NL-16-2002, Supplemental Response to NRC Generic Letter 2004-02

Supplemental Response to NRC Generic Letter 2004-02

ML17116A098
Person / Time
Site:	Vogtle
Issue date:	04/21/2017
From:	Hutto J Southern Nuclear Operating Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
GL-04-002, NL-16-2002
Download: ML17116A098 (415)

Text

{{#Wiki_filter:A Southern Nuclear J. J. Hutto Regulatory Affairs Director

• 40 Inverness Center Parkway Post Office Box 1295 Birmingham, AL 35242 205 992 5872 tel 205 992 7601 fax jjhutto@southemco.com APR 2 1 2017 Enclosure 2 to this letter contains Proprietary Information to be withheld from public disclosure under 10 CFR 2.390. When separated from Enclosure 2, this transmittal document and the other Enclosures are decontrolled.

Docket Nos.: 50-424 NL-16-2002 50-425 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Vogtle Electric Generating Plant - Units 1 & 2 Supplemental Response to NRC Generic Letter 2004-02 Ladies and Gentlemen: The purpose of this report is to provide, for Nuclear Regulatory Commission (NRC) staff review and approval, the Southern Nuclear Operating Company (SNC) supplemental response for Vogtle Electric Generating Plant Units 1 & 2 (VEGP) to Generic Letter (GL) 2004-02, dated September 13, 2004, "Potential Impact of Debris Blockage on Emergency Recirculation during Design Basis Accidents at Pressurized-Water Reactors." This supplemental response supersedes previous responses and uses a risk informed approach to evaluate effects of debris. This report is organized as described below:

• Enclosure 1 provides a high-level summary of Enclosures 2 through 4, and is organized with the same layout as draft Regulatory Guide (RG) 1.229 Section C.
• Enclosure 2 provides a detailed description of the plant-specific conditions and models related to generic safety issue (GSl)-191 (including proprietary information). This enclosure is organized in accordance with the content guideline for GL 2004-02 responses. This enclosure also includes a response to each of the previous requests for additional information (RAls) that VEGP had received on earlier GL 2004-02 submittals.

Accordingly, the responses provided in this enclosure supersedes those provided in previous SNC responses. Additionally, this enclosure contains attachments with affidavits for withholding of proprietary information.

• Enclosure 3 provides a description of the risk quantification using the NARWHAL computer code and the VEGP probabilistic risk assessment (PRA) model. This enclosure is organized with the same layout as draft RG 1.229 Appendix A The enclosure explains how all the individual parts are combined to quantify risk. It also provides discussion on uncertainty quantification.

U.S. Nuclear Regulatory Commission NL-16-2002 Page 2

• Enclosure 4 provides a summary of defense-in-depth and safety margin. This enclosure shows that the health and safety of the public are not adversely affected by debris-related failures of the strainers, pumps, downstream components, or core.
• Enclosure 5 is a duplicate of Enclosure 2 with the proprietary information redacted.

The determination of in-vessel debris limits is necessary to support the final VEGP risk-informed resolution to GL 2004-02 (including a corresponding license amendment request). Please note that the methodology SNC intends to use to determine in-vessel debris limits is currently under NRC review. By letter dated February 14, 2017, the NRC stated that it would not be appropriate for the staff to accept for review a requested licensing action (RLA) that relied upon an unapproved methodology. However, in this letter, the NRC stated their support for an SNC technical report that does not rely on an unapproved methodology. The intent of this technical report is that it will be used to prepare an NRC staff evaluation to support a subsequent RLA submittal after the in-vessel debris limits methodology is approved. Accordingly, the purpose of this report is to receive NRC review and approval of the SNC supplemental GL 2004-02 response which uses a risk-informed methodology to evaluate debris effects, excepting in-vessel fiber limits, which will be provided with the RLA. To support a timely RLA (including a license amendment request pursuant to 10 CFR 50.90) that will resolve this safety issue, SNC requests NRC approval of this report by April 21, 2018. If you have any questions, please contact Ken McElroy at 205.992.7369. Mr. J. J. Hutto states he is Regulatory Affairs Director of Southern Nuclear Operating Company, is authorized to execute this oath on behalf of Southern Nuclear Operating Company and, to the best of his knowledge and belief, the facts set forth in this letter are true. Respectfully submitted, 9<fbc-:> J. J. Hutto Regulatory Affairs Director JJH/RMJ Sworn to and subscri ed before me this _d}_ day of A~i '2017. i(/V\. --=-----1..~~---=-------=~~----J'-- My commission expires: /!J - I?-;)__() ( 1

U. S. Nuclear Regulatory Commission NL-16-2002 Page 3

Enclosures:

1. Introduction and Overall Summary

2. Supplemental Response to NRC Generic Letter 2004-02 (Proprietary)

3. Risk Quantification

4. Defense-in-Depth and Safety Margin

5. Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) cc: Southern Nuclear Operating Company Mr. S. E. Kuczynski, Chairman, President & CEO Mr. D. G. Bost, Executive Vice President & Chief Nuclear Officer Mr. R. D. Gayheart, Fleet Operations General Manager Mr. M. D. Meier, Vice President - Regulatory Affairs Mr. B. K. Taber, Vice President - Vogtle 1 & 2 Mr. B. J. Adams, Vice President- Engineering Mr. D. D. Sutton, Regulatory Affairs Manager - Vogtle 1 & 2 RType: CVC?OOO U.S. Nuclear Regulatory Commission Ms. C. Haney, Regional Administrator Mr. M. D. Orenak, NRR Project Manager - Vogtle 1 & 2 Mr. M. F. Endress, Senior Resident Inspector - Vogtle 1 & 2 State of Georgia Mr. R. E. Dunn, Director- Environmental Protection Division

CAW-17-456'5 AFFIDAVIT COMMONWEALTH OF PENNSYLVANIA: SS COUNTY OF BUTLER: . I, James A: Gresham, am authorized to execute this Affidavit on behalf of Westinghouse Electric

• Co1Upany LLC ("Westingliouse~') ahd declare that the averments of fact set forth in this. Affida.vit are true*

and correct to the best of my knowledge, information, atid belief. EXecuted on: +/tr fe !

                                                      .. ~Jaines A. Gresham, Manager v Regulafory Compliance

3 CAW-17-4565 (1) I am Manager, Regulatory Compliance, Westinghouse Electric Company LLC ("Westinghouse"), and as such, I have been specifically delegated the function of reviewing the proprietary information sought to be withheld from public disclosure in connection with nuclear power plant licensing and rule making proceedings, and am authorized to apply for its withholding on behalf of Westinghouse. (2) I am making this Affidavit in conformance with the provisions of 10 CFR Section 2.390 of the Nuclear Regulatory Commission's ("Commission's") regulations and in coajunction with the Westinghouse Application for Withholding Proprietary Information from Public Disclosure accompanying this Affidavit. (3) I have personal knowledge of the criteria and procedures utilized by Westinghouse in designating information as a trade secret, privileged or as confidential commercial or financial information. (4) Pursuant to the provisions of paragraph (b)(4) of Section 2.390 of the Commission's regulations, the following is furnished for consideration by the Commission in determining whether the information sought to be withheld from public disclosure should be withheld. (i) The information sought to be withheld from public disclosure is owned and has been held in confidence by Westinghouse. (ii) The information is ofa type customarily held in confidence by Westinghouse and not customarily disclosed to the public. Westinghouse has a rational basis for determining the types of information customarily held in confidence by it and, in that connection, utilizes a system to determine when and whether to hold certain types of information in confidence. The application of that system and the substance of that system constitute Westinghouse policy and provide the rational basis required. Under that system, information is held in confidence if it falls in one or more of several types, the release of which might result in the loss of an existing or potential competitive advantage, as follows: (a) The information reveals the distinguishing aspects of a process (or component, structure, tool, method, etc.) where prevention of its use by any of

4 CAW-17-4565 Westinghouse's competitors without license from Westinghouse constitutes a competitive economic advantage over other companies. (b) It consists of supporting data, including test data, relative to a process (or component, structure, tool, method, etc.), the application of which data secures a competitive economic advantage, e.g., by optimization or improved marketability. (c) Its use by a competitor would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing a similar product. (d) It reveals cost or price information, production capacities, budget levels, or commercial strategies of Westinghouse, its customers or suppliers. (e) It reveals aspects of past, present, or future Westinghouse or customer funded development plans and programs of potential commercial value to Westinghouse. (f) It contains patentable ideas, for which patent protection may be desirable. (iii) There are sound policy reasons behind the Westinghouse system which include the following: (a) The use of such information by Westinghouse gives Westinghouse a competitive advantage over its competitors. It is, therefore, withheld from disclosure to protect the Westinghouse competitive position. (b) It is information that is marketable in many ways. The extent to which such information is available to competitors diminishes the Westinghouse ability to sell products and services involving the use of the information. (c) Use by our competitor would put Westinghouse at a competitive disadvantage by reducing his expenditure of resources at our expense.

5 CAW-17-4565 (d) Each component of proprietary information pertinent to a particular competitive advantage is potentially as valuable as the total competitive advantage. If competitors acquire components of proprietary information, any one component may be the key to the entire puzzle, thereby depriving Westinghouse of a competitive advantage. (e) Unrestricted disclosure would jeopardize the position of prominence of Westinghouse in the world market, and thereby give a market advantage to the competition of those countries. (f) The Westinghouse capacity to invest corporate assets in research and development depends upon the success in obtaining and maintaining a competitive advantage. (iv) The information is being transmitted to the Commission in confidence and, under the provisions of 10 CFR Section 2.390, is to be received in confidence by the Commission. (v) The information sought to be protected is not available in public sources or available information has not been previously employed in the same original manner or method to the best of our knowledge and belief. (vi) The proprietary information sought to be withheld in this submittal is that which is appropriately marked in "Vogtle Electric Generating Plant Supplemental Response to NRC Generic Letter 2004-02, Enclosure 2, 'Supplemental Response to NRC Generic Letter 2004-02' "(Proprietary), for submittal to the Commission, being transmitted by Letter GP-19572. The proprietary information as submitted by Westinghouse is that associated with resolution of and response to NRC Generic Letter 2004-02 and may be used only for that purpose. (a) This information is part of that which will enable Westinghouse to provide commercial support for resolution of and response to NRC Generic Letter 2004-02.

6 CAW-17-4565 (b) Further this information has substantial commercial value as follows: (i) Westinghouse plans to sell the use of similar information to its customers for the purpose of providing support for resolution of and response to NRC Generic Letter 2004-02. (ii) Westinghouse can sell support and defense of industry guidelines and acceptance criteria for plant-specific applications. (iii) The information requested to be withheld reveals the distinguishing aspects ofa methodology which was developed by Westinghouse. Public disclosure of this proprietary information is likely to cause substantial harm to the competitive position of Westinghouse because it would enhance the ability of competitors to provide similar technical evaluation justifications and licensing defense services for commercial power reactors without commensurate expenses. Also, public disclosure of the information would enable others to use the information to meet NRC requirements for licensing documentation without purchasing the right to use the information. The development of the technology described in part by the information is the result of applying the results of many years of experience in an intensive Westinghouse effort and the expenditure ofa considerable sum of money. In order for competitors of Westinghouse to duplicate this information, similar technical programs would have to be performed and a significant manpower effort, having the requisite talent and experience, would have to be expended. Further the deponent sayeth not.

