191 failure modes with three new top events added to accommodate composite GSI-191 failure processes. The engineering analysis supports the three new top events by testing the outcome of every postulated LOCA scenario against seven performance thresholds, discussed in detail in Enclosure 4, and summarized below.

New Top Events Performance Thresholds

1. Failure at sump strainers

1.

Strainer DP > NPSH margin

2.

Strainer DP > P-buckle

3.

Strainer F-void > 0.02

2.

Boron precipitation in the core

4.

Core fiber load > cold leg break fiber limit for boron precipitation

5.

Core fiber load > hot leg break fiber limit for boron precipitation

3.

Core flow blockage

6.

Core fiber load > cold leg break fiber limit for flow blockage

7.

Core fiber load > hot leg break fiber limit for flow blockage

NOC-AE-1 3002954 ENCLOSURE 2 Request for Exemption for STP Piloted Risk-Informed Approach to Closure for GSI-191 NOC-AE-13002954 Page 1 of 11 Request for Exemption for STP Piloted Risk-Informed Approach to Closure for GSI-191 Purpose and Objective of the Exemption Request Pursuant to 10 CFR 50.12(a), STP Nuclear Operating Company (STPNOC) requests an exemption from certain requirements specified under § 50.46 and Appendix A to 10 CFR Part 50, "General Design Criteria." The exemption request is for implementation of a risk-informed approach to closure for Generic Safety Issue (GSI)-1 91, "Assessment of Debris Accumulation on Pressurized-Water Reactor Sump Performance," to support closure of Generic Letter (GL) 2004-02 for South Texas Project (STP) Units 1 and 2, and corresponding changes to the licensing basis subject to NRC approval. Under § 50.12, a licensee may request and the NRC may grant exemptions from the requirements of 10 CFR Part 50 which are authorized by law, will not present an undue risk to the public health and safety, are consistent with the common defense and security, and special circumstances are present. STP Units 1 and 2 are the lead plants for the proposed risk-informed approach to resolution of GSI-191.

STPNOC seeks exemption from criterion (b)(5), "Long-term cooling," as specified in

§ 50.46, "Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors," specifically related to the performance of Emergency Core Cooling System (ECCS) and Containment Spray System (CSS) during the recirculation mode in containment following loss of coolant accidents (LOCAs). As stated in § 50.46(d), "The criteria set forth in paragraph (b), with cooling performance calculated in accordance with an acceptable evaluation model, are in implementation of the general requirements with respect to ECCS cooling performance design set forth in this part, including in particular Criterion 35 of appendix A."

As such, STPNOC also seeks an exemption from part of Appendix A of 10 CFR Part 50, General Design Criterion (GDC) 35, as it relates to ECCS performance criterion (b)(5). STPNOC also seeks an exemption from part of other requirements, as specified in the "Regulatory Requirements to Which the Exemption Would Apply" section below.

The STP risk-informed approach, described in Enclosure 1 and in detail in Enclosure 4, uses the STP PRA to quantify the residual risk from those issues related to GSI-1 91 concerns which have not been resolved using other methods. The supporting engineering analysis, including evaluation of defense-in-depth and safety margin, is developed to conform to RG 1.174.

STPNOC requests NRC approval based upon the determination that the risk-informed method is acceptable, and that the calculated risk meets the acceptance guidelines of RG 1.174.

2.

Background

GSI-191 concerns the possibility that debris generated during a LOCA could clog the containment sump strainers in pressurized-water reactors (PWRs) and result in loss of net positive suction head (NPSH) for the ECCS and CSS pumps, impeding the flow of NOC-AE-1 3002954 Page 2 of 11 water from the sump. GL 2004-02 requested licensees to address GSI-1 91 issues, focused on demonstrating compliance with the ECCS acceptance criteria in § 50.46.

GL 2004-02 requested licensees to perform new, more realistic analyses using an NRC-approved methodology and to confirm the functionality of the ECCS and CSS during design basis accidents that require containment sump recirculation.

STP Units 1 and 2 have implemented compensatory and mitigative measures in response to Bulletin 2003-01 and GL 2004-02 to address the potential for sump strainer clogging and other concerns associated with GSI-191. Larger containment sump strainers have been installed that greatly reduce the potential for loss of net positive suction head (NPSH). Additional compensatory measures such as operating procedures and instrumentation to monitor core level and temperature, and actions taken by operators if core blockage is indicated, have been described.

In response to SECY-10-01 13, "Closure Options for Generic Safety Issue - 191, Assessment of Debris Accumulation on Pressurized Water Reactor Sump Performance,"

the Commission issued Staff Requirements Memorandum (SRM)-SECY-10-0113, "Closure Options for Generic Safety Issue (GSI) - 191, Assessment of Debris Accumulation on Pressurized Water Reactor (PWR) Sump Performance," directing the staff to consider alternative options for resolving GSI-1 91 that are innovative and creative, as well as risk-informed and safety conscious, while the industry completed testing in 2011. Subsequently, STPNOC, through interactions with the staff, developed a risk-informed approach for the resolution of GSI-1 91 using the methods described in Regulatory Guide (RG) 1.174. By Reference 6, STPNOC submitted to the NRC the preliminary results showing that the risks, Core Damage Frequency (CDF) and Large Early Release Frequency (LERF), associated with GSI-191 concerns are in Region III, "Very Small Changes," of RG 1.174 acceptance guidelines, and notified the NRC of the intent to seek exemption from certain requirements of § 50.46. SECY-12-0093, "Closure Options for Generic Safety Issue - 191, Assessment of Debris Accumulation on Pressurized-Water Reactor Sump Performance," described the staff plans to use STPNOC as a pilot for other licensees choosing to use this approach.

As stated in SECY-12-0093, the STP piloted risk-informed approach is a graded approach with actions and schedule based on the amount of installed fibrous insulation, and is consistent with the risk management goal of NUREG-2150, "A Proposed Risk Management Regulatory Framework" (Reference 9). The approach incorporates defense-in-depth measures to mitigate the residual risk of strainer or in-vessel issues that have not been resolved, as applicable, during the time required to provide closure for GSI-191.

Confirming the preliminary results, the results from the analysis provided in Enclosure 4 show that the risk associated with GSI-191 concerns is in Region III, "Very Small Changes," of RG 1.174. As such, no additional physical modifications to the South Texas Project (STP) Units 1 and 2 are proposed.

3.

Regulatory Requirements to Which the Exemption Would Apply The exemption request is limited to compliance with certain regulatory requirements specific to addressing the concerns associated with GSI-191. This exemption request is NOC-AE-1 3002954 Page 3 of 11 submitted as part of a risk-informed approach for using probabilistic risk assessment (PRA) to provide closure for GSI-1 91, developed to conform to RG 1.174 and the principles for risk-based applications, i.e., that it is consistent with the existing defense-in-depth framework, has no significant impact on safety margin, and can be implemented with appropriate monitoring capability.

STPNOC seeks exemption from part of certain regulatory requirements for the purpose of closure of GSI-191. STPNOC seeks approval of the request for exemption on the basis that the STP probabilistic risk assessment (PRA) results and supporting engineering analysis are acceptable and show that the residual risk associated with GSI-191 meets the acceptance guidelines of RG 1.174, consistent with the Commission's Safety Goals for public health and safety.

The request for exemption applies to the regulatory requirements listed below, which are discussed in GL 2004-02 and associated regulatory guidance. STPNOC seeks exemption to the extent that the applicable regulatory requirement requires additional calculation or other analysis or evaluation to demonstrate compliance, considering only the concerns raised by GSI-191 and specifically the potential effects on plant performance during the recirculation mode in containment following a LOCA, beyond those already provided as part of the current licensing basis. The proposed changes to the licensing basis, shown in Enclosure 3, provide closure to GSI-191 on the basis that the associated risk is shown to meet the RG 1.174 acceptance guidelines, and that in conjunction with the existing licensing basis, the small risk demonstrates adequate compliance with each of the regulatory requirements. The regulatory requirements listed are associated with the ECCS and CSS required functions in recirculation mode following the DBA LOCA that are potentially affected by postulated effects on containment sump performance due to GSI-191 concerns.

Since sump performance and in-vessel effects were not explicitly described in the regulatory requirements, there is no specific language from which to request exemption.

Specific exemption is requested from the applicable regulations as summarized below:

STP requests exemption from the requirement to use a bounding calculation or other deterministic method to model sump performance, considering the concerns discussed in GL 2004-02 and GSI-191, as a validation of the assumptions made in the licensing basis ECCS evaluation model. Rather, STP proposes a risk-informed approach to validate assumptions in the ECCS evaluation model.

3.1 Requirements that Concern ECCS and CSS Functions for Core Cooling and Containment Heat Removal and Atmosphere Cleanup The exemption request pertains to the following regulatory requirements concerning the ECCS containment sump and associated systems, e.g., the Emergency Core Cooling System (ECCS) and containment spray system (CSS), and GDC that require systems be provided to perform specific functions (i.e., emergency core cooling, containment heat removal, and containment atmosphere cleanup) following a postulated design basis accident (DBA).

NOC-AE-1 3002954 Page 4 of 11

a.

§ 50.46(b)(5), "Long-term cooling," states that after any calculated successful initial operation of the ECCS, the calculated core temperature shall be maintained at an acceptably low value and decay heat shall be removed for the extended period of time required by the long-lived radioactivity remaining in the core."

Impact on § 50.46(b)(5): The exemption request is specific to the requirement for demonstrating long-term core cooling capability as required by § 50.46(b)(5) as it pertains to the validation of assumptions made in the licensing basis analysis, and is not intended to be applicable to the other requirements provided in § 50.46 or Appendix K to Part 50. The exemption request applies to the requirement to use a deterministic method to model sump performance to validate assumptions made in the existing licensing basis.

b.

GDC 35, "Emergency core cooling," states that a system to provide abundant emergency core cooling shall be provided. The system safety function shall be to transfer heat from the reactor core following any loss of reactor coolant at a rate such that (1) fuel and clad damage that could interfere with continued effective core cooling is prevented and (2) clad metal-water reaction is limited to negligible amounts. Suitable redundancy in components and features, and suitable interconnections, leak detection, isolation, and containment capabilities shall be provided to assure that for onsite electric power system operation (assuming offsite power is not available) and for offsite electric power system operation (assuming onsite power is not available) the system safety function can be accomplished, assuming a single failure."

Impact on GDC 35: In conjunction with the existing licensing basis, justification for and approval of the exemption request provides reasonable assurance with high probability that the GDC is met. The exemption request applies to the requirement to use a deterministic method to model sump performance to validate assumptions made in the existing licensing basis

c.

GDC 38, "Containment heat removal," states that a system to remove heat from the reactor containment shall be provided. The system safety function shall be to reduce rapidly, consistent with the functioning of other associated systems, the containment pressure and temperature following any LOCA and maintain them at acceptably low levels. Suitable redundancy in components and features, and suitable interconnections, leak detection, isolation, and containment capabilities shall be provided to assure that for onsite electric power system operation (assuming offsite power is not available) and for offsite electric power system operation (assuming onsite power is not available) the system safety function can be accomplished, assuming a single failure.

Impact on GDC 38: In conjunction with the existing licensing basis, justification for and approval of the exemption request provides reasonable assurance with high probability that the GDC is met. The exemption request applies to the requirement to use a deterministic method to model sump performance to validate assumptions made in the existing licensing basis.

NOC-AE-13002954 Page 5 of 11

d.

GDC 41, "Containment atmosphere cleanup," states that systems to control fission products, hydrogen, oxygen, and other substances which may be released into the reactor containment shall be provided as necessary to reduce, consistent with the functioning of other associated systems, the concentration and quality of fission products released to the environment following postulated accidents, and to control the concentration of hydrogen or oxygen and other substances in the containment atmosphere following postulated accidents to assure that containment integrity is maintained.

Impact on GDC 41: In conjunction with the existing licensing basis, justification for and approval of the exemption request provides reasonable assurance with high probability that the GDC is met. The exemption request applies to the requirement to use a deterministic method to model sump performance to validate assumptions made in the existing licensing basis.

3.2 Requirements that Concern Crediting CSS with Reducing Accident Source Term:

With respect to regulatory requirements and GDC that concern crediting a CSS with reducing the accident source term, the current licensing basis evaluations are described below:

a.

§ 50.67, "Accident source term," states the requirements for acceptable radiation dose for the design basis radiological consequence analyses.

Impact on § 50.67: In conjunction with the existing licensing basis, justification for and approval of the exemption request provides reasonable assurance with high probability that the requirements of § 50.67 are met. The exemption request applies to the requirement to use a deterministic method to model sump performance to validate assumptions made in the existing licensing basis.

b.

GDC 19, "Control Room," states that a control room shall be provided from which actions can be taken to operate the nuclear power unit safely under normal conditions and to maintain it in safe condition under accident conditions, including LOCA. Adequate radiation protection shall be provided to permit access and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 5 rem whole body, or its equivalent to any part of the body, for the duration of the accident. Equipment in appropriate locations outside the control room shall be provided (1) with a design capability for prompt hot shutdown of the reactor, including necessary instrumentation and controls to maintain the unit in a safe condition during hot shutdown, and (2) with a potential capability for subsequent cold shutdown of the reactor through the use of suitable procedures.

c.

Impact on GDC 19: In conjunction with the existing licensing basis, justification for and approval of the exemption request provides reasonable assurance with high probability that the GDC is met. The exemption request applies to the requirement to use a deterministic method to model sump performance to validate assumptions made in the existing licensing basis.

NOC-AE-13002954 Page 6 of 11 3.3 Evaluation of Impact on the Balance of § 50.46 and Appendix K to Part 50:

The exemption request to support closure for GSI-191 is intended to address ECCS acceptance criterion for long-term cooling as presented in § 50.46(b)(5) and is not applicable to the other acceptance criteria of § 50.46 (peak cladding temperature, maximum cladding oxidation, maximum hydrogen generation, and coolable geometry).

For the purposes of demonstrating the balance of the acceptance criteria of § 50.46(b),

the design and licensing basis descriptions of accidents requiring ECCS operation, including analysis methods, assumptions, and results, which are provided in South Texas Project Electric Generating Station (STPEGS) Updated Final Safety Analysis Report (UFSAR) Chapters 6 and 15 remain unchanged. The performance evaluations for accidents requiring ECCS operation described in UFSAR Chapters 6 and 15, based on the Appendix K Large-Break Loss-of-Coolant Accident (LBLOCA) analysis, demonstrate that for breaks up to and including the double-ended severance of a reactor coolant pipe, the ECCS will limit the clad temperature to below the limit specified in

§ 50.46 and assure that the core will remain in place and substantially intact with its essential heat transfer geometry preserved.

The reference to "acceptable evaluation mode!' in § 50.46(d) is discussed in

§ 50.46(a)(1) and defined in § 50.46(c)(2). The purpose of the risk-informed approach is to evaluate the ECCS sump performance to determine if the sumps are configured properly to provide enough flow to ensure criterion § 50.46(b)(5), "Long-term cooling," is met. This is discussed in GL 2004-02, and in SECY-04-0150 and the NRC safety evaluation report on NEI 04-07 which state:

"While not a component of the 10 CFR 50.46 ECCS evaluation model, the calculation of sump performance is necessary to determine if the sump and the residual heat removal system are configured properly to provide enough flow to ensure long-term cooling, which is an acceptance criterion of 10 CFR 50.46.

Therefore, the staff considers the modeling of sump performance as the validation of assumptions made in the ECCS evaluation model. Since the modeling of sump performance is a boundary calculation for the ECCS evaluation model, and acceptable sump performance is necessary for demonstrating long-term core cooling capability (10 CFR 50.46(b)(5)), the requirements of 10 CFR 50.46 are applicable."

The STP risk-informed approach, as described in Enclosures 1 and 4, uses the PRA model to quantify the risk associated with GSI-1 91, thereby quantifying the residual risk from those issues which have not been resolved using other methods, and to show that it meets the acceptance guidelines defined in RG 1.174. By quantifying this risk, the approach validates assumptions in the STP Units 1 and 2 Appendix K Large-Break Loss-of-Coolant Accident analysis associated with the concerns raised by GSI-1 91.

Therefore, the exemption request is specific to the requirement for demonstrating long-term core cooling capability as required by § 50.46(b)(5) as it pertains to the validation of assumptions made in the ECCS evaluation model, and is not intended to be applicable to other requirements provided in § 50.46 or Appendix K to Part 50.

NOC-AE-1 3002954 Page 7 of 11

4.

Technical Justification for the Exemption Regulatory Guide (RG) 1.174 provides technical guidance for licensees who request NRC approval for changes in the licensing basis using a risk-informed approach. This guidance establishes five principles that should be considered for risk-informed changes to the licensing basis. The requested exemption is part of a risk-informed approach that addresses the principles stated in RG 1.174, as described below.

a. Compliance with Current Regulations The proposed change in the licensing basis should meet current regulations, unless it is explicitly related to a requested exemption or rule change.

This exemption request implements this principle.

b.

Defense-in-Depth The proposed change in the licensing basis should be consistent with the defense-in-depth philosophy.

The exemption request is consistent with the defense-in-depth philosophy in that the following aspects of the facility design and operation are unaffected:

Functional requirements and the design configuration of systems Existing plant barriers to the release of fission products Design provisions for redundancy, diversity, and independence Plant's response to transients or other initiating events Preventive and mitigative capabilities of plant design features The STP risk-informed approach analyzes a full spectrum of LOCAs, including double-ended guillotine breaks for all piping sizes up to and including the largest pipe in the reactor coolant system (RCS). By requiring that mitigative capability be maintained in a realistic and risk-informed evaluation of GSI-191 for a full spectrum of LOCAs, the approach ensures that defense-in-depth is maintained. Additional discussion on defense-in-depth is summarized in Enclosure 4.

c. Safety Margins The proposed change in the licensing basis should maintain sufficient safety margins.

The requested exemption does not involve a change in any functional requirements or the configuration of plant structures, systems and components (SSCs). Because of the very small risk associated with the change, STPNOC does not expect the need to change any of the safety analyses in the UFSAR. Therefore, sufficient safety margins associated with the design will be maintained by the exemption.

NOC-AE-13002954 Page 8 of 11

d. Changes in Risk When proposed changes result in an increase in CDF or risk, the increases should be small and consistent with the Commission's Safety Goal Policy Statement.

The proposed change is defined as the risk associated with GSI-191 concerns.

Using engineering analysis and the PRA this risk has been calculated and shown to be in Region III, "Very Small Changes," and is therefore consistent with the Commission's Safety Goal Policy Statement.

e. Monitoring the Impact of the Proposed Change The impact of the proposed change in the licensing basis should be monitored using performance measurement strategies.

The PRA analysis supporting the change is performed using STPNOC PRA procedures as required for PRA analyses and assessments. Code required inspection programs and other requirements provide monitoring capability.

Additional discussion on monitoring the proposed change is summarized in. provides a summary of the STP PRA, risk assessment methodology, and engineering analysis, including modeling of physical plant properties and treatment of uncertainties, and references to other supporting information. The results of the risk-informed approach demonstrate that the calculated risk associated with GS-1 91 concerns for STP Units 1 and 2 meets the acceptance guidelines defined by RG 1.174, and provides a basis for the exemption request:

Change in CDF is - 1.1E-8/yr Change in LERF is - 8.6E-12/yr The STP approach models the physical characteristics of debris generation and transport over a full range of plausible conditions in order to provide inputs to the STP PRA. The PRA is used to calculate the risk (CDF and LERF) associated with GS1-191 for the as-built, as-operating plant, to quantify risk benefit associated with no additional changes to the facility required to address the residual risk associated with GSI-191.

Justification for the exemption, and for closure for GSI-1 91, is based on:

(1)

NRC review and approval of the risk-informed approach, and (2)

The calculated risk associated with GSI-191 meeting the acceptance guidelines in RG 1.174.

5.

Justification for Exemption Pursuant to 10 CFR 50.12(a)(1)

The exemption is authorized by law.

NOC-AE-13002954 Page 9 of 11 The NRC has authority under the Atomic Energy Act of 1954, as amended, to grant exemptions from its regulations if doing so would not violate the requirements of law.

This exemption is authorized by law as is provided by 10 CFR 50.12 which provides the NRC authority to grant exemptions from Part 50 requirements with provision of proper justification. Approval of the exemption would not conflict with any provisions of the Atomic Energy Act of 1954, as amended, the Commission's regulations, or any other law The exemption does not present an undue risk to the public health and safety.

The underlying purpose of § 50.46 is to establish acceptance criteria for ECCS performance, and together with GDC 35, to provide a high confidence that the systems will perform the required functions. The underlying purpose of the other applicable regulatory requirements provided in GDC 38, GDC 41, § 50.46 and GDC 19 also provide a high confidence that the systems will perform the required functions. The proposed exemption does not involve any modifications to the plant that could introduce a new accident precursor or affect the probability of postulated accidents, and therefore the probability of postulated accidents is not increased. The PRA and engineering analysis demonstrate that the calculated risk is small and consistent with the intent of the Commission's Safety Goal Policy Statement, which defines an acceptable level of risk that is a small fraction of other risks to which the public is exposed.

The exemption is consistent with the common defense and security.

The exemption involves a change to the licensing basis for the plant that has no relation to the possession of licensed material or any security requirements that apply to STP Units 1 and 2. Therefore the exemption is consistent with the common defense and security.

6.

Special Circumstances 10 CFR 50.12(a)(2) states that special circumstances are present whenever any of six listed circumstances exist. The following listed circumstances are applicable to this request.

a.

§ 50.10(a)(2)(ii) applies because for the use of the proposed risk-informed approach, application of the regulation would not serve the underlying purpose of the regulatory requirements or is not necessary to achieve the underlying purpose of the regulatory requirements. An objective of the regulatory framework is to maintain low risk to the public health and safety. The supporting analysis demonstrates that the associated risk is consistent with the Commission's Safety Goals for nuclear power plants. Consequently, the special circumstance described in § 50.12(a)(2)(ii) applies.

b.

§ 50.10(a)(2)(iii) applies because compliance with the applicable rules would result in undue hardship or other costs that are significantly in excess of those actions already taken to demonstrate compliance, but without a compensating increase in the level of quality and safety. Absent the exemption, the amount of debris generating contributors in the plant design would need to be reduced. The risk assessment shows that any such modifications to the plant would have a NOC-AE-13002954 Page 10 of 11 very small change in plant risk. The cost and radiological exposure estimates for removal of insulation are significant, approaching tens of millions of dollars and hundreds of person-Rem per unit, depending on the scope of modifications. In comparison, the risk-informed approach is estimated to be a fraction of the cost of the deterministic resolution approach, with no personnel dose consequences based on no projected need for any further modifications needed for closure of GSI-191. Consequently, the special circumstance described in § 50.12(a)(2)(iii) applies.

7.

Conclusion Issuance of an exemption to authorize the use of the risk-informed approach is consistent with the provisions of § 50.12(a)(1) and special circumstances required by

§ 50.12(a)(2) are present as discussed above.

The PRA assessment is used to quantify the risk associated with GSI-191 in order to determine the risk impact on the requirement that long-term cooling will be maintained for § 50.46(b)(5). Based on the determination that the risk meets the acceptance guidelines of RG 1.174, the results demonstrate with reasonable assurance that this requirement and other regulatory requirements that rely on ECCS and CSS functions in the recirculation mode are met.

8.

Implementation To support the completion of work and resolution schedule for closure of GSI-1 91 as described in SECY-12-0093, STPNOC requests that the exemption request be approved for implementation by December 2014.

9.

References

1)

Generic Safety Issue (GSI)-1 91, "Assessment of Debris Accumulation on Pressurized-Water Reactor Sump Performance"

2)

Generic Letter (GL) 2004-02, "Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors," dated September 13, 2004 (ML042360586)

3)

Bulletin 2003-01, "Potential Impact of Debris Blockage on Emergency Sump Recirculation at Pressurized-Water Reactors," dated June 9, 2003 (ML031600259)

4)

SECY-1 0-0113, "Closure Options for Generic Safety Issue - 191, Assessment of Debris Accumulation on Pressurized Water Reactor Sump Performance," dated August 26, 2010 (ML101820212).

5)

Staff Requirements Memorandum (SRM)-SECY-1 0-0113, "Closure Options for Generic Safety Issue [GSI] - 191, Assessment of Debris Accumulation on Pressurized Water Reactor [PWR] Sump Performance," dated December 23, 2010 (ML103570354)

NOC-AE-1 3002954 Page 11 of 11

6)

Letter, J. W. Crenshaw, STPNOC, to NRC Document Control Desk, "Status of the South Texas Project Risk-Informed (RI) Approach to Resolve Generic Safety Issue (GSI)-191," NOC-AE-11002775, dated December 14, 2011 (MLl 1354A386)

7)

Regulatory Guide (RG) 1.174, Revision 2, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis" (ML100910006)

8)

Commission SECY Paper, "Closure Options for Generic Safety Issue - 191, Assessment of Debris Accumulation on Pressurized-Water Reactor Sump Performance," SECY-12-0093, dated July 9, 2012 (ML121320270)

9)

NUREG-2150, "A Proposed Risk Management Regulatory Framework" (Appendix H, Alternative 1), dated April 2012 (ML12109A277)

10)

Letter, T. J. Jordan to Document Control Desk, "Response to a Request for Additional Information Regarding the 60 Day Response to Bulletin 2003-01:

"Potential Impact of Debris Blockage on Emergency Sump Recirculation at Pressurized-Water Reactors (TAC Nos. MB9615 and MB9616)," NOC-AE-05001883, dated July 13, 2005 (ML052000279)

11)

SECY-04-0150, "Alternate Approaches for Resolving the Pressurized Water Reactor Sump Blockage Issue (GSI-191), Including Realistic and Risk-Informed Considerations," dated August 16, 2004

12)

GSI-191 Safety Evaluation Report, Rev. 0, "Evaluation of NEI Guidance on PWR Sump Performance," dated December 6, 2004 (ML043280007)

13)

Staff Requirements Memorandum (SRM)-SECY-1 0-0113, "Closure Options for Generic Safety Issue (GSI) - 191, Assessment of Debris Accumulation on Pressurized Water Reactor [PWR] Sump Performance," dated December 23, 2010 (ML103570354)

14) 51 FR 30028, "Safety Goals for the Operations of Nuclear Power Plants; Policy Statement," Federal Register, Volume 51, p. 30028, August 4, 1986

NOC-AE-13002954 ENCLOSURE 3 Changes to the South Texas Project Units 1 and 2 Licensing Basis (Information Only)

NOC-AE-13002954 Page 1 of 7 Changes to the South Texas Project Units 1 and 2 Licensing Basis (Information Only)

The changes to the STP Updated Final Safety Analysis Report (UFSAR) shown below are provided for information purposes to support NRC review and approval of the risk-informed approach and exemption request. Changes to the UFSAR will be implemented pursuant to NRC approval of the risk-informed approach and the exemption request.