Proprietary Information Notice Transmitted herewith are proprietary and non-proprietary versions of a document, furnished to the NRC associated with resolution of and response to NRC Generic Letter 2004-02 and may be used only for that purpose. In order to conform to the requirements of 10 CFR 2.390 of the Commission's regulations concerning the protection of proprietary information so submitted to the NRC, the information which is proprietary in the proprietary versions is contained within brackets, and where the proprietary information has been deleted in the non-proprietary versions, only the brackets remain (the information that was contained within the brackets in the proprietary versions having been deleted). The justification for claiming the information so designated as proprietary is indicated in both versions by means of lower case letters (a) through (f) located as a superscript immediately following the brackets enclosing each item of information being identified as proprietary or in the margin opposite such information. These lower case letters refer to the types of information Westinghouse customarily holds in confidence identified in Sections (4)(ii)(a) through (4)(ii)(f) of the Affidavit accompanying this transmittal pursuant to 10 CFR 2.390(b)(l). Copyright Notice The reports transmitted herewith each bear a Westinghouse copyright notice. The NRC is permitted to make the number of copies of the information contained in these reports which are necessary for its internal use in connection with generic and plant-specific reviews and approvals as well as the issuance, denial, amendment, transfer, renewal, modification, suspension, revocation, or violation of a license, permit, order, or regulation subject to the requirements of 10 CFR 2.390 regarding restrictions on public disclosure to the extent such information has been identified as proprietary by Westinghouse, copyright protection notwithstanding. With respect to the non-proprietary versions of these reports, the NRC is permitted to make the number of copies beyond those necessary for its internal use which are necessary in order to have one copy available for public viewing in the appropriate docket files in the public document room in Washington, DC and in local public document rooms as may be required by NRC regulations if the number of copies submitted is insufficient for this purpose. Copies made by the NRC must include the copyright notice in all instances and the proprietary notice ifthe original was identified as proprietary.

AL I 0 N 5(1f?lCE A~D TECHNOlOejiY ALION Science & Technology AFFIDAVIT We, Andy Roudenko, Project Manager and Martin Rozboril, Jr. Assistant Vice President Division Manager (AVPDM) state as follows: (1) We, Andy Roudenko, Project Manager, and Martin Rozboril, Jr. AVPDM, Nuclear Services, ALION Science & Technology ("Alion") and have been delegated the function of reviewing the information described in paragraph (2) which is sought to be withheld, and have been authorized to apply for its withholding. (2) The information sought to be withheld is contained in all revisions of ALION Science & Technology report "Erosion Testing of Small Pieces of Low Density Fiberglass Debris-Test Report," ALION-REP-ALION-1006-04, with the latest revision to date, Rev. I, dated November 17, 20011. Information from this report was used to support analysis of post-LOCA debris transport in work designed to address GSI-191, Assessment of Debris Accumulation on PWR Sump Performance, issues at Southern Nuclear Operating Company, Vogtle Units 1 and 2. Specifically, the following Sections and Figures are to be withheld, on that basis that these unique attribute of the testing approach, test results and conclusions:

• Background
• Figure 1.1. I
• Figure 2.1.2
• Figure 2.1.3
• Figure 2.1.5
• Figure 2.1.6
• Figure 2.1.9
• Test Results, including Figures and Tables
• Data Analysis, including Figures and Tables
• Conclusions
• Appendices (3) In making this application for withholding of proprietary information of which it is the owner, Alion relies upon the exemption from disclosure set forth in the Freedom of Information Act ("FOIA"), 5 USC Sec. 552(b)(4), and the Trade Secrets Act, 18 USC Sec. 1905, and NRC regulations 10 CFR 9.l 7(a)(4), and 2.390(a)(4) for "trade secrets" (Exemption 4). The material for which exemption from disclosure is here sought also qualify under the narrower definition of "trade secret", within the meanings assigned to those terms for purposes of FOIA Exemption 4 in, respectively, Critical Mass Energy Project v. Nuclear Regulatory Commission, 975F2d871 (DC Cir. 1992), and Public Citizen Health Research Group v. FDA, 704F2dl280 (DC Cir. 1983).

Page I of3 MAS Affidavit

                                        .ALION SCIEMC.E A.ND TECHNOLOGY (4) Some examples of categories of information which fit into the definition of proprietary information are:

a. Information that discloses a process, method, or apparatus, including supporting data and analyses, where prevention of its use by Alion's competitors without license from Alion constitutes a competitive economic advantage over other companies

b. Information which, if used by a competitor, would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing of a similar product;

c. Information which reveals aspects of past, present, or future Alion customer-funded development plans and programs, resulting in potential products to Alion;

d. Information which discloses patentable subject matter for which it may be desirable to obtain patent protection.

The information sought to be withheld is considered to be proprietary for the reasons set forth in paragraphs (4) a, and (4) b, above. (5) To address 10 CFR 2.390 (b) (4), the information sought to be withheld is being submitted to NRC in confidence. The information is of a sort customarily held in confidence by Alion, and is in fact so held. The information sought to be withheld has, to the best of my knowledge and belief, consistently been held in confidence by Alion, no public disclosure has been made, and it is not available in public sources. All disclosures to third parties including any required transmittals to NRC, have been made, or must be made, pursuant to regulatory provisions or proprietary agreements which provide for maintenance of the information in confidence. Its initial designation as proprietary information, and the subsequent steps taken to prevent its unauthorized disclosure, are as set forth in paragraphs (6) and (7) following. (6) Initial approval of proprietary treatment of a document is made by the manager of the originating component, the person most likely to be acquainted with the value and sensitivity of the information in relation to industry knowledge. Access to such documents within Alion is limited on a "need to know" basis. (7) The procedure for approval of external release of such a document typically requires review by the staff manager, project manager, principal scientist or other equivalent authority, by the manager of the cognizant marketing function (or their delegate), and by the Legal Operation, for technical content, competitive effect, and determination of the accuracy of the proprietary designation. Disclosures outside Alion are limited to regulatory bodies, customers, and potential customers, and their agents, suppliers, and licensees, and others with a legitimate need for the information, and then only in accordance with appropriate regulatory provisions or proprietary agreements. (8) The document identified in paragraph (2), above, is classified as proprietary because it contains "know-how" and "unique data" developed by Alion within our research and Page 2 of 3 MAS Affidavit

ft.LION SCIENCE AND TECHNOLOGY development programs. The development of this document, supporting methods and data constitutes a major Al ion asset in this current market. (9) Public disclosure of the information sought to be withheld is likely to cause substantial harm to Alion's competitive position and foreclose or reduce the availability of profit-making opportunities. The information is part of Alion's comprehensive BWR/PWR GSI-191 analysis base, and its commercial value extends beyond the original development cost. The value of the technology base goes beyond the extensive physical database and experimental methodology and includes development of the expertise to determine and apply the appropriate evaluation process. The research, development, engineering, analytical and experimental costs comprise a substantial investment of time and money by Alion. The precise value of the expertise to devise an evaluation process and apply the correct analytical methodology is difficult to quantify, but it clearly is substantial. Alion's competitive advantage will be lost if its competitors are able to use the results of the Alion experience to normalize or verify their own process or if they are able to claim an equivalent understanding by demonstrating that they can arrive at the same or similar conclusions. The value of this information to Alion would be lost if the information were disclosed to the public. Making such information available to competitors without their having been required to undertake a similar expenditure of resources would unfairly provide competitors with a windfall, and deprive Alion of the opportunity to exercise its competitive advantage to seek an adequate return on its large investment in developing these very valuable analytical tools. I declare under penalty of perjury that the foregoing affidavit and the matters stated therein are true and correct to the best of my knowledge, information, and belief. Executed on this 20th day of April 2017. Martin r Digitally signed by Andy Roudenko

                                   ~! DN: cn=Andy Roudenko, o=Alion Rozboril, Jr.

d- r f( V ~I  ;' 1

                                 .*~\Science and Technology, ou=Nudear
                                 ,1 ':services Division,
                                      ~~ail=aro~denko@alionscience.com, 2017.04.20

_ ,/_..,.,.,-~*=us.....,-:-

                       , *,/          Date:2017.04.2012:09:50-07'00' 15:06:17 -06'00'

( .* Andy Roudenko Martin Rozboril, Jr. Project Manager Assistant Vice President ALION Science & Technology Division Manager, Nuclear Services ALION Science & Technology Page 3 of3 MAS Affidavit

Vogtle Electric Generating Plant - Units 1 & 2 Supplemental Response to NRC Generic Letter 2004-02 Enclosure 1 Introduction and Overall Summary

Enclosure 1 Introduction and Overall Summary Table of Contents 1.0 Introduction 2.0 Systematic Risk Assessment of Debris 3.0 Initiating Event Frequencies 4.0 Defense-in-Depth and Safety Margin 5.0 Uncertainty 6.0 Monitoring Program 7.0 Quality Assurance 8.0 Periodic Update of Risk-Informed Analysis 9.0 Reporting and Corrective Actions 10.0 License Application 11.0 References Attachments E1:A1 Resolution of the VEGP Internal Events PRA Peer Review Findings E1 :A2 Comparison of RG 1.200 Revision 1 and Revision 2 Supporting Requirements E1 :A3 Resolution of the VEGP Seismic PRA Peer Review Findings E1-1

Enclosure 1 Introduction and Overall Summary 1.0 Introduction In 2010, due to the ongoing challenges of resolving Generic Safety Issue (GSl)-191, the United States Nuclear Regulatory Commission (NRG) commissioners issued a staff requirements memorandum (SRM) directing the NRG staff to consider new and innovative resolution approaches (Reference 1). One of the approaches included in the SRM was the option of addressing GSl-191 using a risk-informed approach. In 2011, South Texas Project (STP) initiated a multi-year effort as a pilot plant to define and implement a risk-informed approach to address the concerns associated with GSl-191. In 2012, the NRG staff issued SRM-SECY-12-0093 (Reference 2) providing recommendations for closure options, and these options were accepted by the NRG commissioners. In 2013, Southern Nuclear Operating Company (SNC) selected Option 2b (full risk-informed resolution path) for closure of NRG Generic Letter (GL) 2004-02 (Reference 24) at Vogtle Electric Generating Plant (VEGP) Units 1 and 2 (Reference 3). The objective of GSl-191 is to ensure that post-accident debris blockage will not impede or prevent the operation of the emergency core cooling system (ECCS) or containment spray system (CSS) in recirculation mode at pressurized water reactors (PWRs) during loss of coolant accidents (LOCAs) or other high energy line break (HELB) accidents for which recirculation is required (Reference 4). SNC has provided multiple responses to the NRG supporting the resolution of GSl-191. An in-depth history of the VEGP correspondences issued by or submitted to the NRG on the subject of GSl-191 is provided in Sections 1.0 and 2.0 of Enclosure 2, documenting VEGP's compliance with regulatory requirements. This submittal provides a complete summary of the risk-informed GSl-191 evaluation performed for VEGP Units 1 and 2, superseding all previous GL 2004-02 responses. The VEGP_ GSl-191 submittal is organized as described below:

• Enclosure 1 provides a high-level summary of the other enclosures, and is organized with the same layout as draft Regulatory Guide (RG) 1.229 Section C (Reference 4).
• Enclosure 2 provides a detailed description of the plant-specific conditions and models related to GSl-191 (including some proprietary information). This enclosure is organized in accordance with the content guideline for GL 2004-02 responses (Reference 5). This enclosure also includes a response to each of the previous requests for additional information (RAls) that VEGP had received on earlier GL 2004-02 submittals. Additionally, this enclosure contains attachments with affidavits for withholding the proprietary information.
• Enclosure 3 provides a description of the risk quantification using NARWHAL and the VEGP probabilistic risk assessment (PRA) model. This enclosure is organized with the same layout as draft RG 1.229 Appendix A (Reference 4). The enclosure explains how all of the individual parts are combined to quantify risk. It also provides discussion on uncertainty quantification.

E1-2

Enclosure 1 Introduction and Overall Summary

• Enclosure 4 provides a summary of defense-in-depth and safety margin. This enclosure shows that the health and safety of the public are not adversely affected by debris-related failures of the strainers, pumps, downstream components, or core.
• Enclosure 5 is a duplicate of Enclosure 2 with the proprietary information redacted.

The overall evaluation for VEGP is based heavily on models that have been used in the past and accepted by the NRC for GSl-191 resolution. The results of this evaluation show with high confidence that the risk associated with GSl-191 is very low, as defined by RG 1.174 Region Ill (Reference 6). In addition, the analysis includes significant safety margin and does not affect any of the existing defense-in-depth measures that are in place to protect the public. 2.0 Systematic Risk Assessment of Debris As described in RG 1.174 (Reference 6), the systematic risk assessment should consider all hazards, initiating events, and plant operating modes. However, a screening process can be used to eliminate scenarios that are not relevant, not affected by debris, or have an insignificant contribution. The specific GSl-191 failure modes that were considered are:

1. Debris accumulation in an upstream flow path choke point (e.g., a refueling canal drain) exceeds blockage limits and reduces the available sump volume.

2. Strainer head loss exceeds the net positive suction head (NPSH) margin for the ECCS and CSS pumps when the strainer is fully submerged.

3. Strainer head loss exceeds half of the submerged strainer height when the strainer is partially submerged.

4. Strainer head loss exceeds the strainer structural margin.

5. Gas voids (i.e., water vapor due to flashing or air intrusion due to degasification or vortexing) downstream of the strainers exceed the acceptable void fraction limits of the ECCS and CSS pumps.

6. Debris penetration exceeds ex-vessel downstream effects limits for component wear or clogging.

7. Debris penetration exceeds in-vessel downstream effects limits for core blockage.

8. Buildup of oxides and other chemical precipitates on fuel cladding exceed heat transfer limits.

9. Boric acid concentration in the core exceeds the solubility limit resulting in boric acid precipitation.

Failure Modes 1, 6, and 8, as well as the vortexing portion of Failure Mode 5, have been addressed in a bounding manner for the range of possible breaks with no issues of concern (see Enclosure 2-Section 3.1 for upstream blockage, Section 3.f.3 for vortexing, Section 3.m for ex-vessel downstream effects, and Section 3.n.1 for the LOCA deposition model (LOCADM) portion of the analysis of in-vessel effects) and were therefore not explicitly modeled in NARWHAL (a software tool that analyzes the E1-3

Enclosure 1 Introduction and Overall Summary GSl-191 phenomenological models in a self-consistent, time-dependent manner). The remaining failure modes were explicitly modeled. Note that core blockage (Failure Mode 7) and boric acid precipitation (Failure Mode 9) were addressed by using assumed debris limits. Figure 1-1 shows the relationship between the various elements of the risk-informed GSl-191 analysis and documentation. Sump Volume (Enc 2) Various Bounding Analyses (Enc 2) Determine min/max post- Fiber Penetration

• Ex-vessel wear and blockage accident sump volume Testing
• Fuel cladding debris deposition (Enc 2) (2014)
• Vortexing for bounding conditions Debris Transport Debris (Enc 2)

Generation Quantify debris Head Loss Testing (Enc 2) transported to (Enc 2) (2009) Quantify debris strainer for each Risk Assessment generated by break (Flow-3D) (Enc 3) breaks at all Calculate delta core weld locations Chemical Debris damage frequency on primary (Enc 2) Evaluation of and delta large early piping Quantify chemical each break release frequency scenario against (Enc 2 (BADGER) precipitate debris all GSl-191 (Enc 3 -~ Break passes ~ (CAFTA) for each break Yes failure criteria (NARWHAL) CAD Model Develop defense in (Enc 2) No depth and mitigative CAD model of strategies (Enc 4) Vogtle Break fails containment (inventor) Prepare LAR with exemptions and Tech 1 4 ' - - - - - - ' Prepare Licensing DOC lo<'----* Develop GL 2004-02 Spec changes (Future (DOC SNCS45368) Submittal (Enc 2) Submittal) Figure 1 Flow chart illustrating analysis and documentation elements 2.1 Hazards, Initiating Events, and Plant Operating Modes The only scenarios that need to be considered for GSl-191 are scenarios that require recirculation through the ECCS and/or CSS strainers. If recirculation is not required, there is no potential for debris-related failures of the strainers, pumps, downstream components, or core. The hazards and initiating events relevant to GSl-191 at VEGP include:

1. Reactor coolant system (RCS) pipe breaks resulting in small, medium, and large LOCAs

2. Non-piping LOCAs

3. Secondary side breaks inside containment (SSBI) that result in a consequential LOCA upon failure to terminate safety injection or a stuck open power-operated relief valve (PORV)

4. Seismically-induced LOCAs

5. Water hammer-induced LOCAs E1-4 L

Enclosure 1 Introduction and Overall Summary These hazards and initiating events are discussed in more detail in Section 3.0. The quantitative risk assessment was performed for LOCAs and SSBls that occur during full power operation (i.e., Mode 1), which is assumed to be equivalent or bounding compared to the other operating modes. This is a reasonable assumption because the RCS pressure and temperature (key inputs affecting the ZOI size) would either be approximately the same or significantly lower for Modes 2 through 6. In addition, the flow rate required to cool the core (a key input affecting core blockage) would be significantly reduced for low power or shutdown modes. 2.2 Risk Attributable to Debris The risk attributable to debris was quantified in terms of the change in core damage frequency (b.CDF) and the change in large early release frequency (b.LERF) compared to a hypothetical plant condition without any debris. This was done using a conservative approach that results in mean b.CDF and b.LERF values that are skewed high (as opposed to a best-estimate approach that would result in a more accurate prediction of the mean b.CDF and b.LERF values, or a bounding approach that would significantly over-predict the mean b.CDF and b.LERF values). The risk quantification was performed using the NARWHAL software (to calculate the conditional failure probabilities (CFPs) associated with the effects of debris) and the VEGP internal events PRA model (with some modifications to represent the GSl-191 failure events accurately). The PRA model of record is referred to as the "base PRA model", and the modified PRA model is referred to as the "GSl-191 PRA model". The base PRA model has been peer reviewed against RG 1.200 (Reference 7) and is therefore appropriate to use for risk-informed applications. In order to support the detailed quantification of the GSl-191 risk impact, the base PRA model was modified to incorporate events for GSl-191 sump strainer and core blockage failures, along with the LOCA initiating *events and equipment configurations associated with each potential GSl-191 failure. The risk evaluation relies on many engineering calculations and tests that have been developed and conducted for VEGP over the last several years to address GSl-191 and GL 2004-02. These calculations and tests are described in detail in Enclosure 2. The GSl-191 risk quantification for VEGP shows that the overall risk associated with debris (GDF, LERF, b.CDF, and b.LERF) is very low as defined by Region Ill of RG 1.174 (Reference 6). Figure 1-2 and Figure 1-3 show the RG 1.174 risk guidelines. E1-5

Enclosure 1 Introduction and Overall Summary t LL 0 (.)

               <I 10*5 Region II 10* 5    - - - - - - -- --   - -"---'

Region Ill CDF --+ Figure 1 RG 1.174 Risk Acceptance Guidelines for CDF and ACDF t LL a: w

                 <1 10"6 Region II Region Ill LEAF-+-

Figure 1 RG 1.174 Risk Acceptance Guidelines for LERF and ALERF As shown in Table 1-3 the total baseline risk (from internal events , internal fire , internal flood , and seismic events) for the VEGP Unit 1 PRA model is 4.39x10-5 per reactor-year (ry-1 ) for CDF and 1.73x1 Q-6 ry-1 for LERF. The total baseline risk for the VEGP Unit 2 PRA model is 5.05x10-5 ry-1 for CDF and 1.90x1Q-6 ry- 1 for LERF . The change in risk calculated using the VEGP GSl-191 PRA model is shown in Table 1-1. Note that the internal events and therefore the GSl-191 PRA models are identical for Units 1 and 2. E1 -6

Enclosure 1 Introduction and Overall Summary Table 1 VEGP Total Risk Impact due to GSl-191 Failures

                                                               .dCDF         .dLERF Case                              {ry-1)         {ry-1)

Risk increase from GSl-191 failures for high-likelihood 2.32x10-8 3.10x10-11 LOCA confi!:1urations Bounding risk increase from GSl-191 failures for 1.41x10-9 4.09x10-12 unlikely LOCA confiQurations Risk increase from GSl-191 failures for seismically- 1.50x10-10 1.50x10-9 induced LOCAs Risk increase from GSl-191 failures for SSBls 1.39x10-9 8.25x10- 11 Total risk increase associated with GSl-191 2.75x10-s 2.68x10-10 Enclosures 2 and 3 provide a detailed description of the GSl-191 models and risk evaluation that were used to calculate these LlCDF and LlLERF values. 2.3 Technical Adequacy of Probabilistic Risk Assessment Results The systematic risk assessment of debris for the resolution of GSl-191 at VEGP uses the applicable VEGP PRA models. This section provides information on the technical adequacy of the VEGP Internal Events (including internal flooding) and Seismic PRA model results in support of the resolution of GSl-191. The guidance provided in RG 1.200 (Reference 7, Section 4.2) indicates that the following items be addressed in documentation submitted to the NRC to demonstrate the technical adequacy of the PRA:

1. Identification of permanent plant changes (such as design or operational practices) that have an impact on the PRA but have not been incorporated in the PRA.