The changes are based on the exemption request (Enclosure 2) submitted in accordance with the provisions of § 50.12 for purposes of resolving Generic Safety Issue (GSI)-1 91, "Assessment of Debris Accumulation on Pressurized-Water Reactor Sump Performance."

The design and licensing basis descriptions of accidents requiring ECCS operation, including analysis methods, assumptions, and results provided in UFSAR Chapters 6 and 15 remain unchanged. The performance evaluations for accidents requiring ECCS operation described in Chapters 6 and 15, based on the South Texas Project Units 1 and 2 Appendix K Large-Break Loss-of-Coolant Accident (LBLOCA) analysis, demonstrate that for breaks up to and including the double-ended severance of a reactor coolant pipe, the ECCS will limit the clad temperature to below the limit specified in § 50.46, and assure that the core will remain in place and substantially intact with its essential heat transfer geometry preserved.

Changes to the UFSAR are shown below in gray highlight.

TABLE 3.12-1 REGULATORY GUIDE MATRIX ABBREVIATIONS:

A Conform to guide No.

Regulatory Guide Title UFSAR Reference Revision Status STPEGS Position On STPEGS 1.82 Sumps for Emergency Core 6.2.2.1.2 Proposed Rev 1 A

Cooling and Containment 6.2.2.2.3 (5/83)

Spray Systems 6.3.4.1 NOTES NOC-AE-1 3002954 Page 2 of 7 6.2.2.1.2 Containment Emergency Sump Design Bases:

The Containment emergency sump meets the following design bases:

1.

Sufficient capacity and redundancy to satisfy the single-failure criteria. To achieve this, each CSS/ECCS train draws water from a separate Containment emergency sump.

2.

Capable of satisfying the flow and net positive suction head (NPSH) requirements of the ECCS and the CSS under the most adverse combination of credible occurrences. This includes minimizing the possibility of vortexing in the sump.

3.

Minimizes entry of high-density particles (specific gravity of 1.05 or more) or floating debris into the sump and recirculating lines.

4.

Sumps are designed in accordance with RG 1 roposedrevision1,My 1983 and with Generic Letter 2004-02 as described in A

6.2.2.2.3 Containment Emergency Sump

Description:

At the beginning of the recirculation phase, the minimum water level above the Containment floor is adequate to provide the required NPSH for the ECCS and CSS pumps. The sumps are designed to RG 1.82, proposed revision 1, May 1983 and to the requirements of Generic Letter 2004-02 as described in

. The sump structures are designed to limit approach flow velocities to less than 0.009 ft/sec permitting high-density particles to settle out on the floor and minimize the possibility of clogging the strainers. The sump structures are designed to withstand the maximum expected differential pressure imposed by the accumulation of debris.

6.2.2.3.5 Pump Net Positive Suction Head Requirements:

The minimum available net positive suction head (NPSH) for the CSS pumps is such that an adequate margin is maintained between the required and the available NPSH for both the injection and recirculation phase, ensuring the proper operation of the CSS a u

ARecirculation operation gives the limiting NPSH requirements for the CSS pumps The Westinghouse CSS pump design provides for the NPSH requirement to be met by the inherent design of the pump. CSS pumps are vertical motor-driven pumps, each sitting in an individual barrel. The design calls for a distance of 15 ft in this barrel between the suction nozzle centerline and the pump first-stage impeller. The 15-ft liquid-head in the pump barrel is thus expected to inherently satisfy the 15-ft NPSH requirement.

The analysis of available NPSH to the CSS pumps concerns itself with the NPSH at the pump suction nozzle, located at the top of the barrel. Since the pump barrels provide the required NOC-AE-1 3002954 Page 3 of 7 NPSH at the first-stage impeller, the piping layout need provide only sufficient NPSH at the pump suction nozzle to prevent flashing in the barrel.

Two modes of operation have been analyzed for the CSS pumps:

1.

Pump taking suction from the RWST and delivering spray to the Containment

2.

Pump taking suction from the Containment sump and delivering spray to the Containment Case 2 represents the "worst case" since it gives the minimum available NPSH.

The assumptions and conservatisms used in the analysis are listed below. No exceptions are taken to RG 1.1.

1.

Containment pressure equals the vapor pressure of the sump water.

2.

The runout flows of each pump are used to account for maximum friction loses.

The minimum flood level in Containment is determined by considering the quantities of water trapped by the refueling cavity.

The results of the analysis show the available NPSH at the first-stage impeller of the CSS pumps to be reater than the required NPSH and show that the fluid at the suction flange is pumpperfrmace i therecrcultionmod There is sufficient NPSH at the suction nozzle to prevent flashing in the barrel, and the analysis meets the guideline of RG 1.1. The NPSH parameters are listed in Table 6.2.2-4.

NPSH for the ECCS pumps is addressed in Section 6.3.

TABLE 6.2.2-4 CSS PUMP NPSH PARAMETERS Required NPSH at Max Flow Rate, ft (max) 16.4 Available NPSH, ft (from RWST) 56.1 (From RCB Emergency Sump)

>17.6 (

NOC-AE-13002954 Page 4 of 7 6.3.2.2 EauiDment and Comronent Descriptions.

Net Positive Suction Head Available and required net positive suction head (NPSH) for ECCS pumps are shown in Table 6.3-1. The safety intent of Regulatory Guide (RG) 1.1 is met by the design of the ECCS such that adequate NPSH is provided to system pumps.

The NPSH available for the injection mode is determined from the elevation head and the vapor pressure (atmospheric) of the water in the RWST, and the pressure drop in the suction piping from the tanks to the pumps. The NPSH evaluation is based on all pumps operating at maximum flow rate with no credit taken for the elevation head in the tank and full penalty assumed for head loss in the suction lines.

In addition to considering the static head and suction line pressure drop, the calculation of available NPSH in the recirculation mode assumes that the vapor pressure of the liquid in the sump is equal to the Containment ambient pressure. This assures that the actual available NPSH is always greater than the calculated NPSH.

TABLE 6.3-1 EMERGENCY CORE COOLING SYSTEM COMPONENT PARAMETERS High Head Safety Injection Pumps Required NPSH at max. flow rate, ft (max)

Available NPSH, ft (From RWST)

(From RCB Emergency Sump) 16.1 55.8

> 17.8(

Low Head Safety Injection Pumps Required NPSH, ft (max)

Available NPSH, ft (From RWST)

(From RCB Emergency Sump) 16.5 55.1

> 18.0o NOC-AE-13002954 Page 5 of 7 The UFSAR change for Appendix 6A shown below consists entirely of new content, therefore gray highlight is not used.

RISK-INFORMED APPROACH TO POTENTIAL IMPACT OF DEBRIS BLOCKAGE ON EMERGENCY RECIRCULATION DURING DESIGN BASIS ACCIDENTS INTRODUCTION AND

SUMMARY

NRC Generic Letter 2004-02 (GL 2004-02) "Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors," required licensees to perform an evaluation of the ECCS and CSS recirculation functions, and the flowpaths necessary to support those functions, based on the potential susceptibility of sump screens to debris blockage during design basis accidents requiring recirculation operation of ECCS or CSS. This Generic Letter resulted from the Generic Safety Issue (GSI) 191, "Assessment of Debris Accumulation on Pressurized-Water Reactor Sump Performance." As a result of the evaluation required by GL 2004-02 and to ensure system function, sump design modifications were implemented (refer to Section 6.2.2.2.3).

GL 2004-02 sump performance evaluation activities, documented in References 6A-1 and 6A-2, included the following:

Containment walkdowns Debris generation and transport analysis Calculation of required and available net positive suction head (NPSH)

Screen requirements Screen structural analyses

° Potential or planned design/operational/procedural modifications Downstream effects Upstream effects Chemical effects The plant hardware modifications and plant administrative procedures and processes implemented in response to issues identified in GL 2004-02 provide high confidence that the sump design supports long-term core cooling following a design basis loss of coolant accident.

The sump design meets the regulatory requirements listed in GL 2004-02 except as authorized by exemption based on an evaluation, documented in Reference 6A-3, that the risk associated with the current as-built, as-operated plant meets the acceptance guidelines in Regulatory Guide 1.174 (Reference 6A-4).

DISCUSSION The plant licensing basis considers long-term core cooling following a LOCA as identified in 10 CFR 50.46. Long-term cooling is supported by the ECCS which includes the Containment Spray (CS), the High Head Safety Injection (HHSI), the Low Head Safety Injection (LHSI), and the Residual Heat Removal (RHR) systems. The licensing basis requires these particular NOC-AE-13002954 Page 6 of 7 systems to operate with high probability following a LOCA. Using a risk-informed approach to address the concerns of GSI-1 91, the probability and uncertainty associated with the operation of the ECCS to maintain long-term cooling following a LOCA has been quantified. The results show that long-term cooling is ensured with high probability since the risk associated with GSI-191 concerns meets the acceptance guidelines for "Very Small Changes" in Region III as defined in Regulatory Guide 1.174.

The method of analysis uses an integrative approach to explicitly provide the probabilities for a few post-LOCA basic events of the STP plant-specific PRA. This has been done by modeling the underlying physical phenomena of the basic events and by propagating uncertainties in the physical models.

In particular, the demand recirculation failure probability is replaced with the following basic events:

Pressure drop due to build up of debris on the sump strainers with chemical effects resulting in loss of NPSH margin for the ECCS pumps Strainer mechanical collapse Air ingestion through the sump strainers Core blockage with chemical effects Boron precipitation in the core The accident sequences were analyzed in a realistic time-dependent manner with uncertainty propagation to determine the probabilities of various failures potentially leading to core damage from a spectrum of location-specific pipe breaks for input into STP's plant-specific probabilistic risk assessment (PRA). The specific failure modes that were considered are:

1.

Strainer head loss exceeds the NPSH margin for the pumps causing some or all of the ECCS and CSS pumps to fail.

2.

Strainer head loss exceeds the strainer structural margin causing the strainer to fail, which could subsequently result in larger quantities and larger sizes of debris being ingested into the ECCS and CSS.

3.

Air intrusion exceeds the limits of the ECCS and CSS pumps causing degraded pump performance or complete failure due to gas binding.

4.

Debris penetration exceeds ex-vessel effects limits causing a variety of potential equipment and component failures due to wear or clogging.

5.

Debris penetration exceeds in-vessel effects limits resulting in partial or full core blockage with insufficient flow to cool the core.

6.

Buildup of oxides, crud, LOCA-generated debris, and chemical precipitates on fuel cladding exceeds the limits for heat transfer resulting in unacceptably high peak cladding temperatures.

7.

Boron concentration in the core exceeds the solubility limit leading to boron precipitation and subsequently resulting in unacceptable flow blockage or impaired heat removal.

Failure Modes 4 and 6 were conservatively addressed as part of the previous deterministic evaluations for STP with no issues of concern and were therefore not explicitly modeled in the PRA analysis. The remaining failure modes were explicitly modeled.

NOC-AE-13002954 Page 7 of 7 Failure probabilities and associated uncertainties determined in the supporting engineering analysis are passed to the plant-wide PRA, which determines the incremental risk associated with GS1-191 failure modes. Three new top events were added to the PRA assessment model to accommodate composite GSI-191 failure processes:

1.

Failure at the sump strainer

2.

Boron precipitation in the core

3.

Blockage of the core The engineering analysis supports the three composite failure probabilities needed for the PRA by testing the outcome of every postulated break scenario against seven performance thresholds:

1.

Strainer DP > NPSH margin

2.

Strainer DP > P-buckle

3.

Strainer F-void > 0.02

4.

Core fiber load > cold leg break fiber limit for boron precipitation

5.

Core fiber load > hot leg break fiber limit for boron precipitation

6.

Core fiber load a cold leg break fiber limit for flow blockage

7.

Core fiber load > hot leg break fiber limit for flow blockage Using the inputs noted above, the PRA assessment model is informed with risk insights for the failure modes associated with GSI-191 concerns. The PRA analysis yields results that meet the acceptance guidelines for Region III, "Very Small Changes," as defined by RG 1.174, i.e., the change in CDF is less than 1 E-6/yr and the change in LERF is less than 1 E-7/yr.

REFERENCES Appendix 6A:

6A-1 Correspondence NOC-AE-05001922, dated August 31, 2005 6A-2 Correspondence NOC-AE-08002372, dated December 11, 2008 6A-3 Correspondence NOC-AE-13002954, dated January 31, 2013 6A-4 Regulatory Guide 1.174, "An Approach For Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis" (May 2011)

NOC-AE-13002954 ENCLOSURE 4 Risk-Informed Closure of GSI-191 Volume 1.0 Project Summary NOC-AE-1 3002954 RISK-INFORMED CLOSURE OF GSI-191 VOLUME 1.0 PROJECT

SUMMARY

January 31, 2013 as"l aP)

DOCUMENT: STP-RIGSI191-VO1 REVISION: 0 PREPARED BY:

Ernie Kee, Supervisor Risk Projects, STPNOC Risk Management REVIEWED BY:

Zahra Mohaghegh, Ph.D, Soteria Consultants, LLC Seyed Reihani, Ph.D., Soteria Consultants, LLC Reza Kazemi, Ph.D., Soteria Consultants, LLC Ali Mosleh, Ph.D., Soteria Consultants, LLC Don Wakefield, ABS Consulting Bruce C. Letellier, Ph.D., Los Alamos National Laboratory Tim Sande, ENERCON Janet Leavitt, Ph.D., University of New Mexico Coley Chappell, STPNOC Licensing Wes Schulz, STPNOC Design Engineering C. Rick Grantom, PE, STPNOC Risk Projects Steve Blossom, STPNOC Project Management NOC-AE-1 3002954 Contents List of Tables ii List of Figures Acknowledgements Executive Summary 1 Proposed Change 1.1 Method of Analysis iii v

vi 1

3 2 Engineering Analysis 2.1 Defense-in-Depth and Safety Margin 2.1.1 D ID 2.1.2 Safety margin................

2.2 Risk Impact......................

2.3 PRA Adequacy...................

2.3.1 Scope of the PRA............

2.3.2 Level of detail...............

2.3.3 Technical adequacy...........

2.3.4 Plant representation..........

2.4 Acceptance Guidelines..............

2.5 Comparison with Guidelines..........

2.5.1 Types of uncertainties and methods 2.5.2 Parameter uncertainty.........

2.5.3 Model uncertainty of analysis ines 3

5 5

7 8

9 10 10 10 11 16 18 19 21 21 22 23 24 2.5.4 Completeness uncertainty 2.5.5 Comparisons with acceptance guidel 2.6 Decision making...................

3 Implementation and Monitoring 4 Proposed Change 5 Quality Assurance 6 Documentation 6.1 Introduction.........

6.2 Archival Documentation...............................

6.3 Submittal Documentation.......

24 25 25 25 25 25 26 7 Independent Technical Oversight 28 NOC-AE-1 3002954 8 Acronyms 30 9 References 34 Appendices 42 A Appendix A. Reg. Guide 1.174 Checklist Al B Appendix B. NEI 04-07 Comparison B1 C Appendix C. Defense-in-Depth and Safety Margin C1 List of Tables 1

Mean LBLOCA conditional failure probabilities for five plant operating states. Failure probabilities shown are for strainer blockage, core fiber load exceeding flow blockage criteria, and sump differential pressure exceeding Ppbi,,kle. Each Case refers to a plant operating state........

13 2

Distribution of total conditional failure for LLOCA under Case 43 (one train operating).........

15 3

All cold leg split fractions conditioned on LOCA categories small, medium, and large for Case 43. The fraction going to the hot leg is simply the complement of the cold leg fraction..................

16 4

Sample attributes of break cases leading to failure for Case 43. In the table: Pipe is a text string defined in the inservice inspection program program; System refers to STP System (all are RCS); Break Size is the size of the break in inches; LOCA size values of 1,2,3 denote small, medium, large LOCA events (all are large); DEGB, YES denotes the fully severed pipe condition (failures dominated by DEGB); RCS Leg denotes break location (CLB or HLB); and Break Location denotes region in the containment building related to debris transport fractions. 17 5

Checklist for Regulatory Guide 1.174......

Al 6

Comparison of NEI 04-07 recommended engineering models with the models implemented in the STPNOC Pilot Project....

BI 7

Historical STP responses related to concerns raised in GSI-191 included actions taken, site-specific design features, procedures, and programs that provide defense-in-depth measures (preventive, mitigative, and protective) and safety margin. References to letters, procedures and other guidance documents are also provided...................

C3 NOC-AE-13002954 List of Figures 1

Reproduction of "Figure 1, Relationship of Regulatory Guide 1.174 to other risk-informed guidance" [38, Figure 1, Page 6] showing the elements used in the Option 2b analysis....................................

viii 2

Illustration of the engineering model input to the PRA used in the Option 2b GSI-191 resolution.........

ix 3

Conceptual illustration of the uncertainty quantification process used to add detail (basic events) to the STPNOC PRA analysis in the LOCA initiating event sequences for the Option 2b analysis.......................

x 4

Illustration of the major elements of the STPNOC quality assurance process for risk-informed closure of GSI-191....................

xii 5

Linear-linear interpolation of bounded Johnson extrema (solid) with nonuniform stratified random break-size profiles (dashed)...........

13 6

Empirical distribution of total failure probability for Case 43 (one train operating) based on five discrete samples of the NUREG 1829 break-frequency uncertainty envelope. Weighted mean = 4.45 x 10-03 marked as bold dot..........

15 7

Reproduction of Figure 4 from Regulatory Guide 1.174, "Acceptance guidelines for core damage frequency", the ACDF, CDF phase plane.

18 8

Reproduction of Figure 5 from Regulatory Guide 1.174, "Acceptance guidelines for large-early-release frequency", the ALERF, LERF phase plane.........

18 iii NOC-AE-1 3002954 Abstract The PRA analyses that provide the technical background in a project to close Generic Safety Issue 191 at the South Texas Project using a risk-informed ap-proach are summarized. The overall methodology used in the PRA analyses is summarized. The elements of Regulatory Guide 1.174 required for a Risk-Infor-med license submittal are documented. Qualitative and quantitative results of the PRA analyses are presented. The results of the Project Technical Oversight activities are summarized.

iv NOC-AE-1 3002954 Acknowledgements The Risk-Informed GSI-191 Closure Pilot Program is an effort piloted by the STP Nuclear Operating Company and jointly funded with several other licensees. It is a collaborative work of teams of experts from industry, academia, and a national laboratory. In general, all products are developed jointly and reviewed in regularly scheduled (monthly) Technical Team Meetings and weekly teleconferences as well as in specific review cycles by Independent Oversight (technical evaluation of all materials), STP Nuclear Operating Company project management, and STP Nuclear Operating Company quality management. The business entities, the main areas of investigation, and the principal investigators of the Pilot Program are summarized below.

STP Nuclear Operating Company Project Management, Licensing, Quality Assurance Steve Blossom; Rick Grantom; Ernie Kee; Jamie Paul; Wes Schulz Alion Science and Technology GSI-191 Analysis & Methodology Implementation (GAMI)

Tim Sande (Enercon); Gil Zigler (Enercon); Austin Glover, Clint Shaffer, Joe Tezak (Enercon)

The University of New Mexico Corrosion/Head Loss Experiments (CHLE)

Kerry Howe, PhD; Janet Leavitt, PhD Los Alamos National Laboratory Containment Accident Stochastic Analysis (CASA) Grande Bruce Letellier, PhD; Gowri Srinivassan, PhD Soteria Consultants, LLC Independent Oversight Zahra Mohagheghl, PhD; Seyed Reihani 2, PhD Texas A&M University Thermal Hydraulics (TH)

Yassin Hassan, PhD; Rodolfo Vaghetto; Saya Lee The University of Texas at Austin Uncertainty Quantification (UQ), Jet Formation Elmira Popova, PhD (1962-2012); David Morton, PhD; Alex Galenko, PhD; Jeremy Tejada, PhD; Erich Schneider, PhD ABS Consulting Probabilistic Risk Assessment (PRA)

David Johnson, ScD; Don Wakefield KNF Consulting Services, LLC Location-Specific Failure Damage Mechanism (DM)

Karl Fleming; Bengt Lydell (ScandPower) 1From January 2013, Assistant Professor in Nuclear Engineering Department at the University of Illinois at Urbana Champaign 2From January 2013, Research Scientist in Nuclear Engineering Department at the University of Illinois at Urbana Champaign v

NOC-AE-13002954 Executive Summary The main objective of the STPNOC Risk-Informed GSI-191 Closure Pilot Project [10, 39]

is, "Through a risk-informed approach, establish a technical basis that would demonstrate that the STP as-built, as-operated plant design is sufficient to gain NRC approval to close the issues raised in GSI-191 by the end of 2013." In 2012, the STP approach has been referred to as Option 2b in the industry.

The results presented in this summary are the joint work of STPNOC Risk Management, Los Alamos National Laboratory, The University of Texas at Austin, Texas A&M University, Alion Science and Technology, ABS Consulting, The University of New Mexico, Soteria Consulting, and KNF Consulting, LLC. STPNOC has also collaborated with the PWROG and NEI in development of the Pilot Project.

In the risk-informed approach, STPNOC will seek NRC approval for closure of GSI-191 since the associated risk in STP's containment buildings is very small. STP is committed to investigating plant modifications including insulation removal and other measures (such as selective insulation reinforcement or debris transport mitigation) to preserve sufficient margins for nuclear safety if the risk analysis indicates risk is more than very small.

The project is based on a two-phase approach that addresses all the concerns related to GSI-191. For the initial phase, in 2011, a quantification was performed to understand if a risk-informed approach would be feasible [5]. Since it was shown to be feasible, the project proceeded to a licensing action in 2012 and 2013.

In both the initial risk analysis in 2011 and the 2012 final quantification, the risk was analyzed to be very small. That is, the change in risk was shown to be less than 1 x 10-6 in core damage frequency and less than lx 10-7 for large-early release frequency. Although previous realistic testing [7] had shown that chemicals were unlikely to affect the head loss in STP debris beds (sump strainers and fuel assemblies), conservative head loss estimates due to the presence of chemical products were assumed for the initial phase. In 2012, ex-perimental data, specific to the STP units, continued to demonstrate chemical effects are not likely to cause large increases in head loss in STP prototypical post-LOCA environ-ments. Nevertheless, conservative estimates of chemical effects were included in the 2012 quantification.

The methodologies and results from the first phase were presented in the following doc-uments: analysis of results from the physical process solver, uncertainty quantification and RELAP5 thermal-hydraulic analyses [24]; LOCA Frequency analysis [13]; Uncertainty quan-tification methodologies and illustrative examples [41]; Jet formation research [47]; and Chemical effects research and experimental design [46].

For the second phase, the results of the 2012 quantification are documented in this report (Project Summary) and the references cited. This information is provided as the technical basis for the NRC review of the Pilot Project.

vi NOC-AE-1 3002954 Introduction & Background The purpose of this document is to summarize the PRA3 quantification supporting the STPNOC 4 license submittal to resolve concerns raised in GSI-191 5 "Assessment of Debris Accumulation on PWR6 Sump Performance" at the STP 7 plants. GSI-191 describes the NRC concerns with potential blockage of the ECCS8. Over several years of study, the scope of concern has come to include the possibility of effects in the RCS9 including core blockage from debris and in 2012, linkage to boric acid precipitation in the core. All GSI-191 concerns are related to the LOCA 10 in high energy (Class 1) piping that would result in the release of fibrous material and other potential debris to the ECCS Emergency Sump.

The purpose of the PRA quantification is to understand the risk and uncertainty in the as-built, as-operated plant associated with having fibrous insulation and latent debris in the STP containment buildings. In particular, the PRA quantification forms the basis for what has come to be referred to as Option 2b, "Mitigative Measures and Alternative Methods Approach", identified as a GSI-191 closure path by the NRC Staff in 2012 [39]. The basic elements of the Option 2b submittal are shown in Figure 1, reproduced from RG1.174"

[38].

The PRA licensing elements addressed in the analysis are highlighted in Figure 1.

STPNOC operates two identical four-loop Westinghouse-designed NSSSs. Each NSSS12 operated by STPNOC is licensed for 3853 MWth. The NSSS is contained in, and protected by, a large dry containment building with approximately 3,410,000 ft 3 of free volume. The primary elements of the ECCS are the HHSI13, LHSI14, CS15, and RCFC16. The three trains mentioned in the descriptions for the HHSI, LHSI, CS, and RCFC systems are completely independent and piped into a single RCS loop. In addition, the HHSI and LHSI can be independently directed to their respective hot leg at their full (run out) flow rate.

Early in 2011, STPNOC began a project to develop risk-informed closure strategies that would meet the intent of the NRC Commission memorandum promulgated by Vietti-Cook in late 2010, while preparing a site-specific licensing submittal. Several public meetings were conducted to inform the NRC staff of the modeling approach and to solicit feedback on the applicability and use of the approach for resolving GSI-191. These meetings included supporting material so that members of the public, and especially other plants, could be informed as well: [45], [66], [51, 49, 50, 52, 53, 54, 55, 57, 58, 59, 60, 61, 62, 63, 9, 56].

In the meetings referred to above, STPNOC described the additional physical models and necessary experimental studies required to support enhancement of the PRA to include 3Probabilistic Risk Assessment 4The STP Nuclear Operating Company 5Generic Safety Issue 191 6Pressurized Water Reactor.

7South Texas Project electric generating station

'Emergency Core Cooling System 9Reactor Coolant System

"°Loss of Coolant Accident "Regulatory Guide 1.174 12Nuclear Steam Supply System 13High Head Safety Injection 14Low Head Safety Injection "Containment Spray System "The Reactor Containment Fan Coolers vii NOC-AE-1 3002954 u~vie S06 SO.Btc)

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• 0 Guidance INational PRIAConrensusI Standards and Industry Reaed Guidance Figure 1: Reproduction of "Figure 1, Relationship of Regulatory Guide 1.174 to other risk-informed guidance" [38, Figure 1, Page 6] showing the elements used in the Option 2b analysis.

the phenomena associated with concerns raised in GSI-191. The overall approach that was adopted caused minimal impact to the PRA Model of Record [44].

The method of analysis uses an integrative approach to explicitly provide failure probabil-ities for a few post-LOCA basic events of the STPNOC plant-specific PRA (that is, Module 1 of Figure 2). These basic event probabilities are estimated in a separate module (that is, Module 2 of Figure 2) by modeling the underlying physical phenomena of the basic events and by propagating the uncertainties in the physical models. The analysis framework shown in Module 2 is called CASA Grande17 and is explained in detail in Volume 3. The added basic event probabilities are shown as the dotted lines going from Module 2 to Module 1 in Figure 2. A conceptual outline of the uncertainty quantification process used in Module 2 of Figure 2 is illustrated in Figure 3. More details regarding the uncertainty quantification are available in Volume 3.

The added basic events which are related to the recirculation phase of LOCA and shown as the dotted lines coming from the engineering models in Figure 2 are solved outside the PRA in an uncertainty quantification framework. An illustration showing the typical process of uncertainty quantification is shown in Figure 3. As shown, the process models distributions developed in different contexts such as, data measurement analysis and expert judgment.

One challenge that could make sampling and uncertainty propagation difficult would be the potential for fitting and estimating multivariate distributions. In the STPNOC risk-informed methodology, multivariate distributions have been avoided by assuming independence be-tween parameters, where possible, and by enforcing explicit conditional dependencies, where "7Containment Accident Stochastic Analysis (CASA) and Grande referring to the STPNOC large, dry containment viii NOC-AE-1 3002954 MODULE 1 STPNOC PRA with added features to capture details of concerns associated with GSI-191 Sump failure with added possibility to violate NPSHR and mechanical collapse Basic events to add ECCS pump failure due to air ingestion Figure 2: Illustration of the engineering model input to the PRA used in the Option 2b GSI-191 resolution.

appropriate.

In some cases, the distributions needed for the PRA involve relatively broad distributions which need to be carefully sampled so that the "tails" are properly accounted for. In general, NLHS18 strategies have been developed to properly represent distributions with long tails, especially in LOCA frequencies.

A quality assurance plan was developed to include standard STPNOC practice for PRA assessments. Over the nearly two-year project duration, (nominally weekly) technical review teleconferences were conducted and supplemented at critical product development steps with on-site reviews. In addition, monthly face-to-face Technical Team meetings were held in 2012.

In general, the STP PRA analyst (STP Technical Team Lead) is responsible for review and verification of the PRA inputs developed. The STP PRA analyst review is further sup-l"Nonuniform Latin Hypercube Sampling ix NOC-AE-1 3002954 Uncertainty Modeling Expert Elicitation Fitting Distributions Fitting Multivariate of to Data Distributions Input Parameters Sampling From the

/

oneCro Other Sampling Sampling Distributionso Monte Carlo Ohrapig ns Schemes

'Extreme Events' Input Parameters*

C Computer Model Uncertainty Time Dependency Estimating Time epenency Non-standard~and/or Unknown Functions J Output Analysis Fitting Estimation of Estimating Distributions Different Oupur Multivariate Characteristics Distributions Flow of information Methodologies Challenges Figure 3: Conceptual illustration of the uncertainty quantification process used to add detail (basic events) to the STPNOC PRA analysis in the LOCA initiating event sequences for the Option 2b analysis.

plemented by independent critical peer review intended to help disclose any overlooked technical gaps that would compromise results and, although the analysis is developed for the industrial setting, also help ensure that the overall product is academically defensible.

Independent Technical Oversight also helped to further focus the analysis efforts.

The overall quality assurance plan is illustrated in Figure 4 as a flow chart. Due to the diverse technology required to be implemented in the GSI-191 scope, the PRA inputs originate with products developed by experts in their respective fields. The CASA Grande integrating framework uses the inputs to generate the two main inputs to the PRA, the sump demand failure likelihood and the in-vessel cooling failure likelihood (for each category of LOCA and all possible equipment configurations). These elements are documented by the vendor and the normal STP vendor document review process is followed to assure they are suitable for use as input to the PRA. The overall STPNOC Pilot Project 19 quality assurance methodology is expected to be similar to most utilities' processes for PRA applications and is consistent with industry PRA standards, practices and procedures [see 3].

The technical and RG1.174 documentation that establishes a technical basis to close GSI-191 in an Option 2b approach consists of several volumes:

" Volume 1, Summary (this volume);

" Volume 2, The PRA analysis and quantification; 19STPNOC Risk-Informed GSI-191 Closure Pilot Project.

X NOC-AE-1 3002954

" Volume 3, The engineering analysis supporting the added basic events and top events needed by the PRA to address the concerns raised in GSI-191;

• Volume 4, Quality Assurance documentation, approach, and summary;

" Volume 5 Oversight (four Volumes, Volume 5.1, 5.2, 5.3, and 5.4); and

" Volume 6 Comment and Request for Additional Information Resolution.

Additional documentation (for example the PRA Model Revision 7 and support calculations) are also available through reference.

The remainder of this document is developed to reflect the RG1.174 sectioning. That is, starting with Section 1, (Proposed Change) through Section 6, (Documentation), the section numbering corresponds exactly to the RG1.174 numbering. A summary of the STPNOC Pilot Project Oversight activity is given in Section 7, Independent Technical Oversight. There are many acronyms used throughout the text. For most of the acronyms used in this document, in addition to providing the complete name for them as a footnote when first used, Section 8 provides the complete name again with a short description for each.

As mentioned earlier, the first numbered sections, 1 through 6, correspond to the same sections in RG1.174. A checklist (Appendix A, Appendix A. Reg. Guide 1.174 Checklist) is provided as an additional resource for cross referencing RG1.174 items with the text in this document. Appendix B is provided to give an overview of the models implemented in the STPNOC Pilot Project and how they correspond to those recommended in NEI 04-07 [34].

Finally, Appendix C is a summary of the many historical (that is, prior to the STPNOC Pilot Project) STPNOC actions that have been put in place that address the concerns raised in GSI-191 over the several years leading up to the STPNOC Pilot Project.

xi NOC-AE-1 3002954 Responsiblifty. Contracted service organizatton Peocosau Local quality program STPProcedure: OPGP03-ZT-01 38 Contractor/ Staff Augmentation Volunteer Training and.Qualification Program Input develop 0

CASA Frame*

0 Input to PRA pment 1

Vi

'Responsibifity: LANL orkPoces Local Quality Program, Allon Science VerificationNalildation Program 0.

Responsibility-. STP Contract Technical Coordinator, Project Technical Lead Process STP Technical Document Review Process.

Procedure OPGP04-ZA-0328"EngineerIng Document Processing' Interrnal review supplemented and supported by Independent Oversightl Soteria Consultants I

Inputs to PRA.Verified/Reviewed I

PRA Quantification/Output Responsjbildty ABS Consulting Procesm: RISKMAN'quantificasion, STPNOC PRA current plant model STP PRA Anlyses/Assessment Procedure I

Procedure OPGPOS-ZE-OOOI"PRA Analyses/Assessments" I

PRA Application Responslbiity STP Contract Technical Coordinator, Project Technical Lead 1 Process: STP PRA Assessment Process Procedure OPGPO4-ZA-0604'Probabilisitc Risk Assessment Program" Responsibility-. STI Licensing Engineer Proceas: STP Ucense Amendment Process Procedure OPGPOS-ZN-O004"Changes to Licensing Basis Documents and Amendments to the Operating Ucense" License Amendment Figure 4: Illustration of the major elements of the STPNOC quality assurance process for risk-informed closure of GSI-191.

xii NOC-AE-1 3002954 1 PROPOSED CHANGE 1

Proposed Change Part of the STPNOC plant licensing ba-sis change considers long-term core cooling as identified in 10 CFR §50.46 following a LOCA. Long-term cooling is supported by the ECCS which system includes the safety-related CS, the HHSI, LHSI, and the RHR 20 system. The STPNOC licensing basis requires these particular systems to operate with high probability following a LOCA. In addition, the licensing basis requires evaluation of un-certainty associated with proper operation.

[38]

In this licensing basis change, STPNOC ex-plicitly quantifies the probability and uncer-tainty associated with the operation of the ECCS following a LOCA and shows that long-term cooling is ensured with high probability.

In the current license basis neither the prob-ability nor the uncertainty that long-term cooling will operate properly following LOCA is quantified. Therefore, the licensing basis change is to incorporate the probability and uncertainty associated with long-term cool-ing success of the as-built, as-operated plant (as required in the license basis change). This requires NRC approval where the cumulative risk is shown to be very small [38, Figures 4 and 5, Page 16].

History of Defense in Depth and Safety Margin Activities Since the inception of the GSI-191 issue, STP has made significant improvements to pro-cesses, programs, design, and operation that, in the unlikely event of a LLOCA2 1, would mitigate potential consequences. These im-provements include design modifications to the plant hardware, operator training, and procedures. Appendix C is provided to help review the current status and review what is 20Residual Heat Removal System 21Large Break Loss of Coolant Accident in place to address the concerns raised in GSI-191 at the start of the STPNOC Pilot Project.

In the following section, the primary activ-ities from that history that are already in place are summarized.

Procedures and Activities in the Licensing Basis Before the STPNOC Pilot Project started, STPNOC had already taken steps in STP design and operation to help eliminate, or greatly reduce, effects from the concerns raised in GSI-191 on long term cooling at STP. Some of the steps taken include:

" installing very large, uniform-loading ECCS strainers with approximately a factor of ten increased strainer flow area over the strainers originally installed;

• modifying the STP Emergency Operat-ing Procedures to terminate Contain-ment Spray early as a conditional action step as a means to conserve RvWST 22 in-ventory;

• removing effectively all Marinite (Cal-cium Silicate) insulation from the con-tainment building;

" reworking or replacing PWSCC2 3-susceptible welds in the Steam Generators and the Pressurizer safe ends; and

" performing a

comprehensive post-maintenance containment cleanup and inspection following refueling outages to help ensure material that would cause strainer blockage is removed.

The following primary procedures and ac-tivities are implemented that directly or indi-rectly bear on mitigating or eliminating the concerns raised in GSI-191:

22Refueling Water Storage Tank 2 3primary Water Stress Corrosion Cracking 1

NOC-AE-1 3002954 1 PROPOSED CHANGE

" "Condition Reporting Process", STP-NOC plant procedure, OPGP03-ZX-0002: The STPNOC process used to identify plant Management, Opera-tions, and Work Control of any deficien-cies or issues that may arise. This pro-cess requires identification and evalua-tion of the severity and required actions, to be taken as necessary for safe opera-tion.

" "PRA Analyses/Assessments",

STP-NOC plant procedure, OPGP05-ZE-0001: The STPNOC process used in PRA as the basis for applications and risk-based decision making.

" "Design Change Package".

STPNOC plant procedure, OPGP04-ZE-0309: The STPNOC engineering design change pro-cess governing all design changes. Sec-tion 4 of the design change checklist and the supporting descriptions specif-ically address maintaining the assump-tions used for the engineering models in the STPNOC Pilot Project containment analysis.

" "Inspection of Containment Emergency Sumps and Strainers Unit #1 1-A, 1-B, 1-C Unit #2 2-A, 2-B, 2-C", STP-NOC plant procedure, OPSP04-XC-0001: The procedure satisfying Techni-cal Specifications for ECCS sump oper-ability. The specific procedure purpose is to provide instructions for cleanliness and structural inspection of Contain-ment Emergency sumps and strainers 1-A, 1-B, 1-C or 2-A, 2-B, 2-C required by Technical Specifications 4.5.2.d and 4.5.3.1.1.

" "Initial Containment Inspection to Es-tablish Integrity", STPNOC plant pro-cedure, OPSP03-XC-0002: The STPNOC process that ensures no loose debris which could be transported to the Con-tainment Sump and cause restriction of pumps' suctions during LOCA con-ditions is present and is the proce-dure that satisfies Technical Specifica-tions 4.5.2.c.1, 4.6.1.7.1, 4.6.1.7.4, and 3.6.1.7.b.

"ASME Section XI Inservice In-spection",

STPNOC plant procedure, OPSP11-RC-0015:

This procedure ensures that the following requirements of Technical Specifications 4.0.5 /4.4.10 have been satisfied: completion of the inservice inspection (ISI) examina-tions of STP piping and component welds in accordance with the schedule requirements of the ASME Boiler and Pressure Vessel Code,Section XI (2004 Edition No Addenda); Inservice Service Inspections of STP piping and equipment; component supports (excluding snubber assemblies [pin-to-pin]) in accordance with the schedule requirements of the Code; completion of the Inservice Service Inspections of the STP containment metal liner in accordance with the schedule re-quirements of the ASME Boiler and Pressure Vessel Code; completion of the examinations of the STP Reactor Coolant Pump flywheels in accordance with the requirements of Regulatory Guide 1.14.

• "Transient Cycle Counting Limits",

STPNOC plant procedure, OPEP02-ZE-0001: The STPNOC process that pro-vides for the monitoring of the num-ber of primary and secondary plant op-erations that are explicitly considered as design transients for the NSSS pri-mary system and components. This pro-cedure includes the transients listed un-der the normal, upset and test condi-tions in UFSAR Section 3.9, with the 2

NOC-AE-13002954 2 ENGINEERING ANALYSIS exception of particular transients dis-cussed in Step 1.2 of the procedure. This procedure is based on the recommenda-tions of WCAP-12276.

"Shielding" STPNOC plant procedure, OPRP07-ZR-0004: The STPNOC pro-cess for a consistent method of deter-mining the need for, requesting, eval-uating, installing, modifying, account-ing for and removing shielding at STP.

In particular, OPRP07-ZR-0004 requires inspection for signs of wear such as cracking of the blanket material, dam-aged or corroded grommets, or other signs of physical damage. The inspec-tion is performed prior to each removal and storage and thereby minimizes the possibility that transient lead can be in-troduced in the post-LOCA sump chem-istry.

1.1 Method of Analysis The method of analysis uses an integrative approach to explicitly provide the probabili-ties for a few post-LOCA basic events of the STPNOC plant-specific PRA. This has been done by modeling the underlying physical phenomena of the basic events and by prop-agating uncertainties in the physical models.

In particular, the simplistic demand recircu-lation failure probability is replaced with the following basic events:

" Air ingestion through the sump screen;

" Pressure drop due to buildup of de-bris on the sump screens with chemical effects; resulting in NPSHA24 dropping below NPSHR 25 for the ECCS pumps;

• Boron precipitation;

• Core blockage with chemical effects; and 24Net Positive Suction Head Available 2 5Net Positive Suction Head Required

• Strainer mechanical collapse.

In order to assess the potential risk to long-term core cooling due to the issues raised in GSI-191, a theoretical "perfect plant" is hypothesized. The theoretically perfect plant would not be subject to the pos-tulated failure mechanisms that motivated GSI-191, and the as-built, as-operated plant nor the theoretically perfect plant would have any changes in commitments to cur-rent long-term cooling requirements or per-formance of the ECCS.

By adopting an approach that explicitly assesses the potential risk of the issues raised in GSI-191, STPNOC would avoid signifi-cant cost and worker radiation exposure that would be incurred if using an approach that would bound the risk using extreme assump-tions in engineering models of the LOCA, the so-called "deterministic approach" as long as the risk is evaluated to be very small. Cost estimates for the two STPNOC are in the range of $50,000,000 to $60,000,000, consis-tent with other estimates in the industry. Ra-diation exposure to workers is also very high, 10OREM to 200REM.

2 Engineering Analysis Title 10 "Energy" of the Code of Federal Regulations (CFR) is the law that applies to all domestic commercial nuclear power sta-tions. One of the several legal requirements defined in 10CFR§50.46, "Acceptance crite-ria for ECCS for light-water nuclear power reactors" is that events leading to a loss of long-term core cooling must be mitigated with high probability. The main purpose of the ECCS is to mitigate hypothesized LOCA events by supplying cooling water to the reactor. LOCA events can be triggered by a valve failure or a structural failure and the ECCS is designed to mitigate the "worst case" of these failures with high probability.

3 NOC-AE-1 3002954 2 ENGINEERING ANALYSIS Since 2001, GSI-191 has eluded resolution despite significant efforts by industry and the NRC. Although recent thought has been given to risk quantification [65, for exam-ple] and early recognition of the need for risk evaluation was identified, [8, for exam-ple], until now serious investigation into risk quantification has not been undertaken un-til now. Instead, resolution has followed a classical deterministic approach. STPNOC's view, following an initial quantification [5], is that a risk-informed resolution path should be pursued in preference to a determinis-tic approach, thereby quantifying the safety margins and identifying any scenarios that pose significant risk in GSI-191.

The STPNOC PWR RCS operates at tem-peratures higher than about 650'F. As a con-sequence, it is important to use high effi-ciency insulation to prevent exceeding local and general environmental temperatures in the enclosed space of the reactor contain-ment building. NUKON fiberglass insulation is specified for most high-temperature Class I piping and components in the STP contain-ment buildings. The Reactor Vessel and Re-actor Vessel Head are notable exceptions in-sulated with RM12 6.

In addition to the containment building application, insulation similar to NUKON is installed in high temperature steam cycle ap-plications, piping, heaters, valves, etc. Be-cause it is in general usage, STPNOC has a great deal of experience installing and remov-ing fiberglass insulation. Processes and pro-cedures have been in place for many years and, as a result, the plant staff has significant experience with fiberglass insulation leading to maintenance efficiencies.

In the unlikely event of a LOCA, fibrous insulation ablated from piping and compo-nents insulated with NUKON, paint chips dislodged from painted surfaces, latent debris from inefficient containment building house-26Reflective Metal Insulation keeping, ablated concrete, and chemical pre-cipitants, are examples of material that, dur-ing the recirculation phase of a hypothesized LOCA, may cause high differential pressure on the ECCS strainers or reactor core fuel assemblies if they are transported to the containment emergency sump and then to the ECCS filter screens. If the conditions as-sumed in some of the more extreme hypothe-sized cases were realized, the resulting ECCS filter screen differential pressure could be suf-ficient to cause core damage due to the loss of one or more trains of the ECCS. Filter ineffi-ciency may lead to blockage of all the fuel as-semblies which also may result in core dam-age. In addition to the concerns associated with differential pressures mentioned, the is-sue of boron precipitation causing reduced heat transfer in the core has been raised.

In the GSI-191 analysis, the STPNOC PRA shows the risk to core damage or large early release due to the concerns raised in GSI-191 in the as-built, as-operated design is very small. In the analysis, the risk of core dam-age and/or large early release is quantified for a hypothetical plant designed and oper-ated in the same manner as the STP plants except that it is not subject to the concerns raised in GSI-191. The STPNOC PRA meets the ASME/ANS PRA Standard as Capabil-ity Category II and has successfully provided the technical basis for several risk-informed applications at STPNOC RMTS [77, 12]. PRA is relied upon in this analysis to quantify the risk associated with the concerns raised in GSI-191.

The engineering analysis and experimen-tal support for the proposed license basis change are both detailed and broad in scope, commensurate with the perceived complex-ity of the issues raised in GSI-191. The inher-ent uncertainty of the analysis is addressed through the sampling methodology in the uncertainty quantification and by adopting maximum or reasonably high bounds where 4

NOC-AE-1 3002954 2 ENGINEERING ANALYSIS 2.1 Defense-in-Depth and Safety Margin the analyses or experimental data are incom-plete. For example, NLHS is used in the un-certainty propagation methodology to em-phasize random samples from the extreme tails of many uncertain parameters. In par-ticular, when defining random break scenar-ios, the methodology ensures that DEGB27 conditions are included for every weld in the containment within the spectrum of random break sizes that are chosen. NLHS permits a more precise quantification of variability near the extreme conditions for the same number of random scenarios without bias-ing the propagation of uncertainty. Tradi-tional engineering limits are used for equip-ment performance assessment. Examples are:

NPSH for ECCS pumps, air entrainment in the ECCS supply lines, and cooling flow that is required to remove decay heat.

The findings of this analysis indicate that the risk associated with the issues raised in GSI-191 is very small and well within the Commission's safety goal. There are several reasons, many associated with a realistic an-alytical approach, that contribute to a mini-mal risk result. However, it is most important to note that, following the timeline of the is-sues motivating GSI-191, STPNOC took sev-eral steps in the design of the ECCS, contain-ment maintenance, operation of the CS sys-tem, and insulation design that significantly increased the safety margin against the issues that were raised in GSI-191. The most signiif-icant change in design was the introduction of very large ECCS sump strainers that, un-der realistic assumptions of LOCA behavior, shows it is essentially impossible for NPSHA to drop below NPSHR for the ECCS pumps.

Some insulation types have shown in-creased head loss in fiber debris beds. STP-NOC took steps to remove, or ensure that they were not installed, effectively all insula-tion (such as Microtherm and Calcium Sili-cate) that could contribute to increased head 27Double-Ended Guillotine Break loss in fiber debris beds. To prevent introduc-tion of a direct debris path due to strainer damage, the exposed strainer modules have an added protective fence. Taking these steps after the concerns were originally raised in GSI-191 and within the context of continu-ous performance improvement, has greatly improved the safety margin and assurance of DID 28 in the as-built, as-operated STP plants.

2.1 Defense-in-Depth and Safety Margin No changes are proposed to DID or safety margin by this licensing basis change. In-stead, the risk associated with the traditional design basis accident analysis is assessed and quantified. In keeping with the Commission's goal to increase the use of risk analysis in regulation, this analysis quantifies the risk and uncertainty incorporating the impact of steps taken to preserve high levels of nuclear safety against perceived risks, while balanc-ing regulatory cost and the need for signif-icant worker exposure to mitigate concerns where the risk to nuclear safety is significant.

2.1.1 Defense-in-Depth The risk to reliable operation of the as-built, as-operated plant DID systems is analyzed to be very small. STP has three trains of safety injection and three trains of containment fan coolers. The containment fan coolers do not rely on the recirculation mode for cooling the sump water. Decay heat can also be removed by the steam generators using the auxiliary feed water system and the steam generator Power Operated Relief Valves.

The normal charging system is an alter-nate flow path that can be aligned to the RWST if the ECCS pumps become unavail-able for any reason. An entire volume of 28Defense-in-Depth 5

NOC-AE-1 3002954 2 ENGINEERING ANALYSIS 2.1 Defense-in-Depth and Safety Margin the RWST (approximately 500,000 gallons) can be refilled and injected into the contain-ment per design. Normally, STP can refill the RWST in less than a day. When indicated by the Emergency Operating Procedures, the Reactor Coolant Pumps can be operated to cool the core and prevent core damage.

The risk associated with the concerns raised in GSI-191 with the as-built, as-operated plant to the likelihood for radiation release as evaluated by LERF29 is effectively zero. The concerns raised in GSI-191 have no bearing on containment integrity or on the release of radiation.

2.1.1.1 General design criteria.

Because the analysis evaluates the risk of the as-built, as-operated plant, the tradi-tional engineering analysis that forms the basis for the design remains intact and is inherently included in the analysis. That is, the design criteria ultimately result in certain performance standards for the ECCS, such as required flow rates, support system availability, equipment failure combinations, etc. All commitments to design criteria re-main intact, however, they cannot guarantee that core damage or large early release are prevented for every postulated scenario.

Therefore, as previously mentioned, the licensing basis change evaluates the signifi-cance of the (non-zero) risk associated with the as-built, as-operated plant. Because the design criteria are robust and because changes to the design have been made to address specific GSI-191 concerns, the risk has been analyzed and is very small. The analysis incorporates extreme effects of chemical phenomena on debris bed differen-tial pressures as well as boron precipitation.

Even with these extreme assumptions, the probability for core damage is found to be very small and there is no effect on large early release.

29Large Early Release Frequency 2.1.1.2 Defense-in-Depth prin-ciple. The analysis shows that DID is maintained with high probability.

The availability and reliability of the systems supporting DID continue to be assured with high probability with consideration of uncertainty. The analysis shows there is practically no risk to containment integrity associated with the concerns raised in GSI-191 and therefore, the license basis change would indicate that as-built, as-operated containment design remains adequate to prevent a significant release into the en-vironment. In quantification of the risk, no credit is taken for additional operator actions or programatic activities beyond the existing as-built, as-operated plant.