2. The parts of the PRA used to produce the results are performed consistently with the PRA Standard as endorsed by RG 1.200.

3. A summary of the risk assessment methodology used to assess the risk of the application, including how the PRA model was modified to appropriately model the risk impact of the application.

4. Identifications of key assumptions and approximations in the PRA relevant to the results used in the decision making process.

5. A discussion of the resolution of peer review or self-assessment findings and observations that are applicable to the parts of the PRA required for the application.

6. Identification of parts of the PRA used in the analysis that were assessed to have capability categories less than that required for the application.

This section provides the information to address these items. E1-7

Enclosure 1 Introduction and Overall Summary The VEGP PRA models are highly detailed, including a wide variety of initiating events, modeled systems, operator actions, and common cause events. The quantification process used for the VEGP PRA models is based on the event tree I fault tree methodology, which is a well-known methodology in the industry. The VEGP PRA models are controlled in accordance with the SNC procedure for PRA generation, maintenance and updates. The procedure defines the process for implementing regularly scheduled and interim PRA model updates, for tracking issues identified as potentially affecting the PRA models (e.g., due to changes in the plant, errors or limitations identified in the model, industry operating experience, etc.), and for controlling the model and associated computer files. To ensure that the current PRA model remains an accurate reflection of the as-built, as-operated plants, the PRA maintenance procedure requires the following activities be routinely performed:

• Design changes and procedure changes are reviewed for their impact on the PRA model on an on-going basis.
• Reliability data, unavailability data, initiating events frequency data, human reliability data, and other such PRA inputs shall be reviewed approximately every two fuel cycles and updated as necessary to maintain the PRA consistent with the as-operated plant.

As demonstrated by the information presented in the following sections, the VEGP Units 1 and 2 Internal Events and Seismic PRA models are technically adequate for the systematic risk assessment of debris for the resolution of GSl-191. 2.3.1 Plant Changes Not Yet Incorporated into the PRA Model As part of PRA model configuration control, SNC maintains a PRA model maintenance database that tracks all issues that have been identified that could impact the VEGP PRA model. Per the SNC procedure for PRA maintenance, the significance of the pending items in the database is evaluated to determine the impact on model results. Each pending item is prioritized for future model updates per its significance to model results. Based on a review of the VEGP PRA maintenance log, there are no significant outstanding changes that would impact the GSl-191 risk assessment. A summary of plant changes implemented since the cutoff date of the VEGP Seismic PRA and a qualitative assessment of the likely impact of those changes is provided in Table 1-2. E1-8

Enclosure 1 Introduction and Overall Summary Table 1 Summary of Significant Plant Changes Description of Plant Change Impact on PRA Results Safety-related battery chargers are no An assessment of this change on the VEGP longer operated in a load-share internal events PRA model indicated no configuration. Instead, a single charger significant impact. will be in service and if it fails, the other charger will be placed in service by The battery chargers are modeled as operator action. seismically correlated. Thus, modeling the change would not affect the Seismic PRA results. Permanently installed and portable Credit for FLEX equipment is likely to FLEX equipment (other than low improve the PRA results, but the impact is leakage RCP seals) have not been difficult to quantify without detailed modeling. modeled in the PRA. The Westinghouse RCP shutdown seals have been installed at VEGP, and therefore are credited in the PRA model per the guidance in PWROG-14001-P (Reference 29) although the NRC has not yet issued a Safety Evaluation for this guidance. 2.3.2 Parts of the VEGP PRA Used Version 5 of the VEGP Units 1 and 2 Internal Events PRA model is used for the GSl-191 risk assessment. The internal events PRA model is an at-power model (i.e., it addresses Modes 1 and 2 of reactor operation). The model includes both CDF and LERF from internal events, including internal flooding. Version 2 of the VEGP Units 1 and 2 seismic PRA model is used for the assessment of GSl-191 risk from seismically-induced LOCAs. These versions are part of the current VEGP PRA model of record at the time of this analysis. The latest CDF and LERF results for internal events (including internal flooding), fire, seismic, and other external hazards for VEGP Units 1 and 2 are provided in Table 1 VEGP Units 1 and 2 Internal and External Events Summary. E1-9

Enclosure 1 Introduction and Overall Summary Table 1 VEGP Units 1 and 2 Internal and External Events Summary Event Type Unit 1 CDF Unit 1 LERF Unit 2 CDF Unit 2 LERF (per/year) (per/year) (per/year) (per/year) Internal Events 2.52x10-5 7.33x10- 9 2.52x10-5 7.33x10- 9 Fire 3.86x10-5 1.39x1 o-6 4.52x10-5 1.56x1 o-6 Seismic 2.8x1 o- 6 3.3x10- 7 2.8x10- 5 3.3x10-7 Other External Screened Screened out Screened out Screened out out Total 4.39x10- 5 1.73x1o-s 5.05x1Q- 5 1.90x1Q- 6 It is noted that for VEGP Units 1 and 2, the Total CDF for internal and external events is less than 1.0x1 o-4 /year and the Total LERF is less than 1x1 o-5/year, and therefore meets RG 1.174 total risk guidelines (Reference 6). 2.3.3 Summary of the Risk Assessment Methodology The GSl-191 risk assessment methodology used for VEGP Units 1 and 2 involves quantifying the VEGP Units 1 and 2 internal events and seismic PRA models to determine the increase in risk from debris (i.e., the "risk attributable to debris"). The risk increase is defined as the difference in risk calculated considering debris effects and the risk calculated assuming debris is not present to determine both the increase in CDF (LlCDF) and the increase in LERF (LlLERF). Enclosures 2 and 3 provide a detailed description of the GSl-191 models and risk evaluation that were used to calculate LlCDF and LlLERF. 2.3.4 Key Assumptions and Approximations in the PRA Modeling uncertainties are considered in both the internal events PRA and the seismic PRA. Assumptions are made during the PRA development to address a modeling uncertainty because there is not a single definitive approach. The GSl-191 risk assessment methodology also incorporates various assumptions and approximations pertaining to modeling uncertainties. These assumptions and modeling uncertainties are reviewed to determine the impact on the GSl-191 risk assessment, as described in Enclosure 3, Section 14.2.3. 2.3.5 Assessment of PRA Model Technical Adequacy Internal Events PRA Model Numerous assessments of technical capability have been made for the VEGP Units 1 and 2 internal events PRA model: E1-10 L_ ..

Enclosure 1 Introduction and Overall Summary

• An independent PRA peer review was conducted under the auspices of the Westinghouse Owners Group 0fVOG) in 2001, following the industry PRA peer review process described in NEI 00-02 (Reference 30). This peer review included an assessment of the PRA model maintenance and update process. All "Grade B" findings (there were no "Grade A" findings) were resolved in VEGP PRA model Revision 3.
• In 2005, the VEGP PRA model results were evaluated in the WOG PRA cross-comparison study performed in support of implementing the Mitigating Systems Performance Indicator (MSPI) process. After allowing for plant-specific features there were no MSPI cross-comparison outliers for the VEGP PRA.
• In 2006, MAAP (Modular Accident Analysis Program) evaluations performed for the VEGP PRA model were reviewed by an industry MAAP expert (from Fauske Associates, Inc., the company that developed the MAAP code) to check errors and reasonableness of the MAAP results. No significant issues were found from the review.
• In 2006, a gap analysis was performed for Revision 3 of the VEGP PRA model by an independent contractor against the 2005 addenda to the 2002 version of the ASME/ANS PRA Standard and the 2004 trial use version of RG 1.200. The major gaps related to documentation (especially system notebooks), the internal flooding PRA, and the treatment of uncertainty correlations were resolved in VEGP PRA model Revision 4 in 2009.
• In 2007, during the NRG review of severe accident mitigation alternative (SAMA) analysis for VEGP license renewal, the NRG issued RAls for the VEGP PRA "L2UP" model related to dominant minimal cutsets for GDF, LERF, and other Level 2 release categories, questions about PRA quality, and Level 2 methodology. SNC provided responses to the RAls and no additional RAls were received from the NRG.
• In 2008 and 2009, as a part of MSPI margin improvement, the VEGP PRA Level 1 model was reviewed by an independent contractor (Westinghouse) to identify any excessive conservatism in the PRA model. The review concluded that there were no significant issues or excessive conservatism in the VEGP PRA that needed to be revised or refined.
• In 2008, a gap analysis was performed by an independent contractor (ERIN) for the VEGP internal flooding PRA. Issues from the gap analyses were resolved before finalizing the internal flooding PRA update.
• In 2008, a gap analysis was performed by an independent contractor (Scientech) for the human reliability analysis (HRA) and dependency analyses for post-initiator human failure events. No significant issues were found.
• An industry peer review was performed in May 2009. The Pressurized Water Reactor Owners Group (PWROG) peer review was based on the 2007 addenda to 2002 version of the ASME/ANS PRA standard and RG 1.200 Revision 1. The VEGP PRA was found to meet Capability Category II (CC-II) or better for most of the supporting requirements (SRs) in the PRA standard. The outstanding issues primarily pertained to documentation. A total of 46 facts and observations (F&Os) were identified, 11 of which were categorized as "Findings" (which were related to documentation). Seven of the F&Os recognized areas of strength in the PRA.

E1-11

Enclosure 1 Introduction and Overall Summary

• In 2011 the VEGP PRA model was reviewed along with F&Os from the 2009 peer review to determine if model changes were necessary to be able to assess the risk impact of the proposed surveillance frequency change per NEI 04-10, "Risk Informed Method for Control of Surveillance Frequencies". Open F&Os related to the systems of interest or that could potentially impact the results of the assessment were dispositioned as having no impact, incorporated into the model, or addressed with sensitivity analyses. PRA modeling changes were identified and incorporated into the model. Several components were added to the VEGP PRA model during this task.
• In 2011 the VEGP internal events PRA model (including flooding) was reviewed (along with the fire PRA model) to determine the technical capability for use in supporting the license amendment request to implement NEI 06-09, "Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines". The review demonstrated and documented that the VEGP at-power internal events PRA model (including flooding) and the fire PRA model conform to the PRA standard at CC-II which satisfies the guidance of RG 1.200, Revision 2. In addition, the VEGP PRA model complies with all requirements for technical adequacy of the baseline PRA as defined in NEI 06-09.
• In 2013 a review and update of the*MAAP parameter file was performed by an independent contractor (Fauske) to check for errors and reasonableness of MAAP results. A significant upgrade in MAAP capabilities was initiated.
• In 2014 a review was performed by an independent contractor (Reliability and Safety Consulting Engineers, Inc.) for initiating events and data update.
• In 2014 a review by an independent contractor (Scientech) was performed for the HRA and dependency analysis for post-initiator human failure events. Human error '

probabilities (HEPs) were reviewed, scenario timing verified, and a new dependency analysis was implemented.