2.1.1.3 Uncertainties of chemical effects.

As part of the analysis, experi-ments have been developed to investigate the significance of the concerns raised in GSI-191 for post-LOCA environments specific to the STP plants. The experiments performed examined conditions under which specific forms of chemical precipitates, particularly A1OOH, can be formed: in-situ over short time frames (on the order of hours and days) by, for example, direct injection of aluminum salts; ex-situ (as in surrogate preparations developed elsewhere in the industry); as well as those formed by actual corrosion sources (such as aluminum, zinc, concrete, etc.) in prototypical post-LOCA environments.

Experiments have shown that using ex-situ methods of precipitate formation pro-duced precipitate forms that are much more likely to result in head loss impacts in debris beds than those formed in-situ. Finally, and consistent with previous observations [7], the more recent experimental work performed for this analysis provides evidence that the chemical corrosion process that would take place in an actual post-LOCA environment is significantly more benign to debris bed 6

NOC-AE-1 3002954 2 ENGINEERING ANALYSIS 2.1 Defense-in-Depth and Safety Margin head loss than any of the surrogate (in-situ or ex-situ) methods. The results of the chem-ical effects experimental program that are most similar to the actual post-LOCA sump conditions give confidence that experiments performed with surrogate preparations rep-resent an upper bound for chemical effects on debris bed head loss.

2.1.1.4 Uncertainties of head loss.

The head loss associated with debris beds can be shown to be dependent on not only chemicals, but also on the presence of par-ticulates transported to the sump area. Such particulates have been hypothesized to re-sult from failure of coatings unqualified for high radiation and post-LOCA fluid chem-istry. The transport and failure extent of such particulates have been conservatively estimated in the STPNOC Pilot Project analy-sis so as to preserve their effect on the result.

The failure extent and rate of failure used in the STPNOC Pilot Project is supported by ex-perimental evidence.

Experiments have been conducted in a high temperature vertical loop using ex-pected post-LOCA fluid conditions (pH, boron and buffer chemical concentrations, and temperature) to examine the uncertainty of coefficients derived in correlations com-monly used in the analysis of head loss concerns raised in GSI-191, for example, NUREG 6224. The experiments investigated a wide range of particulate size distribution and types (for example, different forms of sil-icon carbide and iron oxide) and, in these ex-periments the NUREG 6224 correlation has been shown to bound actual head loss in beds with post-LOCA fluid flow, chemistry, partic-ulate, and bed formation prototypical of the STP plants. The experiments help in under-standing the uncertainty and margin in the analysis where the head loss from many dif-ferent hypothesized break sizes and locations with different debris loads had to be exam-ined.

Because current testing of STP conditions has not been fully comprehensive, a multi-plier has been applied to all debris-bed head loss calculations to compensate for residual uncertainties.

2.1.2 Safety margin.

In each scenaxio, the tails of extreme distri-butions are sampled and propagated through to the PRA. Where appropriate, the uncer-tainty distributions envelope attributes of both aleatory uncertainty and epistemic un-certainty. As explained later in this report, the only component of epistemic uncertainty that is explicitly preserved in the present analysis is the component attributable to the break-frequency size distributions taken from NUREG-1829. All other sources of variabil-ity have been integrated into the estimates of failure probability reported for the compos-ite failure modes used in the PRA. Compos-ite failure modes applied in the PRA include:

(1) strainer failure by excessive differential pressure, excessive deaeration, and mechani-cal buckling; (2) core blockage; and (3) boron precipitation. Also, experimental results for chemical effects were obtained with existing amounts of aluminum exposed to post-LOCA fluids and they indicated very little to no pre-cipitate formation.

Although such an extreme scenario would never be expected based on realistic analy-sis of the LOCA response, thermal-hydraulic engineering evaluations of core flow block-age scenarios were conducted to understand safety margin in these scenarios. In these evaluations [69], assessments of extreme con-ditions of core blockage are included. In these analyses, it was shown that with complete blockage of the core inlet and all bypass paths, only medium and large break cold leg LOCA would result in core damage. In addi-tion, detailed modeling of the core and re-actor vessel showed that only one fuel as-7 NOC-AE-1 3002954 2 ENGINEERING ANALYSIS 2.2 Risk Impact sembly flow passage needs to remain clear to prevent fuel overheating. The analyses in-cluded locating the open fuel assembly either at the core center or at an extreme periphery location. Multi-dimensional vessel and core simulations at the time of recirculation show that the core inflow is highly asymmetric in-dicating that it would be likely that several fuel assemblies would not be blocked by de-bris that might penetrate the ECCS sump screens.

The chemical effects testing that has been conducted so far has shown that chemical precipitation does not tend to occur in so-lution in the STP post-LOCA environment.

In cases where precipitation does occur, the current test results suggest that the precipi-tates that actually form in solution have dif-ferent morphology from the surrogate pre-cipitates and are likely to have less impact on total head loss. It is possible that under some extreme scenarios, chemical effects may be more significant than those observed dur-ing the recently completed tests. To address this possibility, a chemical effects bumnp-up factor probability distribution with a tail in-cluding 15x, 18x, and 24x increases for small, medium, and large breaks, respectively, was included in the CASA Grande evaluation. The purpose of the extreme tail was to preserve a 10-05 probability of meeting or exceeding the stated limits while also preserving expec-tation values between 2 and 3 (factors of 2x to 3x) for each LOCA category. In addition, the contributions of chemical effects from limiting experiments with ex-situ prepared precipitates [23] are inherently assumed in the core flow blockage success criteria which was developed as a bounding value for all PWRs. Several other conservative assump-tions leading to safety margin in the as-built, as-operated plant are detailed by NEI [35].

All STP large bore piping PWSCC-susceptible welds (nozzle welds) have been replaced or otherwise mitigated with the ex-ception of the Reactor Vessel nozzle welds.

The reactor vessel nozzle welds are less of a concern in the GSI-191 analysis than other break locations since the reactor vessel is cov-ered with RMI, and the primary shield wall would protect the majority of fiberglass insu-lation in the steam generator compartments.

STPNOC is in compliance with ASME Sec-tion XI weld inspections.

The insulation, paint and concrete damage choice of the Z0130 used in the STPNOC en-gineering calculation is expanded to account for pipe whip. No credit is taken in the cal-culations for piping constraints (especially on large bore pipes) that would reduce the ZOI based on pipe whip restraint. Finally, Ballew et al. [1] have shown the choices of the ZOIs used in the GSI-191 risk analysis is signifi-cantly overestimated [34, Section 3.4.2].

2.2 Evaluation of risk impact The risk assessment shows that any increases in CDF 31 and risk are very small and con-sistent with the intent of the NRC's Safety Goal Policy Statement. The expected change in CDF and LERF is very small in the analy-sis which includes internal and external haz-ards in an at-power model which bounds risk contribution. An in-depth and comprehen-sive risk assessment using the STPNOC PRA was used to derive the quantified estimate of the total impact of the proposed change as opposed to a qualitative assessment using, for example, performance measures.

Because pressures and temperatures are greatly reduced in plant operating Modes 4, 5, and 6, the concerns raised in GSI-191 can not be realized in these shutdown modes of operation. For Mode 3, the at-power model is bounding and can be used as a surrogate for Mode 3 operation.

The quantitative risk metrics evaluated in 30Zone of Influence 31Core Damage Frequency 8

NOC-AE-1 3002954 2.3 PRA Adequacy 2 ENGINEERING ANALYSIS the analysis are CDF and LERF. There may be risk metrics that are not reflected (or are inadequately reflected) by changes to CDF and LERF. Other risk metrics were consid-ered, especially effects on containment in-tegrity. However, there are no concerns re-lated to GSI-191 that have a bearing on con-tainment integrity following a LOCA identi-fied in the analysis. Therefore, there is no effect on LERF and, therefore, no impacts to offsite consequences.

The STPNOC PRA has been reviewed on multiple occasions by the NRC. The last in-dependent peer review was for STPNOC PRA Revision 5 and Revision 5 was assessed in that review to be adequate for use in STP-NOC PRA applications. Since Revision 5, there have not been any major changes to the PRA that require additional peer review.

The PRA is currently at Revision 7, released late in 2012. The concerns raised in GSI-191 are isolated to long-term cooling in LOCAs.

Other initiating events included in the PRA are, therefore, unimportant compared to the LOCA event trees. The STP baseline CDF and LERF are substantially below the Com-mission Safety Goal when the as-built, as-operated plant risk is evaluated with the con-cerns raised in GSI-191 included. That is to say that there is very little risk associated with the concerns raised in GSI-191 because, when incrementally added into the analyzed average plant risk, the risk contribution is negligible.

LOCA in more detail, the main concerns are with MLOCA 32 and LLOCA. As mentioned earlier in Section 2.1.2, thermal-hydraulic re-sponse analysis shows that long-term core cooling is not challenged in SLOCA 33 sce-narios. The STPNOC PRA, like other simi-lar PRAs, included a very simplistic demand failure probability for recirculation failure.

The GSI-191 risk analysis required a much better understanding of the failure proba-bility and concomitant uncertainty for re-circulation failure than the simplistic basic event value used in the past. In order to support a more informed basis for recircu-lation failure, the basic event likelihood and uncertainty needed engineering analysis sup-port. A detailed uncertainty quantification was performed to solve the required engineer-ing models and propagate their uncertainty to obtain a recirculation failure probability.

Similarly, the basic event failure likelihood and uncertainty for ECCS pump performance only included mechanical and electrical fail-ures. However, the concerns raised in GSI-191 required an assessment of the likelihood for air ingestion and inadequate NPSHA when debris beds are hypothesized to form on the ECCS sump strainers. These added failure mechanisms were included in the PRA with their failure probability and uncertainty de-termined through uncertainty propagation of appropriate physical models as described in detail in Volume 3. The failure thresholds for these kinds of events are from a stan-dard engineering analysis of allowable air and NPSHA for the pumps during a worst case LOCA scenario.

Finally, downstream effects of core block-age and boron precipitation were included with the possibility of recirculation failure.

Again, the added failure mechanisms were included in the PRA with their failure prob-ability and uncertainty determined through 32Medium Break Loss of Coolant Accident 33Small Break Loss of Coolant Accident 2.3 Technical the PRA adequacy Analysis The STPNOC PRA is a full-scope integrated Level I and Level II PRA. Further details con-cerning the technical adequacy of the STP-NOC PRA are found in Volume 4. However, and as mentioned earlier, the GSI-191 con-cerns center around LOCA and in particular, the recirculation phase of LOCA. Going into 9

NOC-AE-1 3002954 2 ENGINEERING ANALYSIS 2.3 PRA Adequaqy uncertainty propagation of appropriate phys-detailed engineering analysis is performed ical models.

in an uncertainty quantification framework 2.3.1 Scope of the PRA The scope of the STPNOC PRA is Level I and Level II, including external and inter-nal hazards such as internal floods, seismic events, internal fires, high winds, external flooding, etc. This level of detail is actually not required because none of the LLOCAs are evaluated in external events so consequently, GSI-191 issues do not appear. The concerns raised in GSI-191 are related to LOCA and, in fact, the at-power LLOCA and MLOCA initiating events are the most important of the concerns. The STPNOC PRA is an at-power PRA and, as such, no shutdown LOCA events are considered. The at-power scenar-ios bound the low power and shutdown LOCA events, not only because the decay heat load is significantly reduced, but because the en-ergy available for debris generation is much less. Therefore, the STPNOC PRA overall scope is sufficient to address the concerns as-sociated with GSI-191.

With relationship to LOCA, the STPNOC PRA Revision 7 initiating event frequency is taken from the most recent database used in PRA analyses [11]. Eide et al. refer to NUREG/CR 1829 [68] as the basis for LOCA initiating event frequencies. The frequencies used in the STPNOC PRA LOCA initiating event trees are preserved in the engineering analysis used to develop failure probabilities at locations throughout the Class 1 piping in the STP containment buildings. Also, the LOCA epistemic uncertainties used in the en-gineering analysis are taken from the same NUREG-1829 table used by Eide et al..

2.3.2 Level of detail As mentioned previously, the PRA is not significantly changed to specifically address the concerns raised in GSI-191. Instead, a that evaluates the required failure modes of ECCS and core cooling (in-vessel effects).

Significant detail is included in the engi-neering analysis used to develop the new basic events and top events required. De-tails include physical models and mecha-nisms known to lead to failure, and the anal-yses include experimental evidence used to support particular areas of concern.

2.3.3 Technical adequacy The safety issues associated with GSI-191 are within the scope of current PRAs that meet Regulatory Guide 1.200 [36, 37], Revision 1 or Revision 2. LOCAs are internal event ini-tiators included in all versions of Regula-tory Guide 1.200. The STPNOC PRA has been peer reviewed relative to internal events (includes LOCA initiators). Since STPNOC's PRA is compliant with RG 1.200, Revision 1 for internal events, it is compliant with Reg-ulatory Guide 1.200, Revision 2 for assessing the risk associated with GSI-191.

The technical adequacy of the PRA analy-sis is robust. The assumptions and/or actual modeling of the concerns raised in GSI-191 are either bounded in other work by exper-imental evidence or analysis, or by analysis and experimentation specifically performed for the STPNOC PRA evaluation. The STP-NOC PRA is used in risk-informed applica-tions extensively at STP.

The methodologies, applications, and re-sults derived from the STPNOC PRA are re-viewed by peers in benchmarking and other activities and are also regularly published in the open literature and symposia. These are, for example, Liming and Kee [25], Liming et al. [26], Moiseytseva and Kee [32], Kee et al. [21], Galenko et al. [15], EPRI [12],

Wang et al. [71], Kee and Yilmaz [22], Kee and Popova [19], Yilmaz et al. [76], Yilmaz and Kee [73], Rodgers et al. [43], Kee et al.

10 NOC-AE-1 3002954 2 ENGINEERING ANALYSIS 2.3 PRA Adequacy

[20], Yilmaz and Kee [74, 75]. In some cases, STPNOC has been the industry leader in PRA applications and application develop-ment, and in setting standards and practices.

In the GSI-191 risk-informed resolution, STP-NOC has followed the practices and methods known to be acceptable and consistent with industry PRA practices and standards.

2.3.4 Plant representation The STPNOC PRA and the engineering anal-ysis supporting the GSI-191 analysis are rep-resentative of the as-built, as-operated plant.

The STPNOC PRA is reviewed for compli-ance/adherence with the plant design and plant data review every 36 months as a UF-SAR Chapter 13.7 commitment required for PRA applications. Section 2.3.4.1 is a sum-mary of the engineering analysis support-ing the PRA analysis in the STPNOC Pi-lot Project. The STPNOC PRA configuration control is in accordance with STPNOC plant processes [4].

2.3.4.1 Model of the LOCA pro-cesses, CASA Grande One of the pri-mary functions served by CASA Grande in the STPNOC Pilot Project is quantifying con-ditional failure probabilities related to GSI-191 phenomena for various plant Modes and ECCS operating states. Failure probabilities are passed to the PRA to determine the de-cision metrics for acceptance. Three new top events are added to the PRA to accommo-date composite GSI-191 failure processes:

• failure at the sump strainer;

" boron precipitation in the core; and

• blockage of the core.

These three composite failure probabilities are calculated by testing the outcome of ev-ery postulated break scenaxio against seven performance thresholds:

• (1) strainer AP > NPSHR Margin;

• (2) strainer AP > Pbuckl3,

• (3) strainer Foi 0 d3 5 > 0.02;

• (4) core fiber load > CLB36 fiber limit for boron precipitation;

• (5) core fiber load > HLB 37 fiber limit for boron precipitation;

• (6) core fiber load > CLB fiber limit for flow blockage; and

• (7) core fiber load > HLB fiber limit for flow blockage.

(1) through (3) above are counted as fail-ures if any single operable strainer exceeds the performance thresholds at any time dur-ing the 36-hornu calculation. (4) through (5) are assessed against the accumulated fiber penetration from all operable strainers, and they must exceed the performance thresh-old before the time of Hot-Leg injection to be counted as failures. The thresholds for (5) were set infinitely high so that only exceedance of the CLB boron precipitation loading (4) was recorded as failure. This approach is reasonable because the thresh-old for failure in (4) is substantially lower than for (5) through (7), and because (4) through (7) all depend on (1) through (3),

and all the performance thresholds depend on the same internal flow distribution and fiber accumulation processes.

Violation of any of the seven performance thresholds is counted as an independent fail-ure. Thus, it is possible that a single scenario can contribute to both a strainer-related fail-ure tally, and a core-fiber-load failure tally.

After a suite of scenarios is performed, the 34Strainer mechanical failure limit 35Void Fraction "6Cold Leg Break 37Hot Leg Break 11 NOC-AE-1 3002954 2 ENGINEERING ANALYSIS 2.3 PR.A Adequacy sum of probability weights for failed scenar-ios within each LOCA category is divided by the sum of probability weights for all scenar-ios within each LOCA category to generate the conditional failure probabilities needed for the PRA. Table 1 reports the mean con-ditional failure probability associated with each composite failure mode for each of five plant operating states (Cases). No failures were recorded for small or medium-break events, and later discussion will explain that only the higher range of large-break events contributed to failure. In addition to the composite PRA failure modes, total failure probability conditioned on the LOCA cate-gory is also provided.

Table 1 results can be interpreted in the following ways. Design-basis accident re-sponse with three trains operable (Case 1) is estimated to incur a total failure probability of 0.09% given that a LLOCA occurs (that is, 9 failures in every 10,000 large-break events).

If only one train is operable (Case 43), this estimate increases to 0.45%. The primary contributor to the increase is the additional head loss incurred at the single strainer by collecting all of the debris that is distributed in proportion to flow across three strainers under Case 1. Conversely, failures incurred by exceeding the boron fiber load are reduced (compare first and last columns) because less cumulative fiber is penetrating the sin-gle, highly loaded strainer. Blockage failure is reported as zero probability because the thresholds were set very high, partly to avoid double counting blockage failures for events that first exceed the bounding low value for fiber-load thresholds related to boron precip-itation in the core.

Conditional failure probabilities reported in Table 1 are described as "mean" or "ex-pected" values because five point estimates associated with independent samples of the NUREG-1829 break frequency envelope have been averaged for use in the PRA. The fol-lowing discussion explains the origin and the mechanics of this averaging process.

The NUREG 1829 tables [68] assign con-fidence levels to estimates of annual occur-rence frequency as a function of break size.

This assignment of confidence level defines an envelope of epistemic uncertainty that was fit using bounded Johnson probability density functions at each discrete break size for which percentiles of confidence were tab-ulated. The purpose of these fits was to en-able interpolation of the confidence bands at any intermediate break size of interest.

The relationship defined by NUREG 1829 between annual occurrence frequency (events per year) and break size is presented in terms of a ccdf38. This format implies that under-lying probability density function, pdf3 9 has been integrated, and it is important to con-sider the form of the pdfs before selecting an interpolation scheme that will be applied to the ccdfs. Conversely, any presumption about interpolation of the ccdf has implications for the implied form of the pdf.

A pdf defined for break size must define the probability per unit of size that a break occurs within the interval between the dis-crete sizes tabulated in NUREG 1829. With-out knowing the details of how fracture me-chanics processes were treated during compi-lation of the NUREG 1829 table, it is difficult to defend any assumption other than uni-form probability density between the tabu-lated discrete sizes. Uniform probability den-sity means that any break size within the in-terval is equally likely. Uniform (constant) break-size probability between two ccdf val-ues is easily calculated as the positive dif-ference between the complementary cumula-tive annual frequencies divided by the posi-tive range of size across the interval divided by the total annual exceedance frequency for 3"Complementary cumulative distribution function 3 9Probability density function 12 NOC-AE-1 3002954 2 ENGINEERING ANALYSIS 2.3 PRA Adequacy Table 1: Mean LBLOCA conditional failure probabilities for five plant operating states. Failure probabilities shown are for strainer blockage, core fiber load exceed-ing flow blockage criteria, and sump differential pressure exceeding Ppb,ckle. Each Case refers to a plant operating state.

Case 1 Case 9 Case 22 Case 26 Case 43 Blockage 0

0 0

0 0

Boron 6.94x10-0 4 1.82x10 - 03 7.51x10-05 6.15x10-0 5 3.42x1-°0 6 Fiber Load Sump 2.45x10-0 4 5.39x10-0 4 1.32x10-0 3 9.56x10- 0 4 4.45x10-0 3 Failure Total 9.38 x 10-04 2.35 x 10-03 1.40x 10-03 1.02x 10- 03 4.45 x 10-03 the smallest break size. The integral of a con-stant pdf needed to form a ccdf is a straight line, and this implies that linear-linear in-terpolation of the NUREG 1829 table is the treatment most consistent with the assump-tion of constant underlying probability den-sity.

interesting visual effect when plotted on log-log axes. As shown in Figure 5, the linear ccdf appears as a periodically looping curve on a logarithmic scale. Figure 5 illustrates the extreme endpoints of the bounded Johnson fits (solid lines) and several typical random samples of the break-frequency profile that were used in the STPNOC Pilot Project assessment (dashed lines).

NLHS of break-frequency profiles from the Johnson pdf envelope are performed in ex-actly the same manner as for all other ran-dom variables. The nonuniform probability bins are predefined based on the desired number of samples and on the direction of presumed conservatism, then random per-centiles are chosen from within each bin to represent, or "carry", the associated proba-bility weights. For the STPNOC Pilot Project, five independent random samples were ex-tracted from the Johnson envelope for each plant operating state, with an emphasis on upper percentiles of the break frequency un-certainty envelope. Given a sample of five percentiles, the Johnson fits are then in-verted to find the corresponding annual fre-quencies. It is important to note that all Johnson fits are perfectly correlated by us-break size (in)

Figure 5: Linear-linear interpola-tion of bounded Johnson extrema (solid) with nonuniform stratified random break-size profiles (dashed).

Linear-linear interpolation of the NUREG 1829 table values leads to an 13 NOC-AE-13002954 2.3 PRA Adequacy 2 ENGINEERING ANALYSIS ing the same fixed values of the sampled per-centiles. Finally, the set of annual frequencies from each Johnson fit is linearly interpolated to create the break-frequency profiles shown as the dashed lines in Figure 5.

Each break frequency profile is fully ana-lyzed in CASA Grande using a set of three batch replicates containing approximately 2,250 break scenarios each to obtain a point estimate of failure probability for the com-posite modes. Residual sampling imprecision of 20% between the three replicates is typi-cal of this scenario sampling size. Probabil-ity weights from stratified sampling of the Johnson envelope are then used to form the weighted conditional means reported in Ta-ble 1.

The current resolution used for batch size (2250 breaks), replicates (3) and break-frequency sampling (5) was dictated by prac-tical evaluation times. Table 2 summarizes the five point estimates and their associ-ated probability weights generated for the total failure probability under plant operat-ing state, Case 43 (one train operable). The weighted mean is formed simply by multi-plying each point estimate by its probability weight and adding the products. Similar dis-tributions were formed for all composite fail-ure modes and for all plant operability states, but only the weighted means are presented in Table 1.

The cumulative distribution defined for to-tal failure probability under Case 43 (one train operable) in Table 2 is plotted in Fig-ure 5 to illustrate how epistemic quantiles could be preserved from the GSI-191 engi-neering analysis CASA Grande. This distri-bution reflects only the uncertainty inher-ent to the estimation of annual break fre-quency. All other random variability, includ-ing ranges on physical phenomena and de-cision criteria, has been integrated into each point estimate. As shown in Table 1 and Fig-ure 5, typical variation in failure probability estimates spans a factor of 2 to 4 between the minimum and maximum values (0.012/0.003

= 3.8). This variation is caused solely by the shape of the randomly selected break-frequency profiles, which dictate the relative proportion of break frequency by size.

It is important to reemphasize that CASA Grande never makes any direct use of the an-nual break frequency as a time-rate quan-tity. All analyses proceed conditioned on the assumption that a break has already oc-curred. Sample profiles taken from the break-frequency envelope then describe how to par-tition the relative occurrence of breaks by size. CASA Grande further redistributes the relative size probability across weld types in order to map the cumulative probability of a break as a function of size to discrete loca-tions in the plant [42].

The PRA samples directly from the NUREG 1829 Johnson pdf fits in each cat-egory to preserve the epistemic uncertainty in LOCA frequency. It is important for CASA Grande to use exactly the same represen-tation of the epistemic uncertainty. The Johnson fits are evaluated analytically in CASA Grande to generate a table of em-pirical pdfs that are manually passed to the PRA (RISKMANTM model) for repeated sampling in the risk quantification. Although the PRA generates thousands of samples from the Johnson pdf during quantification, CASA Grande samples relatively sparsely here. CASA Grande uses one quantification loop to generate point estimates of failure probability that are based on parameter vari-ations and model uncertainties like chem-ical effects bump up, and an outer loop to preserve the epistemic quantiles of the break-frequency envelope (see Section 2.5.1).

Sparse sampling of the epistemic envelope is a consequence of placing emphasis on aleatory uncertainties (inner loop) that drive the outcome of each break scenario and re-lies on NLHS for generating unbiased esti-14 NOC-AE-1 3002954 2 ENGINEERING ANALYSIS 2.3 PRA Adequacy Table 2: Distribution of total conditional failure for LLOCA under Case 43 (one train operating).

Point Failure Johnson prob-Cumulative Probability ability weight probability 0.0 0.0 0.0 3.13 x 10- 0 3 8.22 x 10-0 8.22 x 10-01 7.49 x 10-03 4.62 x 10-o3 8.27x 10-01 1.03x 10-02 1.46x 10-01 9.733x 10-01 1.15 x 10-02 1.00 x 10-03 9.74 x 10-01 1.2x 10-02 2.60x 10-o.

1.0 4.45 x 10-03 weighted mean mates of the mean failure probability. Failure distributions similar to those shown in Fig-ure 6 could alternatively be sampled by the PRA to generate distributions of incremen-tal risk attributable to GSI-191 phenomena.

A sampling scheme would necessarily pre-serve epistemic correlation in the distribu-tion of failure probability that is generated by CASA Grande (Figure 6) and shared by the RISKMANTM model.

Another key piece of information passed from CASA Grande to the PRA through the basic events supported is the conditional split fraction for cold leg breaks in each LOCA category. The total break size prob-ability for a single NUREG 1829 profile is distributed across all welds in containment using the hybrid weighting scheme [42] to ac-count for the contributions of small breaks on large pipes to the small and medium LOCA categories. Each break scenario sam-pled from this process carries a specific size and location and a fractional weight of the total break-size probability. Before any other physical parameters are considered, the dis-tribution of probability weight can be par-titioned into HLB and CLB events and by LOCA size.

0.92 0.96 0 94 0.92 COUo.adw U,11 of Total faduMe PMfro96Oy fa C~se 43 2

4 6

a 10 12 14 Figure 6: Empirical distribution of total failure probability for Case 43 (one train operating) based on five discrete samples of the NUREG 1829 break-frequency un-certainty envelope. Weighted mean

= 4.45x 10- 0 3 marked as bold dot.

15 NOC-AE-1 3002954 2 ENGINEERING ANALYSIS 2.4 Acceptance Guidelines Table 3: All cold leg split fractions conditioned on LOCA categories small, medium, and large for Case 43. The fraction going to the hot leg is simply the complement of the cold leg fraction.

Total Small Medium Large 4.2052034x 10-0' 4.2962813x 10-01 4.8133459x 10-01 4.3059826x 10-01 4.2052034 x 10-°'

4.2962813 x 10-01 4.8133459 x 10-01 4.3059826x 10-01 4.2052034x 10-01 4.2962813 x 10-°'

4.8133459 x 10-01 4.3059826x 10-0' 4.2015626x 10-01 4.2933789x 10-°1 4.8133521 x 10-01 4.3048163x 10"'

4.2015626x 10-01 4.2933789,< 10-0 4.8133521 x 10-01 4.3048163x 10-01 4.2015626x 10"1 4.2933789x 10-1i 4.8133521 x 10-01 4.3048163x 10-0i 4.2014556x 10-01 4.2932931 x 10"1 4.8133576 x 10-01 4.3044256 x 10-01 4.2014556 x 10-01 4.2932931 x 10-01 4.8133576x 10-01 4.3044256 x 10-0i 4.2014556 x 10-01 4.2932931x 10-01 4.8133576 x 10-01 4.3044256x 10-01 4.2210420x 10-01 4.3087029 x 10-01 4.8133228x 10-01 4.3115092x 10-'1 4.2210420x 10-01 4.3087029x 10-01 4.8133228x 10-01 4.3115092x 10-01 4.2210420x 10-01 4.3087029x 10-01 4.8133228x 10-01 4.3115092x 10-01 4.3407111x 10-01 4.3931916x 10-01 4.8118954x 10-01 4.3960731x 10-0' 4.3407111 x 10-01 4.3931916x 10-01 4.8118954xx10-1 4.3960731 x 10-01 4.3407111 x 10-01 4.3931916x 10-01 4.8118954xx10-1 4.3960731x 10-°1 Table 3 itemizes all cold-leg split fractions obtained for the fifteen batches associated with Case 43. These values were obtained by dividing the sum of probability weights for CLBs in each LOCA category by the sum of probability weights for all breaks in the LOCA category. HLB split fractions are sim-ply the complement of any single entry in the table. Three replicates of 2,250 scenarios are evaluated for each of five break-frequency profiles for a total of 3x2250x5 = 33,750 scenarios per plant operating state. CLB split fractions are mildly dependent on the break-frequency profile shape (note repetition in successive groups of three rows), but they are independent of plant operating state. It is interesting to note that proportion of large CLBs is substantially smaller than the 50%

proportion assumed in the 2011 [5] quantifi-cation.

Table 4 lists a sample of the specific welds, break sizes, and general containment zones that are associated with one or more fail-ure modes in Case 43. This list includes only the first 34 of 1659 failed scenarios that were tallied during the analysis. The fact that no SLOCA or MLOCA events have been recorded as failure for any scenario evaluated in this quantification is a strong indication that there is a minimum size break below which insufficient debris can be formed to challenge the safety systems. The same con-sideration explains why most failure scenar-ios involve the DEGB assumption of spheri-cal ZOI simply because more insulation vol-ume can be involved in debris generation.

The above illustration regarding Case 43 in-dicates the kinds of insights that can be re-alized in the STPNOC Pilot Project analysis approach.

2.4 Acceptance Guidelines Regions are established on the phase planes defined by ACDF, CDF and ALERF, LERF, as illustrated in Figure 7 and Fig-16 NOC-AE-1 3002954 2 ENGINEERING ANALYSIS 2.4 Acceptance Guidelines Table 4: Sample attributes of break cases leading to failure for Case 43. In the table: Pipe is a text string defined in the inservice inspection program program; System refers to STP System (all are RCS); Break Size is the size of the break in inches; LOCA size values of 1,2,3 denote small, medium, large LOCA events (all are large); DEGB, YES denotes the fully severed pipe condition (failures dominated by DEGB); RCS Leg denotes break location (CLB or HLB); and Break Location denotes region in the containment building related to debris transport fractions.

Pipe 12RC-1112-BB1 12RC-1112-BB1 12RC-1 12-BB1 12RC-1125-BB1 12RC-1125-BB1 12RC-1125-BBl 12RC-1125-BBl 12RC-1125-BB1 12RC-1125-BB1 12RC-1212-BB1 12RC-1212-BB1 12RC-1212-BB1 12RC-1212-BB1 12RC-1221-BB1 12RC-1221-BB1 12RC-1221-BB1 12RC-1221-BB1 12RC-1221-BB1 12RC-1221-BB1 12RC-1221-BB1 12RC-1312-BB1 12RC-1312-BB1 12RC-1312-BB1 12RC-1312-BB1 12RC-1312-BB1 12RC-1322-BB1 12RC-1322-BB1 12RC-1322-BB1 12RC-1322-BB1 16RC-1412-NSS 16RC-1412-NSS 16RC-1412-NSS 16RC-1412-NSS 16RC-1412-NSS 16RC-1412-NSS 16RC-1412-NSS 16RC-1412-NSS 16RC-1412-NSS 16RC-1412-NSS System Break Size RCS 10.126 RCS 10.126 RCS 10.126 RCS 10.126 RCS 10.126 RCS 10.126 RCS 10.126 RCS 10.126 RCS 10.126 RCS 10.126 RCS 10.126 RCS 10.126 RCS 10.126 RCS 10.126 RCS 10.126 RCS 10.126 RCS 10.126 RCS 10.126 RCS 10.126 RCS 10.126 RCS 10.126 RCS 10.126 RCS 10.126 RCS 10.126 RCS 10.126 RCS 10.126 RCS 10.126 RCS 10.126 RCS 10.126 RCS 12.814 RCS 12.036 RCS 12.814 RCS 11.273 RCS 12.090 RCS 12.118 RCS 12.814 RCS 12.814 RCS 12.814 RCS 12.814 LOCA Size 3

3 3

3 3

3 3

3 3

3 3

3 3

3 3

3 3

3 3

3 3

3 3

3 3

3 3

3 3

3 3

3 3

3 3

3 3

3 3

DEGB RCS Leg YES Hot YES Hot YES Hot YES Cold YES Cold YES Cold YES Cold YES Cold YES Cold YES Hot YES Hot YES Hot YES Hot YES Cold YES Cold YES Cold YES Cold YES Cold YES Cold YES Cold YES Hot YES Hot YES Hot YES Hot YES Hot YES Cold YES Cold YES Cold YES Cold YES Hot NO Hot YES Hot NO Hot NO Hot NO Hot YES Hot YES Hot YES Hot YES Hot Break Location SC Compartment SC Compartment SG Compartment SC Compartment SG Compartment SG Compartment SC Compartment SG Compartment SG Compartment SC Compartment SG Compartment SG Compartment SG Compartment SG Compartment SG Compartment SG Compartment SG Compartment SG Compartment SG Compartment SG Compartment SG Compartment SG Compartment SG Compartment SG Compartment SC Compartment SG Compartment SG Compartment SC Compartment SG Compartment SG Compartment SG Compartment SG Compartment SG Compartment SG Compartment SG Compartment SC Compartment SG Compartment SG Compartment SG Compartment 17 NOC-AE-1 3002954 2 ENGINEERING ANALYSIS 2.5 Comparison with Guidelines ure 8. Acceptance guidelines are established for each region as discussed below. The fig-ures show shading as the values increase on either axis. The shading indicates that greater scrutiny and support would be re-quired for values that approach the region boundaries. Also illustrated, in the figures, is the desired trajectory for changes. That tra-jectory can be realized by using resources on projects that have the maximum risk benefit, a concept that is consistent with the Com-mission's direction to use risk insight to best achieve safety goals.

The comparison in the STPNOC GSI-191 analysis uses the full-scope (including in-ternal and external hazards, at-power, low power, and shutdown) assessment of the change in risk metric and the baseline value of the risk metric (CDF and LERF). As noted above, the shutdown PRA analysis is bounded by the at-power model. In the STP-NOC GSI-191 analysis, the maximum accept-able increase in CDF is 10-06 and the maxi-mum acceptable increase in LERF is 10-07.

t0 is' 10' lo" LERF 01 Figure 8: Reproduction of Figure 5 from Regulatory Guide 1.174, "Ac-ceptance guidelines for large-early-release frequency",

the ALERF, LERF phase plane.

2.5 Comparison of PRA 0

10 results with acceptance guidelines The STPNOC Pilot Project PRA quantifica-tion is detailed in Volume 2. As mentioned previously, the quantification shows that the risk associated with the concerns raised in GSI-191 are very small when compared to the acceptance criteria of RG1.174.

The PRA used in the GSI-191 licensing basis change does not rely solely on nu-merical results for change in risk. Instead, the choice of models, solution methodology and incorporation of uncertainties provides a high level of confidence that the uncertainties in models' parameters has been properly ac-counted for in the results. The safety margin described in Section 2.1.2 associated with use of the methodology reflected in the license basis change analysis provides assurance that safe operation can be expected without re-liance on numerical results alone.

As mentioned in Section 2.3.3, the STP-NOC PRA is an integrated Level 1 model 16' to' 10" CDF 3-Figure 7: Reproduction of Figure 4 from Regulatory Guide 1.174, "Ac-ceptance guidelines for core damage frequency", the ACDF, CDF phase plane.

18 NOC-AE-1 3002954 2 ENGINEERING ANALYSIS 2.5 Comparison with Guidelines that includes all internal and external events, Level 1 and Level 2 analysis, the focus of the GSI-191 concerns are related to LOCA.

The analysis of LOCA initiating event fre-quencies and local pipe failure probabili-ties included in development of the basic events for the scenarios that address the con-cerns raised in GSI-191 include the full range of the epistemic uncertainty at each break size. Qualitative conservatisms that increase safety margin (as previously mentioned in Section 2.1.2) are included along with the quantifiable uncertainties to increase confi-dence in the adequacy of the results.

The STPNOC PRA analysis includes un-certainties that have been postulated in de-terministic analyses for the concerns related to GSI-191:

• ZOI;

" Chemical effects;

" Debris transport;

• Head loss;

" Boric acid precipitation; and

• Air ingestion to ECCS pumps.

In some cases, the uncertainties have been addressed through well-known conservative approximations, in other cases, specific ex-perimentation has been performed to analyze the impact of the phenomena on plant per-formance in response to LOCA.

2.5.1 Types of uncertainties and methods of analysis Both aleatory and epistemic uncertainties have been included in the STPNOC PRA.

As mentioned in the previous section (Sec-tion 2.5),

uncertainties have also been addressed using conservative assumptions where appropriate or where large uncertain-ties are seen. For example, assuming a larger ZOI will result in scenarios that are conser-vative.

2.5.1.1 Comments on uncertainty types In the PRA community, the concept of "separate" types of probabilities or un-certainties is discussed frequently. In other communities, probability is simply probabil-ity and following quantification there is no distinction as to the source. (See Chapter 3 of [33] on PHSA 40 for a discussion including, "The panel concludes that, unless one ac-cepts that all uncertainty is fundamentally epistemic, the classification of PHSA uncer-tainty as aleatory or epistemic is ambigu-ous.") So in an uncertainty quantification framework in which the goal is to obtain as output a point estimate or a probability dis-tribution on a key performance measure by propagating the probability distributions as-sociated with multiple sources of input un-certainty, there is typically no attempt to sort out the contribution due to each source of input uncertainty. That said, it is com-mon practice to carry out a parametric anal-ysis in which we effectively remove the prob-ability distribution associated with an input parameter and simply vary the input param-eter over a range of plausible values in or-der to assess the effect on the output for

'the key performance measure. Applying this idea amounts to analyzing the output in a conditional manner, conditioned on the value of the corresponding input parameter. Such parametric analyses are usually done for one source of uncertainty at a time, as opposed to trying to simultaneously vary multiple input parameters.

Now reconsider the probability distribu-tion on the input parameter of focus. Output results for the key performance measure can be reported conditioned on the value of the input parameter, in turn, set to be specific quantiles from the input parameter's proba-4"Probabilistic Seismic Hazard Analysis 19 NOC-AE-13002954 2 ENGINEERING ANALYSIS 2.5 Comparison with Guidelines bility distribution. In this sense we can pre-serve the quantiles associated with a key in-put parameter when analyzing distributional output. The engineering analysis used to de-velop the basic event failure probabilities for the PRA uses an approach, likely new to PRA practitioners, that optionally inte-gTates all uncertainty or preserves the quan-tiles of selected input distributions (which some may wish to label as being epistemic uncertainty). The LOCA frequency, for exam-ple, has a large uncertainty envelope that has been preserved preserved in this manner. An-other large uncertainty envelope that could be preserved in this way is the ECCS strainer differential pressure. By preserving the un-certainty quantiles for selected sources, their effect can be explicitly observed in the resul-tant basic event distributions.

In the STPNOC Pilot Project quantifica-tion, the LOCA epistemic uncertainty on fail-tire probability is quantified separately for each of the five ECCS pump combinations considered in the STPNOC Pilot Project anal-ysis. As a result, the failure probabilities re-sulting from GSI-191 phenomena for the five pump combination cases axe correlated with the correct initiating event frequency associ-ated with the combination.

The RISKMAN TM software used for the STPNOC Pilot Project quantification is specif-ically designed to appropriately correlate el-ements from a group to which the same pa-rameter value applies. This is accomplished using the "Big Loop Monte Carlo" option se-lected for the STPNOC Pilot Project quantifi-cation. Each trial of the "Big Loop Monte Carlo" option, a random set of values is se-lected from all input variables in the PRA model. These sample values are then used to re-evaluate all PRA model elements; that is, basic event probabilities, split fraction fail-ure probabilities, initiator frequencies, and sequence frequencies that are then summed to give the CDF and LERF. Importantly, the option is also selected for the uncertainty quantification of the difference in the PRA metrics of ACDF and ALERF so that the un-certainty in the difference is calculated cor-rectly.

The one exception to this correlation of in-put parameters among PRA model elements are those considered in CASA Grande. By necessity, the PRA is quantified using fail-ure probability distributions developed in the CASA Grande analysis which are themselves functions of many data variables. In the STP-NOC Pilot Project quantification, the GSI-191 failure probabilities are quantified separately for each of the five ECCS pump state com-binations considered in the STPNOC Pilot Project analysis. In this way, the key param-eter of the PRA sequence models (that is sump flow rates) is effectively correlated in RISKMANTM with the CASA Grande analy-sis.

In the CASA Grande analysis, failure prob-abilities associated with engineering models of LOCA phenomena are also evaluated sep-arately for five percentiles of the LOCA fre-quency uncertainty analysis. These five sets of results are the basis for the five-bin uncer-tainty distributions on each of the GSI-191 phenomena failure probabilities.

The sparse sample of five bins on the distribution of failure probability is not an inherent limitation of the CASA Grande methodology, but was chosen only for the sake of current practicality. A more com-plete interrogation of the break-frequency uncertainty distribution can be made de-pending on the needs of the PRA. The ini-tial presumption was that higher percentiles of the break-frequency distribution would lead to more conservative estimates of CDF and LERF, so more sampling resolution was placed in the upper tails of the envelope (see Fig. 5). The shape of each break-frequency profile defines the relative LOCA frequen-cies as a function of break size, as reflected 20 NOC-AE-1 3002954 2 ENGINEERING ANALYSIS 2.5 Comparison with Guidelines in the variation between the five point es-timates of failure probability. RISKMAN TNI samples from the full uncertainty distribu-tion, using 100 percentiles, for the absolute LOCA frequency and correlates each sample when evaluating the MLOCA and LLOCA ini-tiating event frequencies. The correlation be-tween the uncertainties in the relative break sizes used in the CASA Grande analysis and the absolute LOCA frequencies used in the PRA sequences models is not believed to be significant and therefore not modeled.

2.5.2 Parameter uncertainty Parameter uncertainties are addressed perva-sively in the STPNOC PRA analysis. For the physical models addressing the concerns of GSI-191, input parameters were derived from both historical data and physical limits (for example, total contained volume in a tank).

The uncertainty associated with all impor-tant parameters has been included and sam-pling of the parameter distributions was done in LHS 4 1 schemes to accurately preserve the distribution. Human error probabilities are included in the STPNOC PRA however, for the most severe accident scenarios (that is LLOCA), there is very little opportunity for human actions to cause increases in the fail-ure likelihood. In these cases, automatic ac-tuation of the ECCS will occur prior to oper-ator intervention.

2.5.3 Model uncertainty As described on Page vii, the STPNOC PRA is supplied with failure probabilities result-ing from GSI 191 phenomena developed from engineering models of the phenomena asso-ciated with the concerns raised in GSI-191.

That is, in the PRA, the models are devel-oped to be accurate representations of the plant including parameter uncertainties.

4 1Latin Hypercube Sampling Over many years of study, the phenom-ena associated with the concerns raised in GSI-191 have been well characterized. How-ever, the approach taken by most investiga-tors in GSI-191 studies has been to demon-strate margin to performance limits by bias-ing inputs, not by studying uncertainty or ac-tual performance in the as-built, as-operated plant. In the STPNOC PRA, investigators matched the phenomena to the performance of the as-built, as-operated plant.

In all cases, the difference between re-sults of previous studies and results of the STPNOC GSI-191 studies can be explained by well-established analytical methods. The extensive body of work related to the is-sues raised in GSI-191 helps provide assur-ance that adequate models and methods are available to exploit.

Based on the STPNOC Pilot Project anal-ysis performed, the most important con-tribution to CDF is the model of chem-ical effects, both on the strainer and in the core. Although (as mentioned previously in Section 2.1.2) chemical effects in STP post-LOCA fluid conditions are benign com-pared to the conditions assumed for the ex-periments performed in WCAP 16793-NP, the STPNOC Pilot Project assumes that ad-verse chemical effects can occur, both at the strainer and in the core. The STPNOC Pilot Project also uses bounding values for strainer differential pressure, that is, higher differen-tial pressures than observed in experiments representative of STP conditions. The model is less sensitive to strainer differential pres-sure than core failure loading which is cho-sen at one half the 15gm/FA limit found in WCAP 16793-NP as a threshold for the po-tential of boron precipitation.

In a classical interpretation, "model un-certainty" often refers to the degree of credi-bility held by one prediction of physical phe-nomena compared to that held by alterna-tive predictions of the same phenomena. For-21 NOC-AE-1 3002954 2 ENGINEERING ANALYSIS 2.5 Comparison with Guidelines mal methods have been developed to com-2.5.4 Completeness uncertainty pare competing models that have been ini-tialized with as near identical input as possi-ble. Discrepancies between numerical predic-tions can then be used to quantify residual uncertainty in the prediction. These meth-ods can even accommodate subjective mea-sures of confidence that particular models (or none of the models) are more accurate than the others. Often, the primary difference be-tween models lies in the degree of spatial res-olution or physical fidelity, but sometimes, fully mature alternative methods are com-pared.

In the STPNOC Pilot Project, several new predictive models are being applied for the first time. These include the debris penetra-tion/filtration model that was benchmarked to test data, and the time-dependent debris circulation model that addresses coolant by-pass around the reactor core. Relatively sim-ple, first-order models are extremely useful for identifying trends, describing trade-offs between competing mechanisms, and prior-itizing risk contributors; however, additional conservatism is warranted to explicitly ac-knowledge the uncertainty associated with the predictions of first-order models. For this reason, additional conservatism was incorpo-rated in the treatment of both conventional and chemical-induced differential head-loss estimation. Additional testing is planned to more tightly correlate head loss to STP flow conditions, and additional testing is under-way to refine the probability distributions placed on potential chemical head-loss ef-fects. The practice of applying an overall in-flation factor that is distributed in magni-tude according to the best interpretation of available data represents the extent of model uncertainty that has been addressed in the prototype study this far.

Although prior investigations in GSI-191 have focused primarily on "test for success", they have nevertheless resulted in greater under-standing and characterization of the post-LOCA behavior related to the concerns raised in GSI-191. In some cases, greater under-standing has led to adoption of models that bound the experimental evidence simply be-cause the space adopted is too large to fully explore experimentally. As a consequence, simplistic conservative approaches have been adopted where uncertainty is difficult to quantify [see 35]. On the other hand, STP-NOC GSI-191 analysis has helped to extend the completeness of uncertainties associated with the concerns raised in GSI-191 by in-cluding phenomena expected to occur in the recirculation mode of ECCS operation where traditional analyses end. The STPNOC GSI-191 analysis uses realistic or prototypical con-ditions to model anticipated post-LOCA phe-nomena during all LOCA phases. Finally, where possible, uncertainties are quantified based on distributions that encompass plant conditions and equipment operating states that, although important to long-term cool-ing, are not considered in traditional (UF-SAR Chapter 15) analyses.

The confidence in completeness of the modeling scope for the concerns raised in GSI-191 is increased due to the number of years of study and work of independent in-vestigators. In the STPNOC Pilot Project, all known physical models have been adopted and evaluated in the engineering analysis supporting the PRA.

As mentioned in Section 2.5.1, epistemic uncertainty has been considered in the STP-NOC GSI-191 analysis. Examples of complete-ness uncertainties that have been considered and excluded from the current analysis are listed below:

• Multiple simultaneous RCS pipe breaks would result in reduced damage due 22 NOC-AE-13002954 2 ENGINEERING ANALYSIS 2.5 Comparison with Guidelines to the very rapid depressurizaton of the RCS. Although more damage zones would be involved, less damage would be possible at each location.

Physical security events that cause a LOCA. Such events would contribute equally to both the "ideal" plant and the as-built, as-operated plant. The STPNOC security force undergoes con-tinuous evaluation and improvements are made in processes and procedures that would help preclude such events.

" Events occurring during shutdown modes of operation (includes lifting and transport of Heavy Loads). Heavy loads are not being moved during Mode 3.

During the time heavy loads are be-ing moved, the plant is cooled down and depressurized. The STPNOC pro-cess for control of heavy loads [67] com-plies with Generic Letter 81-07, Con-trol of Heavy Loads, ANSI N14.6-1978 and NUREG 0612, and the TRM, Sec-tion 3/4.9.7.

• Structural failures (containment build-ing, interior containment walls or parti-tions, that could be postulated to in-duce a LOCA). These beyond design basis events would contribute equally to both the "ideal plant" and the as-built, as-designed plant. In both cases, it would be assumed that core damage and large early release (in the case of containment failure leading to LOCA) would occur.

" Organizational decision making and safety culture, for example see Mo-haghegh [27]. STPNOC has a STPNOC has a continuous safety culture evalua-tion program that undergoes continuous improvement and examination.

With regard to plant operating states, some can be eliminated from further evaluation.

These are under De-fueled conditions (No Mode), Refueling (Mode 6), and Cold Shut-down (Mode 5). The basis for this is that op-erating pressures and temperatures are suf-ficiently low so that piping failure mecha-nisms typically associated with LOCA events cannot reasonably be expected to occur.

Modes 1, 2, 3 and 4 are bounded by the at-power model.

The uncertainty quantification in the STP-NOC GSI-191 PRA analysis is a significant improvement in the understanding of RCS and containment building behavior under LOCA conditions. Uncertainties, not explic-itly quantified, are either bounded by other uncertainties associated with more dominant contributors or are sources of uncertainty outside the scope and boundary of GSI-191 safety issues.

2.5.5 Comparisons with accep-tance guidelines As mentioned in Section 2, the STPNOC GSI-191 analysis shows that the risk associated with the concerns raised in GSI-191 is very small. Also, as defined byNuclear Regulatory Commission [38, Figures 4 and 5, Page 16]

and previously mentioned in Section 1, the STP average CDF and LERF are also very small. The estimates of ACDF and ALERF from the STPNOC GSI-191 analysis are far less than the Region III acceptance guide-lines.

In the STPNOC GSI-191 PRA analysis, the mean values used to evaluate the accep-tance criteria are probability distributions that come from the propagation of the un-certainties of the input parameters and those model uncertainties explicitly represented in the model. The STPNOC GSI-191 PRA anal-ysis uses a formal propagation of the uncer-tainty to account for any state-of-knowledge uncertainties that arise from the use of the same parameter values for several basic event probability models.

23 NOC-AE-13002954 3 IMPLEMENTATION AND MONITORING Where epistemic uncertainties have been identified in the STPNOC GSI-191 analysis, they have been either reduced through ex-perimental evidence or bounded through as-sumption as previously mentioned in Sec-tion 2.1.2. The STPNOC PRA margin to the acceptance criteria guidelines is significant, providing confidence that any contributor to risk that may have been missed or otherwise not modeled would not make a significant change to the risk determined in the STP-NOC GSI-191 analysis.