• In 2014 a review was performed by an independent contractor (Nuenergy) which focused on the model of record and plant interface. presents the finding-level F&Os from the 2009 VEGP internal events PRA peer review F&Os, based on the 2007 addenda to the PRA standard. The resolution for each finding is described and the manner of that resolution is referenced. All VEGP internal events PRA model peer review findings are resolved. presents the additional/revised requirements associated with the most recent PRA standard (issued in 2009) as amended by RG 1.200, Revision 2. This table also describes the VEGP PRA model and documentation changes that assure consistency with the latest endorsed versions of the PRA standard and RG 1.200.

Seismic PRA Model Version 2 of the VEGP Unit 1 and 2 seismic PRA reflects the as-built and as-operated plant as of August 31, 2015. The VEGP seismic PRA model has been assessed against RG 1.200, Revision 2. Specifically, the model was subject to a self-assessment and a E1-12

Enclosure 1 Introduction and Overall Summary peer review conducted by the PWROG in November 2014. A total of 73 F&Os were identified, 46 of which were categorized as findings and 27 as suggestions. The peer review team determined that the VEGP seismic PRA model is of good quality and integrates the seismic hazard, the seismic fragilities, and the systems-analysis aspects appropriately to quantify COF and LERF. The general conclusion of the peer review was that the VEGP seismic PRA is judged to be suitable for use for risk-informed applications. After the seismic PRA peer review, the peer review finding-level F&Os were appropriately dispositioned, and the seismic PRA model was updated to reflect these dispositions and further refine several fragility values. The seismic PRA peer review conclusions, the disposition of the finding-level F&Os, and the discussion below demonstrate that the VEGP seismic PRA is technically adequate for all aspects of this submittal. Attachment 3 provides a summary of VEGP seismic PRA peer review finding-level F&Os and their disposition, as well as sensitivity analyses performed to address issues identified in the findings. 2.3.6 Capability Categories for Parts of the PRA The 2007 version of the PRA standard used for the May 2009 peer review of the VEGP internal events PRA contains a total of 327 SRs in nine technical elements and one configuration control element. Eleven of the SRs represent deleted requirements (IE-AB, IE-A9, SC-A3, SY-A9, SY-89, HR-GB, IF-A2, IF-84, IF-02, IF-E2, and QU-02), and 20 were determined to be not applicable to the VEGP PRA. Thus, a total of 296 SRs was applicable to the VEGP internal events PRA. Among the 296 applicable SRs, 99% met Capability Category II or higher, as shown in Table 1-4. Ca ability Catego Met No. of SRs  % of total applicable SRs CC-1/11/111 (or SR Met 210 70.9% CC-11/111 24 8.1 % CC-I/II 14 4.7% CC-Ill 7 2.4% CC-II 38 12.8% CC-I 0 0% SR Not Met 3 1.0% Total 296 100% Three SRs were judged to be "Not Met". These were HR-G6, QU-03, and LE-G5. Supporting requirement HR-G6 was characterized as Not Met because the reasonableness check of the HRA was done for the previous revision of the PRA and not the latest revision. Supporting requirement QU-03 was characterized as Not Met because the SR requires the PRA results to be compared with those from similar plants. The VEGP PRA report cited the MSPI benchmark report as evidence of meeting this E1-13 L___

Enclosure 1 Introduction and Overall Summary requirement, which was an outdated comparison. Supporting requirement LE-GS was characterized as Not Met because limitations of the LERF calculations that could impact risk-informed applications were not identified. Resolution of the Findings HR-G6-01, QU-03-01, and LE-GS-01 resulted in SRs HR-G6, QU-03, and LE-GS being met at a Capability Category 1/111111. Thus, the VEGP internal events PRA (including flood) meets the requirements of RG 1.200. The 2013 addenda to the 2008 version of the PRA Standard used for the November 2014 peer review of the seismic PRA contains a total of 77 SRs in three technical elements. Of the 77 SRs, a total of 67 (87%) were met at Capability Category II or higher. The 10 SRs that were judged to be "Not Met" are listed in Table 1-S, along with the associated finding-level F&Os (16 total).

               - - VEGP Se1sm1c Ta bl e 1 5             .  . PRA N0 t Me t/CC -I SRs an dAssoc1a  . te d F"m d"mgs SR                                                       Findings SHA-C4                             12-18, 12-36 SHA-H1                             12-18, 12-36 SHA-11                             12-1S SHA-12                             12-1S SHA-J1                             12-1, 12-2, 12-11, 12-16 SHA-J3                             12-8 SFR-A2                             14-1 ' 14-7' 14-1 0 SPR-82                             16-4, 16-6 SPR-84                             16-1 SPR-F1                             12-31, 16-S As the table indicates, the following 10 SRs were assessed at less than CC-II.

• Six of the SRs are related to the seismic hazard analysis (SHA), for which the seven findings pertain to: (a) inadequate documentation of the hazard analysis; (b) demonstration that sufficient consideration has been given to more recent geologic events and associated modeling; and (c) sensitivity calculations for the models and parameters used in the site hazard. The documentation items have been addressed, as noted in Attachment 3 for the affected Findings listed in Table 1-S.
• One of the SRs is related to the seismic fragility analysis (SFR). Two of the three findings associated with this SR deal with conservatisms that have now been addressed within the analytical methodology. The remaining finding is associated with a specific polar crane fragility issue, and has also been addressed within the reviewed methodology, as noted in Attachment 3 for the affected findings listed in Table 1-S.
• Three of the SRs are related to the seismic plant response (SPR) model. Three of the five findings associated with this SR are related to implementation of the seismic performance shaping factor approach in the HRA. Those findings have been addressed and implemented in the seismic PRA model, without significant impact on E1-14

Enclosure 1 Introduction and Overall Summary the results. One finding is related to the relay chatter evaluation, and is resolved in the latest model update. The last finding is related to the SPR documentation, which has been updated to resolve the issue. The information provided in this section demonstrates that the VEGP internal events and seismic PRA models meet the technical adequacy requirements of RG 1.200, Revision 2 and is of sufficient quality and level of detail to support the risk informed approach for GSl-191. 3.0 Initiating Event Frequencies The initiating events relevant to GSl-191 at VEGP are LOCAs and SSBls. The LOCA and SSBI frequencies from the base PRA model were used (with some modifications for the GSl-191 evaluation as described below). The initiating event frequencies used in the VEGP base PRA model are consistent with the requirements of the ASME/ANS PRA Standard (Reference 10) as endorsed by RG 1.200 (Reference 7), and confirmed by the VEGP base PRA model peer review. The LOCA frequencies in the base PRA model are based on the geometric mean aggregation in NUREG-1829 (Reference 8) for medium and large LOCAs and the NRC initiating event database for small LOCAs (Reference 9). Although the small break LOCA frequency is nearly an order of magnitude lower than that produced from NUREG-1829 data, it has been demonstrated that the GSl-191 risk impact is not sensitive to the initiating event frequency for small LOCAs, because these breaks are not predicted to generate enough debris to cause strainer or core failures at VEGP (see , Section 14.1 ). The parametric uncertainty associated with using the mean frequency was addressed by a sensitivity analysis using the 5th and 95th percentile frequency to calculate the GSl-191 CFPs and LlCDF (see Enclosure 3, Section 14.2.2). In addition, the uncertainty associated with the use of the geometric aggregation for medium and large LOCA frequencies was assessed by performing a sensitivity analysis using the arithmetic mean aggregation (see Enclosure 3, Section 14.2.3). The SSBI frequency in the base PRA model is based on updated industry initiating event data (Reference 11 ). For the GSl-191 evaluation, the SSBI frequency was separated for main steam line breaks (MSLBs) and feedwater line breaks (FWLBs) due to the significant difference in debris quantities that could be generated by these breaks. Based on a closer review of the FWLB frequencies, two industry events between 1987 and 1995 actually occurred outside containment, and therefore the FWLB frequency contribution was recalculated using a Jeffrey's non-informative distribution for the GSl-191 evaluation. 3.1 SSBI Frequency Allocation As discussed in Enclosure 3, Section 14.1, the CFPs for SSBls were calculated independently for MSLBs and FWLBs. This evaluation conservatively assumed that all E1-15

Enclosure 1 Introduction and Overall Summary SSBI breaks were double-ended guillotine breaks (DEGBs). Therefore, the MSLB frequency was split evenly among all of the breaks analyzed on the main steam lines to calculate the MSLB GSl-191 CFP, and the FWLB frequency was split evenly among all of the breaks analyzed on the feedwater lines to calculate the FWLB GSl-191 CFP. 3.2 LOCA Frequency Allocation For the calculation of small, medium, and large LOCA CFPs, the LOCA frequencies were allocated to individual pipe welds using a top-down distribution methodology. The top-down LOCA frequency allocation methodology essentially treats all breaks of a similar size as having an equivalent LOCA frequency regardless of the weld size and associated degradation mechanisms. For specific break sizes within the small, medium, and large LOCA categories, the PRA model LOCA frequencies were interpolated using a semi-log interpolation scheme (i.e., linear interpolation between break sizes and logarithmic interpolation between frequencies). The uncertainty associated with the top-down LOCA frequency allocation was assessed using a sensitivity analysis with different weld-specific LOCA frequency allocation weighting schemes. For this sensitivity, welds were classified as having a high, medium, or low rupture probability based on the weld-specific degradation mechanisms, and the frequency allocations were weighted accordingly (see Enclosure 3, Section 14.2.3). Pipe LOCAs were postulated at weld locations. As described in Enclosure 2 Section 3.a.1, a range of break sizes and orientations were evaluated for all in-service inspection (ISi) welds in the unisolable portion of the Class 1 pressure boundary. Non-pipe LOCAs were not explicitly evaluated. Non-pipe components whose failure would result in a LOCA include nozzles, component bodies, pressurizer heater sleeves, manways, control rod drive mechanism penetrations, safety relief valves, reactor coolant pump seals, the reactor vessel, the pressurizer vessel, the steam generator vessels, welded caps on retired lines, and other components. It was reasonably assumed that breaks at any of these non-piping components would be bounded by already-analyzed breaks at pipe weld locations. With the exception of non-pipe components that are located in the reactor cavity, all of these non-pipe components are located at or near pipe welds. For example, there are many weld locations in lines around the pressurizer vessel including the surge line, spray lines, and the safety and relief valve lines that could be used to estimate debris generated from non-pipe components in that area of containment. In addition, there are many welds distributed along the cold legs, including those near the reactor coolant pumps, that could be used to estimate debris generated from non-weld locations in those areas. The modeled welds that are located at the safe ends on the nozzles at the reactor vessel, the pressurizer vessel, and the steam generator vessels are reasonably close to the associated nozzle welds and are close enough to the vessels to produce significant debris from the insulation around those vessels. Non-pipe components associated with E1-16