In the STPNOC Pilot Project analysis re-liance on importance measures is not neces-sary nor used. The focus of the analysis is to understand the risk associated with the con-cerns raised in GSI-191 and importance mea-sures, while useful in evaluations concerned with other applications, are not useful in the STPNOC Pilot Project.

As discussed in Section 2.3, the STPNOC PRA is an integrated-level model that in-cludes all internal and external events (refer-ring to Level I and Level II analysis) related to the GSI-191 post-LOCA concerns. Care has been taken in the STPNOC GSI-191 PRA to ensure that all concerns associated with GS1-191 have been addressed in the analysis.

2.6 Integrated decision mak-ing As discussed extensively in Section 2.1, there are many qualitative insights that form the basis for the conclusion in the STPNOC GSI-191 PRA analysis that there is a very small risk for the concerns associated with GSI-191.

A significant effort has been expended to ex-perimentally and analytically investigate the risk and uncertainties associated with the concerns raised in GSI-191.

Traditional engineering analysis, which generally ignores uncertainty, has been en-hanced in the STPNOC GSI-191 PRA analysis by including parameter uncertainties. In as much of the analysis as possible, uncertain-ties of input parameters in the traditional engineering models are propagated through the uncertainty quantification of basic events and aggregated (with uncertainty distribu-tions) for use in PRA basic events or top events. By integrating qualitative insights, bounding uncertainties, and quantifying the uncertainties inherent in engineering mod-els, the STPNOC GSI-191 PRA analysis is a robust, integrated analysis that can be re-lied on to accurately evaluate the risk asso-ciated with the concerns raised in GSI-191.

Although the STPNOC GSI-191 PRA analy-sis relies on a full scope PRA, the analysis is specifically focused on the concerns raised in GSI-191. In particular, only the LOCA initi-ating events are of concern and the physical models are directed at long-term cooling.

3 Implementation and Monitoring As stated in Section 5, no changes are pro-posed to any programs, processes, or de-sign with regard to the current as-built, as-operated plant that would result in a sig-nificant reduction to safety margin or DID.

In particular, no changes are proposed to any ASME Section XI inspection programs

[16, 64] or mitigation strategies that have been shown effective in early detection and mitigation of weld and material degradation in PWR Class I piping applications. STPNOC has adopted other programs that help pro-vide early detection and mitigation of leak-age in other applications [17]. Additionally, no changes are proposed to design modifica-tions, processes, or programs that have re-sulted from addressing the concerns related to GSI-191 such as those mentioned in Sec-tion 2.1. In particular, design modifications that could affect any of these measures is specifically checked for in any design change 24 NOC-AE-13002954 6 DOCUMENTATION

[18, Checklist, Page 38].

6 Documentation 6.1 Introduction 4

Submittal of Pro-posed Change Proposed changes to the STP UFSAR, based on NRC approval of the STPNOC Pilot Project and LB 4 2 change to resolve GSI-191, are submitted in the attachments to letter NOC-AE-13002954 [6].

5 Quality Assurance No design, operational, or performance changes are proposed to existing safety re-lated systems, components, or structures in this analysis. Existing procedures and pro-grams are unchanged by this license basis change. The STPNOC PRA analysis support-ing the licensing basis change is performed using STPNOC PRA procedure as required for PRA analyses and assessments [3]. This is the STPNOC approved methodology for ap-plication evaluations using the PRA.

The support provided for the STPNOC PRA is performed by personnel qualified in their fields of expertise. All work performed in the licensing basis analysis is done follow-ing STPNOC procedures for contract person-nel. An oversight program, Section 7, is in effect for the duration of the entire project.

All records and documentation are controlled under the STPNOC Document Control and Records Management systems. A detailed description of the Quality Assurance pro-gram supporting the STPNOC Pilot Project is provided in Volume 4.

The total technical documentation consists of several volumes, Volume 1, Summary, Volume 2 PRA, Volume 3, the support-ing engineering analysis, CASA Grande, Vol-ume 4, Quality Assurance, and Volumes 5.1 through 5.4, Oversight. Additional documen-tation such as the PRA Model Revision 7 and support calculations are also made available through reference. In any case, all documen-tation is available in the STPNOC Records Management program.

6.2 Archival Documentation Volumes 2 and 3 of the STPNOC GSI-191 li-cense basis change submittal are detailed de-scriptions of the PRA and supporting engi-neering analyses conducted and results ob-tained. The analyses are primarily based on traditional engineering analyses that include experimental data obtained to specifically support the engineering models and analy-ses conducted as part of the licensing basis change. The full set of documentation cre-ated for this analysis are maintained as qual-ity documents for the life of plant in the RMS 4 3 and can be retrieved using the fol-lowing search fields and keywords:

" FSUG: D07090703,

" TYPE: VENDREC, and

" SUBTYPE: GSI191.

The STPNOC PRA model of record is also maintained in the RMS according to the nor-mal PRA maintenance process and can be re-trieved using the following search fields and keywords:

• FSUG: D6412, 43Records Management System 42Licensing Basis 25 NOC-AE-1 3002954 6 DOCUMENTATION 6.3 Submittal Documentation

" TYPE: DATA, and

" DOCUMENT NUMBER:

OPGPO1ZA0305.

STPNOC PRA analyses are maintained in the STPNOC RMS. The PRA analysis performed for this work can be retrieved using the fol-lowing search fields and keywords:

" FSUG: D64,

" TYPE: ANLYS, and

" DOCUMENT NUMBER: PRA13001.

6.3 Submittal Documenta-tion The STPNOC proposed license basis change is consistent with the key principles of risk-informed regulation and NRC staff expecta-tions based on the following points:

" The requirements for Long-Term Core Cooling summarized in 10 CFR§50.46 require the supporting systems to op-erate with a high level of probabil-ity including considerations of uncer-tainty. The licensing basis change re-quested quantifies the probability and uncertainty associated with long-term core cooling following the requirements as described in RG1.174. Based on the evaluation documented in the change request showing that the probability is very high that long-term core cooling will be satisfied, the impact to the li-censing basis is insignificant.

" The proposed change has no impact on existing equipment performance re-quirements or performance assessment (equipment surveillance) requirements.

For certain extremely low probability scenarios, when the extreme extent of the associated uncertainty is taken into account, the analysis shows that core damage could occur.

" No change to offsite dose or worker radi-ation dose is evaluated to occur. By im-plementing the proposed licensing basis change, a large worker radiation dose that would be incurred to mitigate a hypothesized event having insignificant likelihood is avoided.

" No change to existing DID is proposed.

All equipment, as designed, is expected to be available and to continue to func-tion with high probability.

• The proposed change is documented in the UFSAR, Chapter 6. No changes are proposed to any high-risk equipment.

In addition to the items listed above, the fol-lowing also support consistency with the key principles of risk-informed regulation and NRC staff expectations:

" The integrity of the Class 1 welds, pip-ing, and components are maintained at a high level of reliability through the ASME Section XI inspection program;

• The materials stored in Contaimnent, especially any transient lead, should be stored as required by Wire [72]. In ad-dition, plant transients are monitored in the Transient Cycle Counting Limits Program [48];

" The structural integrity and cleanli-ness of the Containment Sump Strain-ers is monitored prior to leaving the containment [40, 14]. In particular, any condition noted that would result in direct passage of debris is evaluated through the Station Corrective Action Program

[2] and repaired as neces-sary prior to Containment closeout. The PRA is maintained to reflect the as-built, as-operated plant as described in the STPNOC UFSAR, Section 13.7.2.3 to reflect the current plant design not 26 NOC-AE-1 3002954 6 DOCUAIIENTATION 6.3 Submittal Documentation to exceed every 36 months and to re-flect the equipment performance (com-prehensive data update) not to exceed 60 months. Unless major modifications are made to the containment design or insulation design, no changes should be required to the PRA analysis docu-mented in this licensing submittal;

" Information to be provided as part of the plants LB (e.g., FSAR, technical specifications licensing condition);

" The GSI-191 PRA analysis is not used to enhance or modify safety-related func-tions of SSCs. The STPNOC GSI-191 PRA analysis is controlled under the ex-isting STPNOC PRA application analy-sis and assessment process [3]; and

" There are no other changes to the exist-ing requirements to any systems, struc-tures or components as a consequence of this licensing basis change.

The program used to develop the results of the license basis change included an in-dependent critical peer review oversight pro-cess requiring quarterly reporting and criti-cal review question resolution. A summary of Independent Oversight activities and obser-vations is addressed in Section 7 of this docu-ment. More details including Oversight com-ments and follow-up resolutions are available upon request (Independent Technical Over-sight, Quarterly Reports [28, 29, 30, 31]).

As discussed on Page vii, minimal changes were made to the STPNOC PRA such that a new peer review would not be required.

Although detailed models of post-LOCA be-havior are included in the risk analysis, the models are not embedded in the PRA. In-stead, detailed models of post-LOCA behav-ior are solved in an uncertainty quantifica-tion framework outside of the PRA and the results are supplied to the PRA as discrete probability distributions. In this way, con-tributions of specific issues raised in GSI-191 are encapsulated in familiar models and are therefore more easily scrutinized and under-stood, especially by investigators more famil-iar with the engineering models of behavior.

Since much of the previous investigation into the issues raised in GSI-191 was not based on risk methodologies, the STPNOC GSI-191 analysis method is expected to be familiar to the majority of previous GSI-191 investi-gators.

STPNOC's PRA complies with Regulatory Guide 1.200, Revision 1, however; it does not comply with Regulatory Guide 1.200, Revi-sion 2 with respect to Fire PRA and Seismic PRA requirements. Even though STPNOC's PRA contains both Fire and Seismic PRAs, they do not meet all the standards require-ments in the current ASME/ANS RA-S-2009 PRA Standard, as endorsed by RG 1.200, Rev. 2, at a Capability Category II level.

PRA model changes since the peer review are detailed in Volume 4 but are minimal. The Findings and Observations from the peer re-view are also reviewed in Volume 4.

STPNOC's PRA remains technically ade-quate to evaluate and quantify the risk as-sociated with the concerns raised in GSI-191.

GSI-191 is concerned with LOCA events and these events are explicitly modeled in the STPNOC PRA. STPNOC's PRA does meet Regulatory Guide 1.200, Revision 2 at Ca-pability Category II for LOCA events. For the risk-informed GSI-191 methodology de-scribed in this study, the technical rigor pro-vided to the PRA exceeds that performed in PRAs used today and is technically more than adequate to perform a risk-informed ap-plication meeting RG1.174 guidance.

27 NOC-AE-13002954 7 INDEPENDENT TECHNICAL OVERSIGHT 7

Independent Techni-cal Oversight Since January 2012, Soteria Consultants, LLC (Soteria) has provided Independent Technical Oversight of the STPNOC STP-NOC Pilot Project. STPNOC commissioned the oversight group to help ensure the quality and validity of the research and development undertaken. The main objective of Indepen-dent Technical Oversight has been to per-form an in-depth scientific review of the phe-nomenological models and experiments de-veloped and conducted for the STPNOC Pilot Project.

Soteria's approach included both "active" and "passive" oversight activities. Two mem-bers of Soteria Consultants (Dr. Zahra Mo-haghegh44 and Dr. Seyed Reihani45) inter-acted and collaborated with the technical teams to provide feedback and to offer active oversight services. Since the project involved new research, and because of its multidisci-plinary and integrative nature, it required the oversight group to participate in meet-ings and to follow up on discussions and com-ments with the other team members. Specific areas of concerns and reviews were also dis-cussed with Soteria's associate experts (that is, passive oversight members) including Dr.

Ali Mosleh 46 and Dr. Reza Kazemi47 Soteria was involved in both "informal" and "formal" oversight activities for the STP-44From Janary 2013, Assistant Professor in Nuclear Eng. Department at the University of Illinois at Urbana Champaign.

4 5From January 2013, Research Scientist in Nuclear Eng. Department at the University of Illinois at Urbana Champaign.

46Also, Professor of Mechanical Eng. Depart-ment at the University of Maryland, College Park.

47Also, Operations Research Analyst at the FDA (Individual's opinion and input to this project are his own personal views and do not reflect in any way that of the FDA).

NOC Pilot Project. Examples of informal ac-tivities were: (1) reviewing pre-meeting tech-nical reports and documents related to NRC public meetings and providing comments; (2) providing technical support in develop-ing ACRS presentations, and; (3) participat-ing in brainstorming sessions on diverse tech-nical topical areas with the required follow-up on the proposed ideas. Some of the for-mal Oversight activities included: (1) partic-ipating in weekly technical team teleconfer-ences and providing feedback; (2) participat-ing in monthly technical meetings and pro-viding comments, and; (3) developing four Oversight Quarterly Reports [28, 29, 30, 31].

In order to make the review process more thorough and to enhance the effects and effi-ciency of having an oversight function for the STPNOC Pilot Project, Soteria asked the tech-nical team members to provide responses re-garding each of Soteria's specific comments.

The main objectives of Oversight Quar-terly Reports were to: (1) analyze, the re-sponses that Soteria had received from the members of the teams regarding oversight comments. The teams' responses were doc-umented along with Soteria's responses, res-olutions, and feedback on the unresolved is-sues; (2) provide an up-to-date report of So-terias activities during the quarter; (3) com-municate additional comments based on the review of recent reports and participation in the technical meetings and teleconferences, and; (4) facilitate the interaction and col-laboration of the oversight team with mem-bers of other technical teams. The Over-sight Quarterly Reports contributed to the progress of the project by addressing critical peer review of the documents and by high-lighting an up-to-date elaboration of areas of concern that required further investigation from the technical teams.

From Soteria's perspective, the STPNOC Pilot Project is an outstanding blend of ad-vanced and conventional methods that not 28 NOC-AE-1 3002954 7 INDEPENDENT TECHNICAL OVERSIGHT only contributes towards the closure of the GSI-191 issues, but also makes a significant contribution to the formal incorporation of underlying physical failure mechanisms of certain post-LOCA events into PRA. Soteria's oversight activities have concluded that the STPNOC Pilot Project, having a well-designed combination of probabilistic and determinis-tic methodologies, has made important con-tributions to the closure of GSI-191 issues.

The detailed technical results of Soteria's critical reviews are available in the four Over-sight Quarterly Reports [28, 29, 30, 31].

In addition to reviewing the various work-ing documents and analyses in FY 2012, Soteria has been reviewing Volumes 1, 2, 3, and 4 of the submittals and their sup-porting documents. The members of tech-nical teams (that is, PRA GSI-191 Analy-sis & Methodology Implementation; GAMI, Corrosion/Head Loss Experiments; CHLE, CASA Grande, Thermal Hydraulics; TH, Un-certainty Quantification; UQ, and Jet For-mation; JF) have responded to and imple-mented the majority of Soteria's comments.

Some specific comments (e.g., related to ver-tical head-loss tests and blender bed tests, etc.) have not yet been implemented, mainly due to time and budget constraints. The plan is to address these along with NRC's addi-tional comments in FY 2013. The four Over-sight Quarterly Reports [28, 29, 30, 31] in-clude the resolution status of Soteria's com-ments.

Because of the large-scale nature of the STPNOC Pilot Project, Soteria believes that follow-up research, implementation, and ex-periments in FY 2013 would certainly im-prove the quality and validity of the project.

During FY 2013, Soteria team members, who have joined the academic staff of the University of Illinois at Urbana Champaign, will continue the technical oversight function during ongoing technical work and the NRC review process.

29 NOC-AE-1 3002954 8 ACRONYMS 8

Acronyms CAD Computer Aided Design a computer aided design model STPNOC is using to rep-resent the containment buildings that includes piping welds and insulation details in order to help accurately assess ablated materials following an hypothesized LOCA.

CASA Grande Containment Accident Stochastic Analysis (CASA) and Grande referring to the STPNOC large, dry containment the framework used to perform the computer-ized uncertainty quantification (sampling of distributions, propagating uncertainties) to develop basic events that address the issues raised in GSI-191.

ccdf Complementary cumulative distribution function as normally defined: F(x) = 1 -

f %. f(t)dt.

CDF Core Damage Frequency STPNOC calculates core damage frequency using the STP-NOC PRA.

cdf Cumulative distribution function as normally defined: F(x) = f o f(t)dt.

CLB Cold Leg Break is a failure in the RCS piping between the steam generator cold leg nozzle and the reactor vessel cold leg nozzle.

CS Containment Spray System a part of the STP Engineered Safety Systems and consists of three trains (Trains A, B, and C). Only two Containment Spray trains are required to meet the system's spray flow requirements. The STPNOC Containment spray does not pass through the RHR heat exchanger.

DEGB Double-Ended Guillotine Break is a hypothetical condition that can be realized mathematically whereby a pipe instantaneously shears around its circumference and in the same instantaneous time, completely offsets such that the jets from each end of the shear plane can't interfere with each other.

DID Defense-in-Depth is the design concept that includes redundant and/or multiple bar-riers to a particular consequence.

ECCS Emergency Core Cooling System part of the STPNOC engineered safety features.

Foid Void Fraction is the liquid vapor fraction just downstream of the ECCS strainer.

GL 2004-02 NRC Generic Letter 2004-02 was issued in response to the concerns raised in GSI-191 for PWRs.

GSI-191 Generic Safety Issue 191 the NRC Generic Safety Issue number 191.

HHSI High Head Safety Injection a part of the ECCS. The STPNOC plants have three HHSI trains (Trains A, B, and C) that can provide ECCS flow at pressures up to around 1600 psi.

HLB Hot Leg Break is a failure in the RCS piping between the steam generator hot leg nozzle and the reactor vessel hot leg nozzle including the Pressurizer (D Loop).

30 NOC-AE-1 3002954 8 ACRONYMS LB Licensing Basis is the collection of commitments and requirements that licensee makes to the regulatory authority (in this case, the NRC) over the course of time.

LERF Large Early Release Frequency STPNOC calculates large early release frequency using the STPNOC PRA.

LHS Latin Hypercube Sampling is a method used in uncertainty quantification to sample a distribution as efficiently as possible while preserving the variability.

LHSI Low Head Safety Injection part of the ECCS. The STPNOC plants have three LHSI trains (Trains A, B, and C) that can provide ECCS flow at pressures up to around 400 psi. The LHSI train is the only ECCS train that uses the RHR heat exchangers for decay heat removal.

LLOCA Large Break Loss of Coolant Accident a hypothetical instantaneous pressure boundary failure that is defined for STPNOC as greater than 6 inch equivalent di-ameter.

LOCA Loss of Coolant Accident a hypothetical instantaneous pressure boundary failure.

MLOCA Medium Break Loss of Coolant Accident a hypothetical instantaneous pressure boundary failure that is defined for STPNOC as greater than 2 inch equivalent diameter but less than 6 inch equivalent diameter.

NLHS Nonuniform Latin Hypercube Sampling is the stratified LHS scheme that divides the cumulative probability into into unequal segments that are (each) randomly sampled to form a sample design matrix.

NPSHA Net Positive Suction Head Available is the total pressure at the eye of the pump impeller. As long as the net positive suction head available is higher then the net positive suction required, the pump will have sufficient pressure at the impeller inlet to operate without cavitation.

NPSHR Net Positive Suction Head Required is the total pressure at the eye of the pump impeller required for the pump to operate properly, without excessive cavitation.

NSSS Nuclear Steam Supply System the nuclear reactor, piping, pumps, steam genera-tors, pressurizer, and auxiliary equipment associated with operation and control of the reactor system.

STPNOC Pilot Project STPNOC Risk-Informed GSI-191 Closure Pilot Project. The NRC works with licensees as they develop methods to address new regulatory ap-proaches. STPNOC requested and was granted Pilot Project status for the methodol-ogy for closing GSI-191 using Option 2b Pbuckle Strainer mechanical failure limit is the differential pressure across the ECCS strainers at which they are analyzed to suffer mechanical damage. The failure limit is approxi-mately 9.35 ftWC.

pdf Probability density function is a differential function having units of "# per unit x" as in, "probability per inch of break size."

31 NOC-AE-1 3002954 8 ACRONYMS PHSA Probabilistic Seismic Hazard Analysis is the probabilistic study of seismic events on systems, structures, and components to obtain failure likelihoods.

PRA Probabilistic Risk Assessment the STPNOC PRA is the platform for all quantitative risk assessment licensing activities at STPNOC. The current model (Model of Record) is Revision 7.

PWR Pressurized Water Reactor. The STPNOC site consists of two, four loop, approxi-mately 3800 MWth, Westinghouse Nuclear Steam Supply System reactors.

PWSCC Primary Water Stress Corrosion Cracking is a degradation mechanism for certain types of weld materials, especially Alloy 600.

RCFC The Reactor Containment Fan Coolers a part of the STP Engineered Safety Systems and consist of three trains (Trains A, B, and C).

RCS Reactor Coolant System the STPNOC reactor coolant system is a four loop Westing-house design RG1.174 Regulatory Guide 1.174 is a regulatory guidance document that describes the overall methodology to quantify risk using the PRA together with deterministically-based criteria to evaluate the acceptability of a particular change. The quantitative risk measures are CDF and LERF. The risk is deemed to be "very small" when the change increases CDF less than 10-6 and the LERF less than 10-7.

RHR Residual Heat Removal System a shutdown cooling system consisting of three inde-pendent trains. The RHR heat exchangers are shared with the LHSI train. If the LHSI train is using the heat exchanger for that train, the RHR train must be secured and vice versa.

RM1II Reflective Metal Insulation is a fitted, rigid insulation that uses metal radiation heat shields and dead air space to reduce heat loss.

RMS Records Management System is the STPNOC document storage and retrieval system meeting the requirements of Regulatory Guide 1.33, Revision 2, Quality Assurance Program Requirements (Operation).

RMTS Risk Managed Technical Specifications the allowed outage time for risk significant equipment derived from the configuration risk during the outage time.

RWST Refueling Water Storage Tank the STPNOC reactor water storage tank holds ap-proximately 500,000 gallons of water borated to the all rods out, xenon free boron concentration, approximately 2800 ppm.

SLOCA Small Break Loss of Coolant Accident a hypothetical instantaneous pressure boundary failure that is defined for STPNOC as less than 2 inch equivalent diame-ter and greater than 1/2 inch equivalent diameter.

STP South Texas Project electric generating station is the two commercial nuclear electric generating units located near Wadsworth, TX.

32 NOC-AE-1 3002954 8 ACRONYMS STPNOC The STP Nuclear Operating Company is the organization responsible for the safe and efficient operation of the South Texas Project electric generating station.

ZOI Zone of Influence refers to the enclosed volume where damage to materials is hypoth-esized or assumed to occur. The damage assumed is from the energetic jet associated with the hypothesized instantaneous failure of Class 1 piping in the containment build-ing.

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Risk-Informed Resolution of GSI-191 at South Texas Project.

Technical Report Revision 0, South Texas Project, Wadsworth, TX.

[25] Liming, J. and E. Kee (2002, April). Integrated Risk-Informed Asset Management for Commercial Nuclear Power Stations. In Proceedings of the 10th International Conference on Nuclear Engineering, Number 10-22033 in ICONE.

[26] Liming, J. K., E. J. Kee, and G. G. Young (2003, April). Practical application of deci-sion support metrics for power plant risk-informed asset management. In Proceedings of the 11th International Conference on Nuclear Engineering, April 20-23, Tokyo, JAPAN.

[27] Mohaghegh, Z. (2009, March). Socio-Technical Risk Analysis. VDM Verlag.

Discusses how cultural effects such perception of safety of workers, manage-ment communication of safety, etc., can be integrated into a "classic PRA".

[28] Mohaghegh, Z. and S. A. Reihani (2012a, April 14).

1st Oversight Quarterly Report for STP Risk-Informed Approach to NRC Generic Safety Issue 191 (GSI-191). Quarterly Oversight Report 1, SOTERIA Consultants, LLC, Boston, MA.

[29] Mohaghegh, Z. and S. A. Reihani (2012b, July 11).

2 nd Oversight Quarterly Report for STP Risk-Informed Approach to NRC Generic Safety Issue 191 (GSI-191). Quarterly Oversight Report 2, SOTERIA Consultants, LLC, Boston, MA.

[30] Mohaghegh, Z. and S. A. Reihani (2012c, October 14). 3rd Oversight Quarterly Report for STP Risk-Informed Approach to NRC Generic Safety Issue 191 (GSI-191). Quarterly Oversight Report 3, SOTERIA Consultants, LLC, Boston, MA.

[31] Mohaghegh, Z. and S. A. Reihani (2013, January 24). 4th Oversight Quarterly Report for STP Risk-Informed Approach to NRC Generic Safety Issue 191 (GSI-191). Quarterly Oversight Report 4, SOTERIA Consultants, LLC, Boston, MA.

[32] Moiseytseva, V. E. and E. Kee (2004, April). Using RIAM for Optimizing Reactor Vessel Head Leak Failure Mode Maintenance Strategies. In Proceedings of the 12th In-ternational Conference on Nuclear Engineering, Number 12-49376 in ICONE.

[33] National Research Council (1997). Review of Recommendations for Probabilistic Seis-mic Hazard Analysis:Guidance on Uncertainty and Use of Experts. Panel on Seismic Hazard Evaluation, Committee on Seismology, Commission on Geosciences, Environment, 36 NOC-AE-1 3002954 9

REFERENCES and Resources, National Research Council. Washington, DC: The National Academies Press.

[34] NEI (2004, May). Pressurized Water Reactor Sump Performance Evaluation Method-ology. Technical Report 04-07, Nuclear Energy Institute, 1776 I Street, Washington, DC.

[35] NEI (2009).

ECCS Recirculation Performance Following Postulated LOCA Event:

GSI-191 Expected Behavior. White Paper.

[36] Nuclear Regulatory Commission (2007, January). AN APPROACH FOR DETER-MINING THE TECHNICAL ADEQUACY OF PROBABILISTIC RISK ASSESSMENT RESULTS FOR RISK-INFORMED ACTIVITIES. Regulatory Guide 1.200, Nuclear Reg-ulatory Commission, WVashington, DC.

[37] Nuclear Regulatory Commission (2009, March). AN APPROACH FOR DETERMIN-ING THE TECHNICAL ADEQUACY OF PROBABILISTIC RISK ASSESSMENT RE-SULTS FOR RISK-INFORMED ACTIVITIES. Regulatory Guide 1.200, Nuclear Regu-latory Commission, Washington, DC.

[38] Nuclear Regulatory Commission (2011, May). REGULATORY GUIDE 1.174 An Ap-proach for Using Probabilistic Risk Assessment In Risk-Informed Decisions On Plant-Specific Changes to the Licensing Basis, Revision 2. Regulatory Guide 1.174, Nuclear Regulatory Commission, Washington, DC.

[39] Nuclear Regulatory Commission (2012, July).

CLOSURE OPTIONS FOR GENERIC SAFETY ISSUE - 191, ASSESSMENT OF DEBRIS ACCUMULATION ON PRESSURIZED-WATER REACTOR SUMP PERFORMANCE. Letter (SECY) 12-0093, Nuclear Regulatory Commission, Washington, DC.

[40] Page, M. (2010). Initial Containment Inspection to Establish Integrity. South Texas Project Plant Procedure, OPSP03-XC-0002.

[41] Popova, E. and A. Galenko (2011, December). Uncertainty Quantification (UQ) Meth-ods, Strategies, and Illustrative Examples Used for Resolving the GSI-191 Problem at South Texas Project. Technical Report Revision 0, The University of Texas at Austin, Austin, TX.

[42] Popova, E. and D. Morton (2012, May). Uncertainty modeling of LOCA frequencies and break size distributions for the STP GSI-191 resolution. Technical report, The University of Texas at Austin, Austin, TX.

[43] Rodgers, S. S., C. D. Betancourt, E. Kee, F. Yilmaz, and P. Nelson (2011). Integrated Power Recovery Using Markov Modeling. ASME Journal of Engineering for Gas Turbines and Power Volume 133.

[44] Rodgers, S. S. and R. F. Dunn (2012, August 30). PRA Reference Model Update From STP Rev. 6 to STP Rev. 7. Procedure OPGP01-ZA-0305, Rev. 9 STI 33590701, STPNOC Risk Management, STPNOC, PO Box 289, Wadsworth, TX 77414.

37 NOC-AE-1 3002954 9 REFERENCES References the documentation set for the STPNOC PRA Revision 7 released in 2012.

[45] Rosenburg, S. (2011, January). PUBLIC MEETING WITH THE NUCLEAR EN-ERGY INSTITUTE ON STATUS AND PATH FORWARD TO RESOLVE GSI-191.

Memorandum.

[46] Sande, T., K. Howe, and J. Leavitt (2011, October). Expected Impact of Chemical Effects on GSI-191 Risk-Informed Evaluation for South Texas Project. White Paper ALION-REP-STPEGS-8221-02, Revision 0, Jointly, Alion Science and Technology and Univiersity of New Mexico, Albuquerque, NM.

White paper developed in anticipation of the STPNOC GSI-191 chemical ef-fects experimental program. Actual amounts of corrosion materials in the plant and preliminary hypotheses are developed.

[47] Schneider, E., J. Day, and W. Gurecky (2011, December). Simulation Modeling of Jet Formation Progress Report, August - December 2011. Internal Report Revision 0, University of Texas at Austin, Austin, TX.

[48] Shojaei, S. (2010). Transient Cycle Counting Limits. South Texas Project Plant Procedure, 0PEP02-ZE-0001.

[49] Singal, B. K. (2011a, June). FORTHCOMING CONFERENCE CALL WITH STP NUCLEAR OPERATING COMPANY (TAC NOS. ME5358 and ME5359). Memoran-dum.

Initial meeting on the STPNOC overall approach to risk-informed slolution to GSI-191. Overviews of the PRA approach, treatment of LOCA frequen-cies, thermal-hydraulics, jet formation, and downstream effects. The licensing strategy was presented as well (DRAFT Exemption request language).

[50] Singal, B. K. (2011b, July). FORTHCOMING CONFERENCE CALL WITH STP NUCLEAR OPERATING COMPANY (TAC NOS. ME5358 and ME5359). Memoran-dum.

Follow-up discussion to the public meeting held on June 2, 2011, between STP Nuclear Operating Company (STPNOC) and the U.S. Nuclear Regula-tory Commission (NRC) staff to discuss Generic Safety Issue (GSI) 191, "As-sessment of Debris Accumulation on PWR (Pressurized-Water Reactor])Sump Performance." At the June 2 nd meeting, STPNOC discussed a risk-informed GSI-191 resolution option approach regarding Texas Project, Units 1 and 2.

[51] Singal, B. K. (2011c, May). FORTHCOMING MEETING WITH STP NUCLEAR OPERATING COMPANY (TAC NOS. ME5358 and ME5359). Memorandum.

[52] Singal, B. K. (2011d, August). FORTHCOMING MEETING WITH STP NUCLEAR OPERATING COMPANY (TAC NOS. ME5358 and ME5359). Memorandum.

38 NOC-AE-1 3002954 9 REFERENCES Overview of the CASA Grande calculation flow starting with a loss-of-coolant accident to sump screen performance. Discussion of computational fluid dy-namics verification plans.

[53] Singal, B. K. (2011e, October). FORTHCOMING PUBLIC MEETING VIA CONFER-ENCE CALL WITH STP NUCLEAR OPERATING COMPANY (TAC NOS. ME5358 and ME5359). Memorandum.

Meeting to discuss the topic of Loss-of-Coolant Accident (LOCA) Initiating Event Frequencies and Uncertainties. These discussions were related to the initial approach which we have come to refer to as the "bottom up" approach.

[54] Singal, B. K. (2011f, September). FORTHCOMING PUBLIC MEETING VIA CON-FERENCE CALL WITH STP NUCLEAR OPERATING COMPANY (TAC NOS.

ME5358 and ME5359). Memorandum.

[55] Singal, B. K. (2011g, November). FORTHCOMING PUBLIC MEETING VIA CON-FERENCE CALL WITH STP NUCLEAR OPERATING COMPANY (TAC NOS.

ME5358 and ME5359). Memorandum.

Initial plans and protocol for integrated chemical effects testing. This testing described would be performed in 2012.

[56] Singal, B. K. (2012a, November 27).

FORTHCOMING PUBLIC MEETING VIA CONFERENCE CALL WITH STP NUCLEAR OPERATING COMPANY (TAC NOS.

ME5358 and ME5359). Memorandum.

Chemical effects testing, Coatings, Texas A&M bypass test follow up, Bump-up factor, and non-chemical head loss testing.

[57] Singal, B. K. (2012b, January). FORTHCOMING PUBLIC MEETING VIA CON-FERENCE CALL WITH STP NUCLEAR OPERATING COMPANY (TAC NOS.

ME7735 and ME7736). Memorandum.

[58] Singal, B. K. (2012c, February 2).

FORTHCOMING PUBLIC MEETING VIA CONFERENCE CALL WITH STP NUCLEAR OPERATING COMPANY (TAC NOS.

ME7735 and ME7736). Memorandum.

Chemical Effects testing plan review.

[59] Singal, B. K. (2012d, February 3).

FORTHCOMING PUBLIC MEETING VIA CONFERENCE CALL WITH STP NUCLEAR OPERATING COMPANY (TAC NOS.

ME7735 and ME7736). Memorandum.

[60] Singal, B. K. (2012e, March 29). FORTHCOMING PUBLIC MEETING VIA CON-FERENCE CALL WITH STP NUCLEAR OPERATING COMPANY (TAC NOS.

ME7735 and ME7736). Memorandum.

Follow-up discussion on the topic of Loss-of-Coolant Accident Initiating Event Frequencies and Uncertainties.

39 NOC-AE-1 3002954 9 REFERENCES

[61] Singal, B. K. (2012f, May 31). FORTHCOMING PUBLIC MEETING VIA CONFER-ENCE CALL WITH STP NUCLEAR OPERATING COMPANY (TAC NOS. ME7735 and ME7736). Memorandum.

[62] Singal, B. K. (2012g, August 23). FORTHCOMING PUBLIC MEETING VIA CON-FERENCE CALL WITH STP NUCLEAR OPERATING COMPANY (TAC NOS.

ME7735 and ME7736). Memorandum.

[63] Singal, B. K. (2012h, September 21). FORTHCOMING PUBLIC MEETING VIA CONFERENCE CALL WITH STP NUCLEAR OPERATING COMPANY (TAC NOS.

ME7735 and ME7736). Memorandum.

[64] Spiess, L. (2012). ASME Section XI Inservice Inspection. South Texas Project Plant Procedure, OPSPl1-RC-0015.

[65] Teolis, D., R. Lutz, and H. Detar (2009). PRA Modeling of Debris-Induced Failure of Long Term Cooling via Recirculation Sumps. WCAP 16882, Westinghouse Electric Company, LLC, Pittsburgh, PA.

[66] Thadani, M. (2011, February). FORTHCOMING MEETING WITH STP NUCLEAR OPERATING COMPANY (TAC NOS. ME5358 and ME5359). Memorandum.

[67] Trbovich, J. (2010, November 15). Control of heavy loads. South Texas Project Plant Procedure, OPGP03-ZA-0069.

[68] Tregoning, R., L. Abramson, and P. Scott (2008, April). Estimating Loss-of-Coolant Accident (LOCA) Frequencies Through the Elicitation Process. NUREG/CR 1829, Nu-clear Regulatory Commission, Washngton, DC.

[69] Vaghetto, R. (2013, January).

Core Blockage Thermal-Hydraulic Analysis. South Texas Project Risk-Informed GSI-191 Evaluation, Texas A&M University, College Sta-tion, Texas.

[70] Vietti-Cook, A. L. (2010, December). STAFF REQUIREMENTS - SECY-10-0113 -

CLOSURE OPTIONS FOR GENERIC SAFETY ISSUE-191, ASSESSMENT OF DE-BRIS ACCUMULATION ON PRESSURIZED WATER REACTOR SUMP PERFOR-MANCE. Letter from Annette L. Vietti-Cook to R. W. Borchardt.

[71] Wang, S., E. Kee, and F. Yilmaz (2010, June). Quantification of Conditional Probabil-ity for Triggering Events using Fault Tree Approach. In Proceedings of the Probabilistic Safety Assessment Meeting 2010, Number 10-164 in PSAM.

[72] Wire, C. (2012). Shielding. South Texas Project Plant Procedure, OPRP07-ZR-0004.

[73] Yilmaz, F. and E. Kee (2011, March). Methodology to Rank BOP Components at STP.

In ANS PSA 2011 International Topical Meeting on Probabilistic Safety Assessment and Analysis, Wilmington, NC March 13-17. American Nuclear Society.

[74] Yilmaz, F. and E. Kee (2012a, July 29 - August 2). Return-to-Service Priority determi-nation in RAsCal. In in print, Number 21-15356 in ICONE, Chendu, China. ANS/ASME.

40 NOC-AE-13002954 9

REFERENCES

[75] Yilmaz, F. and E. Kee (2012b, July 29 - August 2). Tier 1 Nuclear Safety Performance Index at STP: Risk Index. In in print, Number 21-15355 in ICONE, Chendu, China.

ANS/ASME.

[76] Yilmaz, F., E. Kee, and R. Grantom (2011, March). Development of Risk Communica-tion Sheet for Daily Operational Focus Meetings at STP. In ANS PSA 2011 International Topical Meeting on Probabilistic Safety Assessment and Analysis, Wilmington, NC March 13-17. American Nuclear Society.

[77] Yilmaz, F., E. Kee, and D. Richards (2009, July). STP Risk Managed Technical Spec-ification Software Design and Implementation. In Proceedings of the 17th Inteirnational Conference on Nuclear Engineering, Number 17-75043 in ICONE.

41 NOC-AE-13002954 Appendices Appendix A is a table with three columns, "Section", "Paragraph Summary", and "Where Addressed" developed to help ensure the requirements of RG1.174 have been addressed in the STPNOC Pilot Project. The first column, "Section", highlights the four elements identi-fied in RG1.174. In an attempt to identify all sub-elements, items that clearly bear on the information needed were pulled out of the text and entered in the column "Paragraph Sum-mary". The "Where Addressed" column primarily refers to the Section in this document (Volume 1) where the requirement is addressed. As mentioned in the Volume 1 Introduc-tion & Background, the numbered sections of Volume 1 correspond to the numbered sections in RG1.174 which should also help in this regard.

Appendix B is a table with four columns, "Topical Area", "NRC-Approved Determinis-tic Methods", "STPNOC Pilot ProjectMethods for 2012 Quantification", and "Comments".

The table is intended to help understand how the engineering analysis supporting the PRA used in the STPNOC Pilot Project relates to the NEI 04-07 [34] recommended models. In particular, the collection of engineering models used in the CASA Grande analysis are item-ized against the recommendations. NEI 04-07. "Topical Area" is the GSI-191 engineering model subject area. "NRC-Approved Deterministic Methods" is the methodology approved by the NRC for the particular topical area (not all topical areas had approved models at the time the STPNOC Pilot Project was completed). "STPNOC Pilot ProjectMethods for 2012 Quantification" is a quick description of the engineering model used in the STPNOC Pilot Project. "Comments" provides information about whether the model is the same (that is, "no difference") or a summary description of how the model adopted differs or in some cases is closely related to the NRC's model choice.

Appendix C is a table having two columns that summarize actions taken over the several years GSI-191 has been of concern. The appendix is provided to help, in some cases, add some specificity to references in the body of the Volume 1 document and, in some cases, to supplement the basis for engineering judgement of DID and safety margin assertions.

42

A Appendix A. Checklist for Regulatory Guide 1.174 Inputs Table 5: Checklist for Regulatory Guide 1.174 Section Paragraph Summary

ý Where addressed Element 1: Define the Identify those aspects of the plants LB that may be affected by the proposed Page 1.

Proposed Change change, including but not limited to rules and regulations, FSAR, technical specifications, licensing conditions, and licensing commitments.

Identify all structures, systems, and components (SSCs), procedures, and ac-Page 1.

tivities that are covered by the LB change being evaluated and should consider the original reasons for including each program requirement Identify all structures, systems, and components (SSCs), procedures, and ac-Prior changes and tivities that are covered by the LB change being evaluated and should consider primary STPNOC pro-the original reasons for including each program requirement cesses bearing on this LB change are summna-rized in Section 1 Identify regulatory requirements or commitments in its LB that it believes are GSJ-191 and Generic overly restrictive or unnecessary to ensure safety at the plant.

Letter 2004-02 overly restrictive based on actual plant analysis.

Identify design and operational aspects of the plant that should be enhanced No additional changes consistent with an improved understanding of their safety significance. Such to the plant are rec-enhancements should be embodied in appropriate LB changes that reflect these ommended beyond enhancements.

the those already implemented. Section 3 continued next page...

z 0

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..continued Section Paragraph Summary Where addressed Identify available engineering studies, methods, codes, applicable plant-specific Overview on Page vii, and industry data and operational experience, PRA findings, and research and Figure 2. Further de-analysis results relevant to the proposed LB change. With particular regard to tails provided in Vol-the plant-specific PRA, the licensee should assess the capability to use, refine, ume 3. The PRA ca-augment, and update system models as needed to support a risk assessment pability is described in of the proposed LB change.

Section 2.3 and further details are provided in Volumes 2 and 4.

Describe the LB change and to outline the method of analysis. The licensee Page 3 should describe the proposed change and how it meets the objectives of the NRCs PRA Policy Statement (Ref. 1), including enhanced decision making, more efficient use of resources, and reduction of unnecessary burden.

Describe the LB change and to outline the method of analysis. The licensee Page 3 should describe the proposed change and how it meets the objectives of the NRCs PRA Policy Statement (Ref. 1), including enhanced decision making, more efficient use of resources, and reduction of unnecessary burden.

Combined Change Re-Licensees may include several individual changes to the LB that have been This section is not ap-quests evaluated and will be implemented in an integrated fashion.

plicable to the STPNOC Pilot Project.

Guidelines for Develop-The changes that make up a CCR should be related to one another.

This section is not ap-ing Combined Change plicable to the STPNOC Requests Pilot Project.

Element 2:

Perform The scope, level of detail, and technical adequacy of the engineering analyses Section 2. Detailed de-Engineering Analysis conducted to justify any proposed LB change should be appropriate for the scription is provided in nature and scope of the proposed change.

Volume 3.

Some proposed LB changes can be characterized as involving the categorization Not applicable to this of SSCs according to safety significance.

LB change.

continued next page...

z0 00o) cIh C

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... continued Section 1 Paragraph Summary Where addressed Evaluation of Defense-Evaluate the proposed LB change with regard to the principles of maintaining Section 2.1 summarizes in-Depth Attributes and adequate defense-in-depth, maintaining sufficient safety margins, and ensuring Defense in Depth and Safety Margins that proposed increases in CDF and risk are small and are consistent with the Safety Margin. The risk intent of the Commissions Safety Goal Policy Statement.

is very small, (Page 5) and well within the Commissioners' safety goal.

Show that the fundamental safety principles on which the plant design was No changes are pro-based are not compromised by the proposed change.

posed to plant design principles as described in Section 2.1.1.1.

Evaluate whether the impact of the proposed LB change (individually and Section 2.1.1.2 cumulatively) is consistent with the defense-in-depth philosophy.

The evaluation should consider the intent of the general design criteria Section 2.1.1.1.

Assess whether the proposed LB change meets the defense-in-depth principle.

Section 2.1.1.2.

Assess whether the impact of the proposed LB change is consistent with the Section 2.1.2 principle that sufficient safety margins are maintained.

Evaluation of Risk Ira-Risk assessment may be used to address the principle that proposed increases Section 2.2.

pact, Including Treat-in CDF and risk are small and are consistent with the intent of the NRCs ment of Uncertainties Safety Goal Policy Statement Impacts of the proposed change on aspects of risk not captured (or inade-Section 2.2.

quately captured) by changes in CDF and LERF should be addressed. For example, changes affecting long-term containment performance would impact radionuclide releases from containment occurring after evacuation and could result in substantial changes to off-site consequences such as latent cancer fatalities.

Technical Adequacy of The scope, level of detail, and technical adequacy of the PRA are to be com-Section 2.3 and Sec-Probabilistic Risk As-mensurate with the application for which it is intended and the role the PRA tion 2.3.1.

sessment Analysis results play in the integrated decision process.

continued next page...

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... continued Section Paragraph Summary Where addressed Both aleatory and epistemic uncertainty should be evaluated. An understand-Section 2.5.3.

ing of the important contributors in the model should be developed.

Acceptance Guidelines Regions are established in the two planes generated by a measure of the base-Section 2.4.

line risk metric (CDF or LERF) along the x-axis, and the change in those met-rics (CDF or LERF) along the y-axis (Figures 4 and 5). Acceptance guidelines are established for each region.

It is recognized that many PRAs are not full scope and PRA information of The scope and technical less than full scope may be acceptable.

adequacy of the STP-NOC PRA is also de-scribed in Section 2.3.3 There are two sets of acceptance guidelines, one for CDF and one for LERF, The STPNOC PRA [44]

and both sets should be used.

evaluates both CDF and LERF.

Both of these metrics are in-cluded in the STPNOC Pilot Project acceptance criteria (Section 2.2).

Comparison of PRA In the context of integrated decision making, the acceptance guidelines should Section 2.5.

results with acceptance not be interpreted as being overly prescriptive. They are intended to provide guidelines an indication, in numerical terms, of what is considered acceptable.

The assumptions made in response to these sources of model uncertainty and Importance measures any conservatism introduced by the analysis approach can bias the results.

are not relied on in the This is of particular concern for the assessment of importance measures with STPNOC Pilot Project respect to the combined risk assessment and the relative contributions of the (Page 24) hazard groups to the various risk metrics.

continued next page...

z0C) mm 00

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continued Section Paragraph Summary Where addressed I

Comparison of the PRA results with the acceptance guidelines must be based on an understanding of the contributors to the PRA results and on the ro-bustness of the assessment of those contributors and the impacts of the uncer-tainties, both those that are explicitly accounted for in the results and those that are not.

Section 2.5. Other con-tributors are captured in epistemic uncertainty as well as adoption of extreme thresholds for

failure, especially in consideration of Boron Precipitation, ECCS strainer differen-tial pressure and core blockage. Appendix B.

See Page 20.

01 The analysis must be done to correlate the sample values for different PRA Section 2.5.1 elements from a group to which the same parameter value applies.

it is important to develop an understanding of the impact of a specific as-Section 2.3.4.1 provides sumption or choice of model on the predictions of the PRA. This is true even an example illustration when the model uncertainty is treated probabilistically, since the probabili-of how the analysis pro-ties, or weights, given to different models would be subjective. The impact vides understanding of of using alternative assumptions or models may be addressed by performing engineering model im-appropriate sensitivity studies or by using qualitative arguments, based on an pacts on the results.

understanding of the contributors to the results and how they are impacted by the change in assumptions or models.The impact of making specific modeling approximations may be explored in a similar manner.

In many cases, the appropriateness of the models adopted is not questioned and these models have become, de facto, the consensus models to use.

Appendix ?? compares models used compared with industry de facto models. Sections 2.5.1 and 2.3 also address model appropriateness.

z0 mm 00 N)C coo MD

.p. rh continued next page...

0

... continued Section Paragraph Summary Where addressed Completeness Uncer-The issue of completeness of scope of a PRA can be addressed for those scope Section 2.5.4 tainty items for 2095 which methods are in principle available, and therefore some understanding of the contribution to risk exists, by supplementing the analysis with additional analysis to enlarge the scope,using more restrictive acceptance guidelines,or by providing arguments that, for the application of concern, the out-of-scope contributors are not significant.

Comparisons with Ac-Comparison with acceptance guidelines.

Section 2.5.5 ceptance Guidelines Integrated decision In making a regulatory decision, risk insights are integrated with considera-Section 2.6 making tions of 2206 defense-in-depth and safety margins.

Element 3: Define Im-Careful consideration should be given to implementation of the proposed Section 3 and Section 1 plementation and Mon-change and the associated performance-monitoring strategies. The primary itoring Program goal of Element 3 is to ensure that no unexpected adverse safety degradation occurs due to the change(s) to the LB.

Element 4:

Submit Requests for proposed changes to the plants LB typically take the form of Section 4 Proposed Change requests for license amendments (including changes to or removal of license conditions), technical specification changes, changes to or withdrawals of or-ders, and changes to programs under 10 CFR 50.54, "Conditions of Licenses" (e.g., quality assurance program changes under 10 CFR 50.54(a)).

Documentation To facilitate the NRC staffs review to ensure that the analyses conducted were Section 6 sufficient to conclude that the key principles of risk-informed regulation have been met, documentation of the evaluation process and findings are to be maintained.

As part of evaluation of risk, licensees should understand the effects of the The STPNOC PRA current application in light of past applications, is maintained current with the plant including application impacts as described in Section 6.3.

z0 M. M 00 0

Mf CD MCD~

40,41

B Appendix B. NEI 04-07 Comparison Table 6: Comparison of NEI 04-07 recommended engineering models with the models implemented in the STPNOC Pilot Project Topical Area NRC-Approved Determinis-STPNOC Pilot Project Comments tic Methods Methods for 2012 Quantifi-cation Debris Generation Use spherical or hemispherical Use spherical or hemispherical No difference ZOI ZOI 17D ZOI for Nukon and 17D ZOI for Nukon and No difference Thermal-Wrap Thermal-Wrap 28.6D ZOI for Microtherm 28.6D ZOI for Microtherm No difference 4D ZOI for qualified coatings 4D ZOI for qualified coatings No difference Truncate ZOI at walls Truncate ZOI at walls No difference 4-category size distribution for Alion proprietary 4-category size Alion 4 category size distribution fiberglass debris including fines, distribution methodology (con-methodology previously accepted small pieces, large pieces, and in-sistent with guidance in SER ap-by NRC for deterministic evalu-tact blankets pendices) ations 100% fines for Microtherm debris 100% fines for Microtherm debris No difference 100% fines (10/t) for qualified 100% fines (10p) for qualified No difference coatings debris coatings debris 100% failure for all unqualified Partial failure of unqualified New methodology documented coatings debris coatings based on available data.

in Volume 3.

Time-dependent failure of un-qualified coatings based on avail-able data.

continued next page...

z0 00 K) C:

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continued Topical Area NRC-Approved Determinis-STPNOC Pilot Project Meth-Comments tic Methods ods for 2012 Quantification Unqualified coatings fail as 10p Unqualified coatings fail in a Similar methods previously ac-particles if the strainer is fully size distribution based on coat-cepted by NRC for deterministic covered or as chips if a fiber bed ing type and available data.

evaluations would not be formed.

Plant-specific walkdowns re-STP-specific walkdown used to No difference quired to determine latent debris determine latent debris quantity quantity Latent debris consists of 85%

Latent debris consists of 85%

No difference dirt/dust and 15% fiber dirt/dust and 15% fiber Debris 'Tr-ansport Logic tree approach to analyz-Logic tree approach to analyz-No difference ing transport phases: blowdown, ing transport phases: blowdown, washdown, pool fill, recircula-washdown, pool fill, recircula-tion, and erosion tion, and erosion All large pieces and a portion of Fines transport proportional to Similar methods previously ac-small pieces are captured when containment flow, grating and cepted by NRC for deterministic blowdown flow passes through miscellaneous obstructions cap-evaluations.

grating.

ture some small and large pieces.

100% washdown of fines, limited 100% washdown of fines. Credit Includes some new methodology credit for hold-up of small pieces, for hold-up of some small piece documented in Volume 3.

and 0% washdown of large pieces debris on concrete floors and through grating grating. 0% washdown of large pieces through grating.