Enclosure 1 Introduction and Overall Summary the reactor vessel such as control rod drive penetrations, manways, and instrument lines connected to the reactor vessel, etc., are located away from the hot and cold leg nozzles and are not near modeled pipe weld locations. However, any quantity of debris generated by non-pipe component welds located in the reactor cavity will be bounded by a reactor vessel nozzle break. 3.3 Seismically Induced LOCAs In order to evaluate the risk impact from GSl-191 due to seismically-induced LOCAs, the VEGP Internal Events PRA model that was modified to perform the risk-informed GSl-191 evaluation was used as a guide to make corresponding modifications to the VEGP seismic PRA model. The GSl-191 risk impact presented in Table 1-1 therefore includes the risk increase from seismically-induced LOCAs. Enclosure 3 provides a description of the method used to calculate the LiCDF and LiLERF values for

                                                                                         \

seismically-induced LOCAs. 3.4 Water Hammer-Induced LOCAs The approach used to demonstrate that the risk of water hammer is acceptably low is to verify that the potential for water hammer is not likely to cause pipe rupture in the break locations that can produce and transport problematic debris. The portions of the VEGP RCS that are subject to a LOCA are designed to the Class 1 requirements of Section Ill of the ASME Boiler and Pressure Vessel Code, which includes consideration of appropriate transients. The reactor coolant pressure boundary (RCPB) is designed to accommodate the system pressures and temperatures attained under the expected modes of plant operation, including anticipated transients, with stresses within applicable limits. Consideration is given to loadings under normal operating conditions and to abnormal loadings, such as pipe rupture and seismic loadings. Pressurizer piping is a primary area of consideration due to its function during RCS pressure transients. The pressurizer safety valve, including valve supports, is designed for loads due to water relief, including the passage of a water slug and the effects of water hammer. The pressurizer is also instrumented to monitor for indications of RCS leakage that would contribute to creating a water hammer condition and the VEGP Technical Specifications (TS) impose limits on RCS operational leakage. Because the RCS is kept water-solid during operation, a water-hammer event can only be introduced from one of the systems that interact with the primary loop piping. At VEGP, the only systems that flow into the primary loop piping are the safety injection (SI) system, the residual heat removal (RHR) system, and charging from the chemical and volume control system (CVCS) (References 13, 14, and 15). The potential for gas accumulation in the ECCS, which includes the CVCS, RHR, and SI sub-systems, is addressed under VEGP's response to GL 2008-01 (Reference 13). To address GL 2008-01, VEGP performed a review of site documents, procedures, and E1-17

Enclosure 1 Introduction and Overall Summary equipment, and implemented modifications and document revisions as necessary. These changes included adding vent valves, revising procedures to include ultrasonic testing for gas voids, and creating/maintaining an active program to prevent, monitor, and trend gas voids in the ECCS and CSS (References 14, 15, 25, and 26). VEGP's documented resolution of GL 2008-01 was accepted by the NRC and deemed effective in precluding gas accumulation in the ECCS and CSS, and, therefore, preventing a water hammer in these systems. (References 27 and 28). Lastly, VEGP performed a search of corrective action program data for water hammer and found no issues in systems related to GSl-191 locations of concern. Based on the fact that the piping is designed to ASME Ill Class 1 standards, the implementation of an approved gas accumulation prevention/monitoring program, and the lack of historical data for water hammer events, the relevance of water hammer events in the context of GSl-191 is deemed insignificant. Therefore, LOCA frequencies are not impacted for VEGP Units 1 & 2 due to water hammer considerations in these systems. 4.0 Defense-in-Depth and Safety Margin As described in RG 1.174 (Reference 6), sufficient defense-in-depth and safety margin must be maintained. Both of these aspects were evaluated in detail as described in . 5.0 Uncertainty Uncertainty quantification is a key requirement in RG 1.174 for a risk-informed evaluation (Reference 6). As defined in RG 1.174 and explained in more detail in NUREG-1855 (Reference 16) and two corresponding EPRI reports (References 17 and 18), there are three types of uncertainty that should be addressed:

1. Parametric uncertainty

2. Model uncertainty

3. Completeness uncertainty Parametric uncertainty refers to the variability in input parameters that are used in the risk assessment. Due to the wide range of plant-specific post-LOCA conditions related to GSl-191 phenomena, this is a very important aspect for understanding the overall uncertainty.

Model uncertainty refers to the potential variability in an analytical model when there is no consensus approach. A consensus approach is a model that has been widely adopted or accepted by the NRC for the application for which it is being used (Reference 16). For example, the use of a spherical zone of influence (ZOI) to model the debris quantity generated by a high energy break is a consensus model that has been widely adopted and accepted by the NRC (References 19 and 20). In general, the VEGP GSl-191 evaluation has been performed using standard models that have been widely accepted for deterministic evaluations (e.g., accepted ZOI sizes and prototypical E1-18

Enclosure 1 Introduction and Overall Summary strainer module testing). By using these consensus approaches, the effort to address model uncertainty is minimized. Completeness uncertainty refers to a) the uncertainty associated with scenarios or phenomena that are excluded from the risk evaluation, and b) the uncertainty associated with unknown phenomena. Although it may not be practical to quantify the uncertainty associated with factors that are not explicitly modeled, their potential impact can be qualitatively assessed. Uncertainties associated with unknown phenomena, on the other hand, cannot even be qualitatively assessed. Uncertainties associated with unknown phenomena are the reason that it is important to maintain defense-in-depth and safety margins (see Enclosure 4). Because all of the cases that were evaluated for model uncertainty and parametric uncertainty resulted in a b.CDF less than 1x1 o-6 (see Section 5.1 and Section 5.2), it can be concluded with high confidence that the risk associated with GSl-191 is very low as defined by the acceptance guidelines in RG 1.174 (Reference 6). 5.1 Parametric Uncertainty The parametric uncertainties associated with the VEGP risk-informed GSl-191 evaluation were evaluated in a very conservative manner by analyzing the worst case combinations of input parameters associated with strainer and core failures. Although the scenario is hypothetically possible, the probability of all of the worst-case conditions occurring simultaneously is extremely unlikely. As described in Enclosure 3, Section 14.2.3, the results of this evaluation showed that the parametric uncertainties are low (i.e., the resulting b.CDF still falls within RG 1.174 Region Ill even under the worst-case combination of input parameter values). 5.2 Model Uncertainty The model uncertainties were quantified using sensitivity analysis for models where no consensus exists. Sensitivities were run for the following models:

• Break model (continuum vs. DEGB-only)
• LOCA frequencies (VEGP PRA vs. NUREG-1829 arithmetic mean)
• LOCA frequency allocation to individual welds (top-down vs. degradation mechanism probability weighting)
• Containment spray (CS) actuation (CS actuates for hot leg breaks larger than 15 inches vs. CS actuating for more or fewer breaks)
• Aluminum metal release equation (UNM vs. WCAP-16530)
• Fiber bed thickness required to apply chemical head loss (0.45 inches vs.

0 inches)

• LBLOCA size range discretization (base case allocation of frequencies vs.

allocations biased to smaller break sizes and larger break sizes) E1-19

Enclosure 1 Introduction and Overall Summary As described in Enclosure 3, Section 14.2.3, the uncertainty associated with each of these models is low (i.e., the resulting ~CDF still falls within RG 1.174 Region Ill for each sensitivity that was evaluated). 5.3 Completeness Uncertainty Completeness uncertainty was qualitatively determined to be low. As described below, the VEGP evaluation was rigorous and comprehensive, and the areas that were not explicitly evaluated have a low potential for any significant risk impact:

• The range of hazards, initiating events, and plant operating modes were considered as described in Section 2.1.
• LOCAs and consequential LOCAs from SSBls were directly evaluated in the risk quantification as described in Section 2.2.

o The SSBI evaluation included an analysis of DEGBs spaced no more than 5 ft apart along each of the main steam and feedwater lines. o The LOCA evaluation included pipe breaks on every ISi weld within the Class 1 pressure boundary inside the first isolation valve. o Break sizes ranging from %-inch to a full DEGB were postulated on each weld. o Partial breaks (i.e., breaks smaller than a DEGB) were evaluated in 45-degree increment orientations around the pipe for each break size. o Debris quantities were calculated for breaks on ISi welds outside the first I I isolation valve, and there is no significant difference between the type and quantity of debris generated for these breaks compared to similar size breaks inside the first isolation valve. Due to the low probability of isolation valve failure, breaks outside the first isolation valve are insignificant with respect to GSl-191 risk at VEGP. o Non-pipe LOCAs were shown to be reasonably represented or bounded by adjacent pipe breaks as described in Section 3.2. o High likelihood equipment configurations were explicitly evaluated. o Low likelihood equipment configurations were addressed using a bounding approach.

• The risk of seismic and water hammer-induced LOCAs was shown to be low as described in Sections 3.3 and 3.4.
• As described in Section 1.0 and Enclosure 2, all known GSl-191 phenomena and debris failure mechanisms were evaluated either in a bounding manner for phenomena not explicitly included in the VEGP risk model or in a reasonably conservative manner for phenomena that were included in the risk model.

Although there is also some uncertainty associated with unknown phenomena, this uncertainty is judged to be small. The nuclear industry has been actively addressing GSl-191 concerns for PWRs for well over a decade. In addition, the boiling water reactor (BWR) strainer performance issue dates back to 1992, and unresolved safety issue (USI) A-43 dates back to 1979. Numerous tests have been performed by the U.S. E1-20

Enclosure 1 Introduction and Overall Summary NRC and industry, as well as regulators and utilities around the world over the last 35 years to resolve issues related to debris and strainer performance. This testing has investigated nearly every aspect of GSl-191 including insulation and coatings destruction from break jets; unqualified coatings failure; blowdown and washdown debris transport; containment pool settling, tumbling, and lift-over-curb debris transport; debris erosion; chemical release, solubility, and precipitation; strainer head loss, vortexing, and penetration; ex-vessel component wear; and in-vessel core blockage and boron precipitation. Based on the extensive research that has been performed, it is unlikely that there are unidentified phenomena that would significantly increase the risk of GSl-191 related failures. 6.0 Monitoring Program VEGP has implemented procedures and programs for monitoring, controlling, and assessing changes to the plant that have a potential impact on plant performance related to GSl-191 concerns. These provide the guidance to inspect the condition of the sump strainers and the ability to assess impacts to the inputs and assumptions used in the PRA and the associated engineering analysis that support the proposed change. Programmatic requirements ensure that the potential for debris loading on the sump does not materially increase. In addition, programs and procedures have been implemented to evaluate and control potential sources of debris in containment. 7.0 Quality Assurance Most of the analysis and testing for the risk-informed GSl-191 evaluation was performed as safety related under vendor quality assurance (QA) programs compliant with 10 CFR 50 Appendix B. The following exceptions are noted:

• The aluminum release equation used for the NARWHAL CFP calculation was developed through testing at the University of New Mexico (UNM). Although the testing was not performed under an Appendix B QA program, it was conducted using standard practices for academic research at the same facility where the NRG-sponsored integrated chemical effects testing (ICET) was conducted (Reference 21). The test results (including the aluminum release equation) were peer reviewed and published in a scientific journal (Reference 22). The UNM aluminum release model was qualified for safety related use at VEGP as described in Enclosure 2, Section 3.o.2.9. Finally, a sensitivity calculation was performed to address the model uncertainty associated with the use of the UNM aluminum release equation (see Enclosure 3, Section 14.2.3).
• VEGP has a relatively high containment pressure setpoint for actuating containment sprays. The design-basis calculations show that this setpoint would be exceeded for a design-basis accident. However, best-estimate thermal hydraulic calculations performed for VEGP by Texas A&M University (TAMU) showed that hot leg breaks smaller than or equal to 15 inches and cold leg breaks up to DEGBs would not initiate containment sprays. Although the TAMU E1-21

Enclosure 1 Introduction and Overall Summary thermal-hydraulics work was not performed under an Appendix B QA program, it was prepared and peer reviewed using standard practices for academic research. Note that the results of the TAMU thermal-hydraulic analysis were only used to define which breaks would initiate containment sprays, and were not used to define the pressure and temperature profiles used in the NARWHAL CFP calculation. In addition, sensitivity calculations were performed to address the model uncertainty associated with the breaks that initiate containment sprays (see Enclosure 3, Section 14.2.3).

• The fiber penetration equations used for the NARWHAL CFP calculation were developed through testing at Alden Research Laboratory. Alden has a 10 CFR 50 Appendix B QA program. Although the testing was not officially conducted under the Alden QA program, it was performed using most of the same processes, reviews, and procedures in the QA program. In addition, sensitivity calculations were performed to evaluate the sensitivity to the fiber penetration fraction (see Enclosure 3, Section 14.2.2).
• The LOCA frequencies were allocated to individual welds using a top-down approach (see Section 3.2). To address the model uncertainty, a sensitivity analysis was performed using an alternate weighting scheme based on weld-specific degradation mechanisms. This weighting scheme was derived in part using information contained in a LOCA frequency evaluation prepared by KNF Consulting Services. Although the KNF evaluation was not performed under an Appendix B QA program, it provides a reasonable set of inputs for the purpose of the sensitivity analysis.
• The GSl-191 PRA calculations are not safety related, but were prepared as safety significant under the SNC QA program.

8.0 Periodic Update of Risk-Informed Analysis The risk-informed GSl-191 analysis will be updated within at least 48 months following initial NRC approval or since the last update. This update should include all parts of the risk-informed evaluation including the systematic risk assessment, consideration of defense-in-depth, and consideration of safety margin. The update should also include any new information on LOCA frequencies that may be developed. Reliability data, unavailability data, initiating event frequency data, human reliability data, and other similar PRA inputs are reviewed approximately every two fuel cycles to maintain the base VEGP PRA model consistent with the as-operated plant. In addition, existing procedures are in place for periodic updates of risk-informed applications. 9.0 Reporting and Corrective Actions Licensees are required to make a report to the NRC and take corrective action in the event that the risk of debris exceeds the NRC acceptance criteria or in the event that defense-in-depth or safety margins have decreased from NRG-approved analysis. The risk of debris is defined in terms of ~CDF and ~LERF, and the acceptance criteria are E1-22

Enclosure 1. Introduction and Overall Summary defined as the upper threshold for RG 1.174 Region 11 (i.e., 1x10-5 for ~CDF and 1x10-5 for ~LERF) (Reference 6). Defense-in-depth measures and safety margin are specifically defined for VEGP in . Any unacceptable changes in risk or reductions in defense-in-depth or safety margins that are identified through the monitoring program (see Section 6.0), the periodic updates (see Section 8.0), or other means will be reported to the NRC. In addition, these issues will be entered and tracked to resolution in accordance with the SNC corrective action program~ 10.0 License Application The specific requirements for the license application described in RG 1.174 (Reference

6) will be addressed at a later date.

11.0 References

1. SRM-SECY-10-0113, "Staff Requirements - SECY-10-0113 - Closure Options for Generic Safety Issue - 191, 'Assessment of Debris Accumulation on Pressurized Water Reactor Sump Performance'," December 23, 2010

2. SRM-SECY-12-0093, "Staff Requirements - SECY-12-0093 - Closure Options for Generic Safety Issue - 191, 'Assessment of Debris Accumulation on Pressurized-Water Reactor Sump Performance'," December 14, 2012

3. SNC Letter NL-13-0953 to NRC (ML13137A130), "Vogtle Electric Generating Plant Proposed Path to Closure of Generic Safety lssue-191, 'Assessment of Debris Accumulation on Pressurized Water Reactor Sump Performance'," May 16, 2013

4. Draft Regulatory Guide 1.229, Revision 0, "Risk-Informed Approach for Addressing the Effects of Debris on Post-Accident Long-Term Core Cooling"

5. NRC Letter (ML073110278), "Revised Content Guide for Generic Letter 2004-02 Supplemental Responses," November 2007

6. Regulatory Guide 1.174, Revision 2, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," May 2011

7. Regulatory Guide 1.200, Revision 2, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities,"

March 2009

8. NUREG-1829, "Estimating Loss-of-Coolant Accident (LOCA) Frequencies Through the Elicitation Process," April 2008

9. U.S. Nuclear Regulatory Commission, "Reactor Operational Experience Results and Databases, Initiating Events," http://nrcoe.inel.gov/resu ltsdb/I nitEvent/

(1988-2010 Summaries) 10.ASME/ANS RA-Sa-2009, "Addenda to ASME/ANS RA-S-2008 Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications," February 2009 E1-23

Enclosure 1 Introduction and Overall Summary

11. NUREG/CR-6928, "Industry-Average Performance for Components and Initiating Events at U.S. Commercial Nuclear Power Plants," Initiating Event Data Sheets Update 2010, January 2012

12. NUREG-1903, "Seismic Considerations for the Transition Break Size," February 2008

13. Generic Letter 2008-01, "Managing Gas Accumulation in Emergency Core Cooling, Decay Heat Removal, and Containment Spray Systems," January 11, 2008

14. NL-10-0062, "Vogtle Electric Generating Plant Unit 1 Nine-Month Supplemental (Post-Outage) Response to Nuclear Regulatory Commission Generic Letter 2008-01," January 20, 2010

15. NL-08-1921, "Vogtle Electric Generating Plant Unit 2 Nine-Month Supplemental (Post-Outage) Response to Nuclear Regulatory Commission Generic Letter 2008-01," January 21, 2009

16. NUREG-1855, Revision 1, "Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decisionmaking," March 2017

17. EPRI Report 1016737, "Treatment of Parameter and Model Uncertainty for Probabilistic Risk Assessments," December 2008

18. EPRI Report 1026511, Technical Update, "Practical Guidance on the Use of Probabilistic Risk Assessment in Risk-Informed Applications with a Focus on the Treatment of Uncertainty," December 2012

19. NEI 04-07, Revision 0, "Pressurized Water Reactor Sump Performance Evaluation Methodology, 'Volume 1 - Pressurized Water Reactor Sump Performance Evaluation Methodology'," December 2004

20. NEI 04-07, Revision 0, "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'," December 2004

21. NUREG/CR-6914, Volume 1, "Integrated Chemical Effects Test Project:

Consolidated Data Report," December 2006

22. Howe, Kerry J., et al., "Corrosion and solubility in a TSP-buffered chemical environment following a loss of coolant accident: Part 1 - Aluminum," Nuclear Engineering and Design, Volume 292, October 2015: 296-305

23. NRC Bulletin 2003-01, "Potential Impact of Debris Blockage on Emergency Sump Recirculation at Pressurized-Water Reactors," June 9, 2003

24. NRC Generic Letter 2004-02, "Potential Impact of Debris Blockage on Emergency Recirculation during Design Basis Accidents at Pressurized-Water Reactors,"

September 13, 2004

25. TSTF-523 (ML13053A075), Revision 2, "Generic Letter 2008-01, Managing Gas Accumulation," February 21, 2013

26. NRC Letter (ML16063A475), "Vogtle Electric Generating Plant, Units 1 and 2, Issuance of Amendments (CAC Nos. MF6213 and MF6214)," March 21, 2016 27.NL-10-1228 (ML102140115), "Vogtle Electric Generating Plant Response to NRC Generic Letter 2008-01 Response to Request for Additional Information,"

July 28, 2010

28. NRC Letter (ML11101A097), "Vogtle Electric Generating Plants, Units 1 and 2 -

Closeout of Generic Letter 2008-01, 'Managing Gas Accumulation in Emergency E1-24

Enclosure 1 Introduction and Overall Summary Core Cooling, Decay Heat Removal, and Containment Spray Systems' (TAC Nos. MD7892 and MD7893)," April 27, 2011 29.Topical Report PWROG-14001-P, Revision 1, "PRA Model for the Generation Ill Westinghouse Shutdown Seal," July 2014

30. NEI 00-02 (ML061510619), Revision 1, "Probabilistic Risk Assessment (PRA) Peer Review Process Guidance," May 2006 E1-25

Enclosure 1 Introduction and Overall Summary Vogtle Electric Generating Plant Supplemental Response to NRC Generic Letter 2004-02 Enclosure 1 Introduction and Overall Summary Attachment 1 Resolution of the VEGP Internal Events PRA Peer Review Findings E1-26

Enclosure 1 Attachment 1 - Resolution of the VEGP Internal Events PRA Peer Review Findings F&O Review cc F&O Description F&O Resolution Reference Number Element IE-A4-01 IE-A4 CCII The SR requires a systematic evaluation of each SR IE-A4 is met at Capability Category II V-RIE-IEIF-Met system to assess the possibility of an initiating event or equivalent level per Peer Review U00-001, occurring due to failure of the system. The reviewers Report. VEGP Electric could not find documentation of such a systematic Generating review. There is no technical issue associated Plant, Initiating with this F&O. Event Additional notes made by review team in response to Notebook, SNC's comments: When the reviewers asked for the A systematic systems evaluation of each June 2014, Initiating Events (IE) notebook (NB), they were told system, including support systems was Table 2 and that Chapter 2 of the main report is the IE NB. performed to assess initiating event Table 3 and Chapter 2 does not contain any evidence that a possibility due to system failure. The Appendix E systematic evaluation of each system was performed. results of this evaluation are documented and F. Nor does Chapter 2 contain a system failure modes in Table 2 and 3 and Appendix E and F and effects analysis (FMEA) as required by the of V-RIE-IEIF-U00-001, VEGP Electric Standard which would have been an acceptable Generating Plant, Initiating Event alternate. The fact that a systematic evaluation was Notebook, June 2014. performed during the Individual Plant Examination (IPE), in of itself, is not sufficient. The evaluation Discussed in Appendix D, table D.1 - performed for the IPE should have been reviewed and (page D.3) - Map of ASME Initiating a statement to that extent should have been Events (IE) SRs to the IE notebook. presented in the Chapter 2. In absence of such evidence, the review comment stays. This F&O is resolved. As noted elsewhere in the report, it is very important to have Qood documentation. IE IE-01 cc The lack of a central place for all the information SR IE-01 is met at Capability Category II V-RIE-IEIF-01 1/11/111 related to initiating events made it difficult for the or equivalent level per Peer Review U00-001, Met review team to review this topic. Most plants have all Report. VEGP Electric this information stored in a separate IE notebook. Generating The review team recommends that VEGP do the There is no technical issue associated Plant, Initiating same. with this F&O. Event Notebook, Additional notes in response to SNC's comments: The A review and update of the VEGP June 2014. review team disagrees with SNC's comments. The initiating events was completed in June Standard requires that the work be documented in a 2014. An IE Notebook was developed as manner that facilitates PRA applications, upgrades RIE calculation V-RIE-IEIF-U00-001 and peer review. The review team does not believe Initiating Events. The updated analysis that the work was documented a manner to facilitate notebook includes all the information E1-27