Pool fill transport to inactive Pool fill transport to inac-Similar methods previously ac-cavities must be limited to 15%

tive cavities is less than 15%.

cepted by NRC for deterministic unless sufficient justification can Methodology is based on expo-evaluations.

be made nential equation with uniform mixing of fines.

continued next page...

z 0

mm 00

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CAC

... continued Topical Area NRC-Approved Determinis-STPNOC Pilot Project Meth-Comments tic Methods ods for 2012 Quantification CFD refinements are appropriate Recirculation transport based on Methodology for CFD modeling for recirculation transport, but a conservative CFD simulations and recirculation transport anal-blanket assumption that all de-developed for the deterministic ysis previously accepted by NRC bris is uniformly distributed is STP debris transport calcula-for deterministic evaluations.

not appropriate.

tion. All debris was not assumed to be uniformly distributed.

90% erosion should be used for Probability distribution with a Values are relatively close to the non-transporting pieces of un-range of less than 10% erosion experimentally determined 10%

jacketed fiberglass in the recircu-based on Alion testing.

erosion value previously accepted lation pool unless additional test-by the NRC for deterministic ing is performed to justify a lower evaluations.

fraction.

1% erosion of small or large 1% erosion of small or large No difference.

pieces of fiberglass held up in up-pieces of fiberglass held up in up-per containment, per containment.

Minimal previous analysis on Time-dependent transport evalu-Several aspects of the time-time-dependent transport.

ated for pool fill, washdown, re-dependent transport are new en-circulation, and erosion.

gineering models documented in Volume 3.

Chemical Effects Corrosion and dissolution of met-Corrosion and dissolution of met-Several aspects of the corro-als and insulation in contain-als and insulation in containment sion and dissolution models axe ment is a function of tempera-is a function of temperature, pH, new engineering models as docu-ture, pH, and water volume. Ac-water volume, and pool chem-mented in Volume 3.

cepted model is WCAP-16530-istry. New model being developed NP.

for STP conditions.

continued next page...

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continued Topical Area NRC-Approved Determinis-STPNOC Pilot Project Meth-Cornments tic Methods ods for 2012 Quantification 100% of material in solution will Some material in solution may New engineering model docu-precipitate.

not precipitate depending on the mented in Volume 3.

solubility limit of the precipitate.

Precipitates can be simulated us-Precipitates are much smaller New engineering model docu-ing the surrogate recipe provided and more benign than WCAP mented in Volume 3.

in WCAP-16530-NP.

surrogate.

Strainer Head Loss Perform plant-specific head Modify the NUREG/CR-6224 Several aspects of the engineer-loss testing of the bounding correlation to address old ACRS ing models are new as docu-scenario(s) with a prototype comments and STP-specific con-mented in Volume 3.

strainer module.

ditions so that head loss can be evaluated at the full range of sce-narios.

Address chemical effects head Address chemical effects head New engineering model docu-loss using WCAP-16530-NP sur-loss with a simple bump-up fac-mented in Volume 3.

rogates in prototype strainer tor similar to the 2011 quantifica-testing.

tion using the CHLE testing that has been performed so far to jus-tify the conservatism.

Minimum fiber quantity equiva-Minimum fiber quantity equiva-No difference lent to 1/16 inch debris bed on lent to 1/16 inch debris bed on the strainers is required to form the strainers is required to form a thin bed.

a thin bed.

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... continued Topical Area NRC-Approved Determinis-STPNOC Pilot Project Meth-Comments tic Methods ods for 2012 Quantification Bounding strainer head loss corn-Time-dependent strainer head Similar engineering model as pared to bounding NPSH margin loss compared to time-dependent documented in Volume 3.

and bounding structural margin NPSH margin and bounding to determine whether the pumps structural margin to determine or strainer would fail.

whether the pumps or strainer would fail.

Air Intrusion Release of air bubbles at the Release of air bubbles at the No difference strainer calculated based on strainer calculated based on the water temperature, submer-the water temperature, submer-gence, strainer head loss, and gence, strainer head loss, and flow rate.

flow rate.

NPSH margin adjusted based on NPSH margin adjusted based on No difference the void fraction at the pump in-the void fraction at the pump in-let let Void fraction at pumps compared Void fraction at pumps compared No difference.

to a steady-state void fraction to a steady-state void fraction of 2% to determine whether the of 2% to determine whether the pumps would fail.

pumps would fail.

Debris Penetration Perform plant-specific fiber pen-Develop a fiber penetration cor-New engineering model Docu-etration testing of the bound-relation as a function of strainer mented in Volume 3.

ing scenario(s) with a prototype flow rate and fiber accumulation strainer module.

based on a series of penetration tests.

100%

penetration of trans-100%

penetration of trans-No difference.

portable particulate and chemi-portable particulate and chemi-cal precipitates.

cal precipitates.

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• .. continued Topical Area NRC-Approved Determinis-STPNOC Pilot Project Meth-Comments tic Methods ods for 2012 Quantification Ex-Vessel Downstream Evaluate ex-vessel wear and clog-Evaluate ex-vessel wear and clog-No difference.

Effects ging based on the methodology in ging based on the methodology in WCAP-16406-P WCAP-16406-P In-Vessel Downstream Compare fiber quantity on core Use RELAP5 simulations to New engineering model docu-Effects to bounding 15 g/FA limit based show that cold leg SBLOCAs and mented in Volume 3.

on WCAP-16793-NP.

all hot leg LOCAs would not go to core damage with full block-age at the base of the core. Use WCAP-17057-P tests with condi-tions closer to the STP to justify an appropriate fiber limit on the core.

Evaluate reduced heat transfer Evaluate reduced heat transfer No difference.

due to deposition on fuel rods us-due to deposition on fuel rods us-ing LOCADM software.

ing LOCADM software.

Boron Precipitation No currently accepted methodol-Evaluate fiber accumulation on New engineering model docu-ogy.

the core for cold leg breaks dur-mented in Volume 3.

ing cold leg injection. Assume that 7.5 g/FA of fiber is sufficient to form a debris bed that would prevent natural mixing between the core and lower plenum. As-sume failure due to boron pre-cipitation if this quantity arrives prior to hot leg switchover.

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Appendix C. Defense-in-Depth and Safety Mar-gin The staff memorandum [39] for options to close GSI-191 concluded that debris could clog the containment sump strainers in PWRs, leading to the loss of net positive suction head for the ECCS and CSS pumps. The NRC issued GL 2004-0248, "Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors" (ADAMS Accession No. ML042360586), dated September 13, 2004, requesting that licensees address the issues raised by GSI-191. GL 2004-02 was focused on demonstrating compliance with 10 CFR 50.46.

Licensees implemented compensatory measures in response to Bulletin 2003-01, "Poten-tial Impact of Debris Blockage on Emergency Sump Recirculation at Pressurized-Water Reactors" (ADAMS Accession No. ML031600259) dated June 9, 2003, and GL 2004-02 to address the potential for sump strainer blockage. Additional compensatory measures could be developed by licensees to specifically address in-vessel blockage. PWRs have instrumen-tation to monitor core water levels and temperatures following a LOCA and operating procedures to initiate hot-leg injection, which may provide an alternate flowpath that by-passes core inlet blockage. For these reasons and others documented in GL 2004-02 that are still applicable, continued operation is justified for each of the recommended options and schedules to resolve GSI-191.

Most licensees implemented mitigative measures for suction strainer clogging following Bulletin 2003-01, "Potential Impact of Debris Blockage on Emergency Sump Recircula-tion at Pressurized-Water Reactors" (ADAMS Accession No. ML031600259) dated June 9, 2003, and GL 2004-02, "Potential Impact of Debris Blockage on Emergency Recircula-tion During Design Basis Accidents at Pressurized-Water Reactors" (ADAMS Accession No. ML042360580) dated September 13, 2004. The staff would expect these measures to be in place while suction strainer performance is resolved, if applicable, and the licensee to implement additional mitigative measures for in-vessel effects.

Plant hardware modifications developed in response to issues identified in GL 2004-02 are installed in STP Units 1 and 2 and are supporting compliance with the regulatory re-quirements for long term cooling following a design basis loss of coolant accident. Similarly, implementation is complete for STPNOC plant administrative procedures and processes needed to support the GL 2004-02 hardware modifications and to support the current as-sumptions, initial conditions and conclusions of GL 2004-02 related evaluations, including the current evaluations of design basis accident debris generation and transport, sump strainer performance, impact of chemical effects and downstream effects of debris. Substantial plant-specific testing that supports assumptions and corresponding conclusions contained in the GL 2004-02 evaluations for STP has been performed.

Since hardware, operating procedures and administrative controls required to support actions taken in response to issues identified in GL 2004-02 are already implemented at STP, STPNOC has high confidence that if an accident of the type described in CL 2004-02 were to occur at STP, plant systems and plant operators would respond in a manner consistent with the intent of the GL 2004-02 corrective actions, including conformance with the regulatory 48NRC Generic Letter 2004-02 C1 NOC-AE-1 3002954 requirements listed in GL 2004-02.

The following table itemizes the historical (that is, prior to the STPNOC Pilot Project)

STP responses related to concerns raised in GSI-191.

C2

Table 7: Historical STP responses related to concerns raised in GSI-191 included actions taken, site-specific design features, procedures, and programs that provide defense-in-depth measures (preventive, mitigative, and protective) and safety margin. References to letters, procedures and other guidance documents are also provided.

Issue or Reference Summary GL 2004-02 response Modifications, mitigative measures, compensatory measures, and/or favorable conditions are in effect at STP, Units 1 and 2, minimizing the risk of degraded ECCS and CS functions.

The three train-specific original design ECCS strainers for both STP units have been replaced with new design strainers. The new design increases the surface area of each strainer from 150.4 square feet to 1818.5 square feet. The diameter of the screen perforations has been reduced from 0.25 inches to 0.095 inches, thus significantly reducing the potential for downstream debris effects.

The surveillance procedure for inspection of the new design strainers has been implemented. The procedure requires a visual inspection of the entire exterior and the interior of the strainers, which includes a visual inspection of the sump and the vortex suppressor.

The procedure for design-change packages has been enhanced with additional controls related to managing potential debris sources such as insulation, post-LOCA recirculation flow paths, qualified coatings, addition of aluminum or zinc, and effect of post-LOCA debris on downstream components.

STPNOC has implemented actions described in its responses dated August 11, 2003, November 11, 2004, and July 13, 2005, to Bulletin 2003-01, Potential Impact of Debris Blockage on Emergency Sump Recirculation at Pressurized Water Reactors. The measures implemented include refilling the refueling water storage tank after verification of proper swap over to cold-leg recirculation, provision of guidance in emergency operating procedures for restoration of recirculation or for alternate cooling methods if flow blockage occurs, and operator training on indications of and response to strainer clogging.

For smaller LOCAs, it is possible to cooldown and depressurize the RCS to cold shutdown conditions before the RWST is drained to the switchover level. Therefore cold leg recirculation is not required to be established, and sump blockage is not an issue.

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continued Issue or Reference Summary Alternative water sources Containment cleanliness and foreign material OPSP03-XC- 0002 OPSP03-XC- 0002A RWST refill is not an assumed evolution in the STP safety analyses and plant design bases. However, post accident response instructions to refill the RWST, once it has been determined a loss of ECCS recirculation capability exists, are provided in OPOP05-EOEC1l, "Loss of Emergency Coolant Recir-culation". Procedure OPOP05-EO-EC1l actions provide guidance that results in reducing outflow from the RWST. The following actions may be taken to address degraded ECCS recirculation flow, which may be caused by the containment recirculation sump clogging: Stopping CS pumps not needed for containment pressure control; Reducing ECCS flow to the minimum required for decay heat removal, adding makeup to the RWST and; Injecting makeup into the RCS from alternate sources.

The RWST level is normally maintained at a nominal level from 490,000 to 500,000 gallons. This RWST level assures capacity above the Technical Specification 3.5.5 minimum required volume of 458,000 gallons, and is also above the current low alarm level of 473,000 gallons.

STPNOC's procedures for establishing and maintaining containment cleanliness are effective barriers for controlling loose debris and potential sources of loose debris. Procedures OPSP03-XC-0002 "Initial Containment Inspection To Establish Integrity" and OPSP03-XC-0002A "Partial Containment Inspec-tion (Containment Integrity Established)", "Visual Inspection of Containment for Loose Debris" are applied to assure containment cleanliness.

Performed prior to entering MODE 4 during plant startup, and details a visual inspection of all accessible areas of Containment prior to establishing Containment Integrity to verify no loose debris is present which could be transported to the Containment Sump and cause restriction of pump suctions during LOCA conditions. Actions performed in the course of this procedure include an elevation-by-elevation check to confirm the absence of loose debris that could clog the sump and confirmation that all temporary storage box lids are in place and secured and all tool cabinet doors are closed and secured.

Performed for containment entries that are NOT under the control of OPSP03-XC-0002. OPSP03-XC-0002A maintains validity of the in-progress or completed requirements of OPSP03-XC-0002 for transition to Mode 4. This procedure may be performed prior to commencing OPSP03-XC-0002 to aid in establishing controls for Containment Building work activities in preparation for establishing Containment Integrity.

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... continued Issue or Reference Summary Performed after Containment Integrity is established for visual inspection of the affected areas within Containment at the completion of each Containment entry to verify no loose debris is present which could be transported to the Containment Sumps and cause restriction of pump suctions during LOCA conditions. When Containment Integrity is established or being established, these procedures apply to all entries into the Containment. They are performed under the direction of the Shift Supervisor.

All Containment entries require a pre-job briefing, which is typically performed by a Senior Reactor Operator, that addresses the requirements for Containment cleanliness, the definition of loose debris, and reinforces a high level of expectations for housekeeping and control of material. Information from Bulletin 2003-01 is included in briefings.

The process of performing containment cleanup prior to restart is a focused effort with experienced individuals assigned responsibility for areas of the containment. In-process walk-downs are performed by station management and Operations and a final acceptance walk-down is performed by Operations to confirm all requirements for Containment Integrity are met. STPNOC has a high level of confidence in the process to assure the containment building is free of loose debris.

n C;1 Containment drainage paths Sump screens are free of ad-verse gaps and breaches Instrumentation STP procedures require confirmation that the flanged flow paths that allow drainage from the reactor cavity are open. In addition, the process of restoring Containment Integrity, including the perfor-mance of the XC-0002/2A inspections provides assurance that the Containment meets its design basis configuration.

STP surveillance procedures require the sumps to be inspected during each refueling.

Indications of pump cavitation (NPSHA dropping below-NPSHR) such as erratic current, flow or dis-charge pressure can indicate a loss of or degraded suction supply, such as that caused by containment recirculation sump clogging. ECCS and CS pump flow and discharge pressure can be monitored for indications of containment sump clogging following establishment of recirculation flow. Specific in-dications available for operators include: SI/CS Pump Flow (Main Control Board, Plant Computer, Qual PAMS; SI/CS Pump Discharge Pressure (local indication).

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... continued Issue or Reference Summary Operator training The indications and consequences of a degraded containment sump condition at STP have been re-viewed for impact to operator training. Licensed Operator Training includes the monitoring of operat-ing ECCS and CS pumps during the evolution for transfer to cold leg recirculation (OPOP05-EO-ES13, "Transfer To Cold Leg Recirculation") and hot leg recirculation (OPOP05-EO-ES14, "Transfer To Hot Leg Recirculation"). Operator training also includes actions required on a total loss of Emergency Sump recirculation capability (OPOP05-EO-EC11, "Loss of Emergency Coolant Recirculation"). Op-erator training currently includes the recognition of indications of pump distress (NPSHA dropping below NPSHR), such as erratic current, flow or discharge pressure. Initial Licensed Operator training material includes the indications of sump clogging.

Licensed operators are trained on actions to respond to Emergency Sump clogging on a biennial basis in the Licensed operator program. Simulator training objectives are trained every two years on the topics of transfer to cold leg recirculation, transfer to hot leg recirculation, and total loss of Emergency Sump recirculation capability. Specific classroom training on indications of and responses to sump clogging are provided for the licensed operators.

Operator actions STP is a three-train plant and has three CS pumps, based on single failure criteria, one of the three CS pumps may be secured and still meet the current design basis. If a CS actuation occurs and all three CS pumps are operating, then the emergency operating procedures require one CS pump to be secured after verifying containment conditions.

With verification of containment cooling and CS pumps not otherwise required to be operating, the action to remove all CS pumps from service is taken during recirculation according to the emergency operating procedures to preserve RWST inventory.

Injecting more than one Refueling Water Storage Tank (RWST) volume from a refilled RWST is incor-porated in the EOPs, and STP has the guidance to inject more than one RWST volume, coordinated with the Technical Support Center (TSC) continued next page...

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... continued Issue or Reference 1 Summary Refilling the RWST is included in STP Loss of Emergency Coolant Recirculation procedure. This action is taken to extend the time that ECCS and CS pumps can take suction from the RWST and provide cooling to the RCS. RWST makeup is provided for extended time for RCS cooling. RWST is refilled when the loss of recirculation capability occurs. In addition, the "TRANSFER TO COLD LEG RECIRCULATION" procedure commences refilling the RWST after verification of proper swap over to cold leg recirculation.

More Aggressive Cooldown and Depressurization Following A SLOCA is initiated by the Loss of Emer-gency Coolant Recirculation emergency operating procedure. This action is taken to reduce the overall temperature of the RCS coolant and metal temperature to reduce the need for supporting plant sys-temns and equipment required for heat removal. Cooldown is established to reduce the heat energy remaining in the primary thus reducing the cooling requirements of the ECCS.

NOC-AE-05001922 STPNOC has several programmatic controls that address potential sump debris items.

Insulation replacement inside containment is either a like-for-like replacement as a maintenance activity

("rework") or is a modification with a design change that has been approved by STPNOC Engineering.

The STPNOC design change process ensures that new insulation material that differs from the initial design is evaluated.

STPNOC has a procedure that governs signs and labels containing the requirements for labeling inside containment. These requirements are used to minimize potential sump debris items.

The latent debris at STP has been evaluated through containment condition assessments. Containment walkdowns were completed for Unit 1 and for Unit 2 in accordance with the guidance of NEI 02-01, "Condition Assessment Guidelines, Debris Sources Inside Containment", Revision 1. The quantity and composition of the latent debris was evaluated by extensive sampling for latent debris (dirt/dust and latent fiber) considering the guidance in NEI 04-07, Volume 2. The results of the latent debris calculation conservatively determined the debris loading to be less than 160 lbm in each containment.

Therefore, it was elected to use a conservative bounding value of 200 Ibm for the latent debris source term in containment.

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... continued Issue or Reference Summary Visual examination of the latent debris showed very low fiber content. In lieu of analysis of samples, conservative values for debris composition properties were assumed as recommended by NEI 04-07 Volume 2. This results in a very conservative estimate of fiber content. The particulate/fiber mix of the latent debris is assumed to be 15% fiber Containment condition assessments included the identification of miscellaneous solid objects such as labels and tags. Qualified tags attached with stainless steel wires were found for much of the equipment. Unqualified items were identified and removed. The total surface area for any remaining debris of this type was determined to be much less than 100 sq-ft. Therefore, as suggested by NEI 04-07, this miscellaneous solid object debris source is bounded by 100 sq-ft in STP debris generation and transport analyses.

STPNOC periodically conducts condition assessments of coatings inside containment. Coating con-dition assessments are conducted as part of the structures monitoring program. Visual inspection of coatings in containment is intended to characterize the condition of the coating systems. If localized ar-eas of degraded coatings are identified, those areas are evaluated and scheduled for repair/replacement as necessary.

COr Plant modifications imple-mented in Unit 1 during the Fall 2006 refueling outage and in Unit 2 during the Spring 2007 refueling outage New strainer details Remove existing sump screens from each of the three emergency sumps.

Maintain existing vortex breakers in place.

The old strainer screen had perforations of 0.25 inches diameter. Water entering the suction pipe from the sump may contain small particles less than 0.25 inches diameter. These particles cannot clog the containment spray nozzles (3/8-inch orifice diameter) which are the limiting restrictions found in any system served by the sump. The new strainers have a screen hole size of 0.095 inches diameter and thus meet this design requirement. The new strainer design provides improved capability to filter fine debris due to their decreased opening size. While the decreased perforation size may tend to increase head loss, under the current licensing basis methodology, this effect is more than offset by the significant increase in strainer area.

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..continued Issue or Reference Summary The previous STP sumps were designed according to Regulatory Guide 1.82, proposed Revision 1, dated May 1983. The guidance in proposed Revision 1 of Regulatory Guide 1.82 recommends a calculation of the sump screen head loss due to debris blockage. The licensee indicated that, utilizing the current licensing basis methodology from proposed Revision 1 of Regulatory Guide 1.82, the NPSHA is sufficient to accommodate this calculated head loss. The new strainers have a surface area of 1818.5 square feet per sump. The old screens had a surface area of 155.4 square feet per sump. Thus, for the current licensing basis debris loading, the debris head loss with the new strainers will be substantially smaller than for the old screens.

The new STP Unit 1 and Unit 2 strainer installation does not affect the independence and redundancy of the ECCS and CS sumps. Three independent sumps are maintained by the new strainer design.

Procedural requirements The sump inspection procedure (required by Technical Specifications Surveillance Requirement 4.5.2.d) includes criteria to assure the following: No external evidence of structural distress or abnormal cor-rosion; No pathways that would allow foreign objects or debris to enter the sump; There are no structural joints with gaps larger than 0.095 inches; There are no gaps in the strainer modules or associated piping fit-up connections; There are no foreign materials remaining on or lodged into the gaps of the strainer modules; There are no foreign materials inside the strainer core tubes, including the two strainer modules connected on a 45-degree angle on sumps "A" and "B"; The sump suction inlet is not restricted; The sump is dry, free of foreign objects, debris, and boron crystal build-up.

Guidance to delay depletion of the RWST after switchover to sump recirculation is currently contained in Emergency Operating Procedure OPOP05-EO-ECli, "Loss of Emergency Coolant Recirculation".

This procedure provides actions to reduce the outflow from the RWST to preserve the RWST inven-tory once it has been determined that a loss of sump recirculation capability exists. The procedure establishes a process to determine the actions for delaying RWST inventory depletion, while ensuring adequate core cooling flow and containment heat removal as necessary.

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continued Issue or Reference Summary For small to medium LOCAs, guidance to delay depletion of the RWST before switchover to sump recirculation currently exists in procedure OPOP05-EO-ES12, "Post LOCA Cooldown and Depressur-ization". This procedure provides actions to cooldown and depressurize the RCS to reduce the break flow, thereby reducing the injection flow necessary to maintain RCS subcooling and inventory. The operating HHSI pumps are sequentially stopped to reduce injection flow, based on pre-established criteria that maintain core cooling, resulting in less outflow from the RWST.

For smaller LOCAs, it is possible to cooldown and depressurize the RCS to cold shutdown conditions before the RWST is drained to the switchover level. Therefore cold leg recirculation is not required to be established, and sump blockage is not an issue.

The new strainer configuration maintains the independence and redundancy of the existing three-train sump configuration 0

Recirculation Break flow The reduced average strainer approach velocity will tend to decrease the potential for large pieces of debris in the flow stream approaching the sump from damaging the strainers. Also, the new strainers are of robust construction. The staff further considers the filtration capability of the new sump strainers to be superior to the combined capability of the old screens and trash racks because of the new strainers larger surface area and complex geometry. The new strainers satisfy the licensing basis functions associated with the previously installed trash racks.

The emergency operating procedure for the loss of emergency coolant recirculation provides guidance (by reference to the ERG) for restoration of recirculation as well as the contingencies for cooling down and depressurizing the RCS in the event that recirculation can not be restored. The major actions of this procedure are: Continue attempts to restore Emergency Coolant Recirculation (ECR), with the first priority to access the equipment needed for ECR and restore that equipment prior to performing any extreme recovery actions; Increase/Conserve RWST level, makeup is added to extend the time available for pumps to take suction from the RWST.

Limit outflow by securing unneeded CS pumps and limiting ECCS pump flowrate(s); Commencing a cooldown/depressurization to Cold Shutdown at a 100°F/hr cooldown to limit coolant leakage while minimizing thermal stresses thus remaining within limits; Depressurize the RCS to minimize RCS subcooling to reduce break flow from the LOCA.

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... continued Issue or Reference Summary Makeup Try to add makeup to the RCS from alternate source utilizing the normal Chemical and Volume Control System equipment; Depressurize steam generators to cool down and depressurize the RCS:

A controlled depressurization of the Steam Generators (SG) will allow the SI accumulators to inject, minimize the break flow and allow the RCS to reach Residual Heat Removal (RHR) System cut-in conditions.

Heat removal Establish and maintain RHR conditions or utilizing steam dumps. Consult the plant engineering staff for further actions at this point for additional recovery actions SRM-SECY-12-0093 Given the vastly enlarged advanced strainers installed, compensatory measures already taken, and the low probability of challenging pipe breaks, adequate DID is currently being maintained.

Compensatory actions and modifications made to date have reduced the risk of strainer clogging. All PWR licensees have made their sump strainers substantially larger.

One of the main objectives of the risk-informed approach is to estimate the difference in risk (delta risk) if some or all fibrous insulation were to remain installed at the plant. The STPNOC Pilot Project approach does not consider a transition break size. Rather the approach analyzes a full spectrum of postulated LOCA, including DEGB for all pipe sizes up to the largest pipe in the RCS, the design basis accident (DBA) LOCA. The STP approach attempts to characterize the physical behavior of debris generation and transport over a full range of plausible conditions. Some aspects of GSI-191 have limited data support; thus uncertainty characterization in the form of CDFs is an important part of the description of the parameters modeled in the PRA basic events.

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