Enclosure 1 Attachment 1 - Resolution of the VEGP Internal Events PRA Peer Review Findings F&O Review cc F&O Description F&O Resolution Reference Number Element peer review. One could almost make a case for 'not related to the initiating events. met' categorization for this element as the Documentation of special initiators and documentation is the weakest link in this whole effort. system dependency analysis is included The F&O stays as written. in appendix E and F respectively. Appendix D provides mapping of ASME IE SRs to locations in the IE notebook. This F&O is resolved. AS-A11- AS-A11 cc Dependencies are not preserved for consequential SR AS-A 11 is met at Capability Category Documented 01 1/11/111 ATWS for the SLOCA initiator and the SGTR initiator. II or equivalent level per Peer Review in PRA-8C-V-Met The existing A TWS trees, based on a LOFW initiator, Report. 07-003 Rev were developed for transients that do not include a 4.0VEGP loss of RCS inventory or operator actions to mitigate a In the revised model, if ATWT occurred Internal Event SGTR. after a SLOCA or SGTR, the accident PRA model, sequence is treated in a similar way as a Chapter 5, Note: The review team decided to leave the F&O as is case with a stuck open PZR safety Section 5.2, after reviewing SNC's comments. valve(s) where inventory make up by item 27. high pressure injection and recirculation (Reference 4) as well as secondary heat removal is required to prevent core damage. This F&O is resolved. SY SY-83 cc The treatment of main or frontline system and SR SY-83 is met at Capability Category V-RIE-IEIF-01 1/11/111 supporting or mitigating system Common Cause II or equivalent level per Peer Review U00-013, Met Failure (CCF) event groupings do not appear to Report. VEGP Electric consistent in the current PRA documents. It appears Generating that for systems considered non-risk significant, CCF documentation is revised. Plant, reviews for CCF groups may not have been Common undertaken since the IPE modeling. A separate CCF Notebook (V-RIE-IEIF- Cause Factor U013 Common Cause Factors) was Notebook. The updated standards require a systematic treatment created. Potential CCF groups (for both of all systems, not just the main systems contributing risk significant and non-risk significant to core damage. New CCF groups may be required systems) are considered and or updated documentation as to why these groups are documented as identified. not required is needed. This F&O is resolved. Note: The review team decided to leave the F&O as is after reviewing SNC's comments. E1-28

Enclosure 1 Attachment 1 - Resolution of the VEGP Internal Events PRA Peer Review Findings F&O Review cc F&O Description F&O Resolution Reference Number Element HR-G6- HR-G6 Not Check of consistency and review for reasonableness This F&O is resolved. Reasonableness PRA-BC-V 01 Met is missing in the Revision 4 updated HRA draft and check for all HRAs for Revision 4 model 003 Rev 4.0, the prior revision document information related to was re-performed. All HRAs have been Chapter 8, these items is not appropriate to use in light of the determined to be reasonable or have Human updates performed and changes to the results. been appropriately revised. The Reliability Section 8 includes a table of human failure events reasonableness check is documented in Analysis for (HFEs) and human error probabilities (HEPs) but Section 8.2.2 of PRA-BC-V-07-003, VEGP PRA does not include HEP reasonableness check, as is Human Reliability Analysis for VEGP Model, Section documented in Section 8.3 of the November 2005 PRA Model Rev. 4.0 (Reference 4). 8.2.2. HRA update for Revision 3. This F&O is resolved. DA-C2- DA-C2 cc Generic data alone was used for the probability that a SR DA-C2 is met at Capability Category PRA-BC-V 01 1/11/111 PORV is blocked -- refer to Table 6.3.9. Since PORV II or equivalent level per Peer Review 003 Rev. 4.0 Met availability is a critical plant feature with respect to Report. VEGP Internal A TWS pressure control, the use of generic data for Events PRA this parameter is deemed a weakness. VEGP specific data was used for the model, probability that a PORV is blocked - Chapter 6, refer to Table 6.4-1 in Chapter 6, VEGP VEGP Data, Data, PRA-BC-V-07-003 Rev. 4.0 VEGP Table 6.4-1. Internal Events PRA model. This F&O is resolved. E1-29

Enclosure 1 Attachment 1 - Resolution of the VEGP Internal Events PRA Peer Review Findings F&O Review cc F&O Description F&O Resolution Reference Number Element IF-C2a- IF-C2a cc Because of a lack of well documented analysis, a lot SR IE-C2a is met at Capability Category V-RIE-IEIF-01 11111111 of information had to be obtained by talking to the II or equivalent level per Peer Review U00-008, Rev. Met analyst who performed the analysis, which is the Report. 1, Internal basis for the F&O. Flooding The VEGP Internal Flooding analysis Notebook Original F&O: From a more detailed review of the IPE uses the internal flooding assessment VEGP Electric flood calculations (which are the main input to defining conducted by ABS Consulting, and PRA- Generating the flood events and consequences), it is noted that BC-V-07-003, Rev.4.0 VEGP Internal Plant Internal successful operator mitigation of ALL flood events is Events PRA model, Chapter 5, Linked Flooding PRA, assumed to occur 30 minutes into any flood scenario Fault Tree, Section 5.3 Internal Flooding Rev. 2, dated and fully terminate the flood flow, and it appears to be IE Integration. These documents were 01/2009, based on assumptions only, as no detailed discussion combined into V-RIE-IEIF-U00-008, Rev. VEGP Design of the actual ability of operators to perform such 1, Internal Flooding Notebook. Manual# DC-actions is given. This appears to be in direct conflict Calculation V-RIE-IEIF-U00-008, Rev. 1, 1009 Flooding with the HFEs included (but not modeled in the PRA Internal Flooding Notebook resolves the lnterdiscipline, model) in the flooding report (assumed perfect comment of "lack of well documented PRA-BC-V response vs. HFE calculation). Also, the report lists analysis". 003, Rev.4.0 hundreds of pages of a detailed analysis approach Automatic and operator actions that have VEGP Internal using screening criteria, flow calculations, etc. and the ability to terminate or contain floods Events PRA only by locating very specific statements. are identified in V-RIE-IEIF-U00-008, model, Rev. 1, Internal Flooding Notebook, Chapter 5, There appear to be conflicts of the inputs to the Section 10, Evaluate Flood Mitigation Linked Fault flooding PRA and the subsequent discussions of Strategies. Actions to mitigate the flood Tree, Section operator mitigation as well as using the information are generally not credited. 5.3 Internal from the IPE calculations for propagation Flooding IE assessments. This is more than an editorial finding Bounding assumptions about flood Integration, and impacts the entire basis of using the older heights, propagation, and impact on VEGP Level 1 calculated results in the current analysis. equipment have been made (V-RIE-IEIF- PRA for At-U00-008, Rev. 1, Internal Flooding Power and Additional notes made in response to SN C's Notebook, Section 9.1, Flood Internal Floods comments: The lengthy flooding methodology outlined Characterization). Engineering (NRC). in the report is not used in the current VEGP flooding calculations for design basis flood results as mentioned in the original F&O. The conditions have been performed for each previous IPE flooding analysis is used as inputs to the flood area, but these calculations are not flooding targets and propagations for a bounding case directly referenced in the flooding estimation, and the more thorough analysis outlined in analysis. The flood water flow was the report is not undertaken but is in place for use (as successfully isolated at 30 minutes, and was explained by the SNC analyst in charge of the all calculations for flood volumes, flooding oroiect durinq the peer review and verv briefly oropaqations, etc. were done with the E1-30

Enclosure 1 Attachment 1 - Resolution of the VEGP Internal Events PRA Peer Review Findings F&O Review cc F&O Description F&O Resolution Reference Number Element mentioned in the flooding report). amount of water generated in this 30 minutes with considerations for system Each and every IPE flooding calculation reviewed characteristics". during the peer review contained the assumption that the flood water flow was successfully isolated at 30 While these actions are not explicitly minutes, and all calculations for flood volumes, modeled, the flooding model uses propagations, etc. were done with the amount of bounding assumptions with respect to water generated in this 30 minutes with operator actions (V-RIE-IEIF-U00-008, considerations for system characteristics. If any IPE Rev. 1, Internal Flooding Notebook, flooding calculations are done, which do not contain Section 9.1, Flood characterization). This this assumption, they were not seen during the peer resolves the comment "successful review. operator mitigation of ALL flood events is assumed to occur 30 minutes into any No change to the F&O is warranted. The problem with flood scenario and fully terminate the this scenario of using the IPE flooding calculations for flood flow". inputs to the described methodology is the following: If the flooding analysis was performed in accordance to The calculations were reviewed by ABS the methodology outlined in the current report, new and RIE analysts and an independent flooding volumes and propagation assessments would plant walkdown was performed. These be required that did not take into account successful activities support that model isolation at 30 minutes (as the IPE calculations do) assumptions are conservative. since another operator isolation assessment is outlined in the flooding report methodology for normal There is no technical issue associated HFE calculations for flow isolation. with this F&O. This F&O is resolved. QU QU-03 Not Reviewer asked the VEGP staff to provide evidence A new comparison study was performed V-RIE-IEIF-01 Met of comparison of the VEGP results to those from by comparing VEGP PRA results with U00-001, similar plants. The VEGP staff presented the two PWR PRAs (Callaway and Wolf Initiating benchmark report for MSPI as evidence of Creek), which are considered relatively Events comparison. Reviewers concluded similar to VEGP. In addition to the Notebook, that report is not sufficient evidence for demonstrating comparison of PRA reports, a plant visit Section 2, compliance to this SR. to Callaway was performed to identify Table 4 more details of Callaway systems and PRA modeling. The plant comparisons were again E1-31

Enclosure 1 Attachment 1 - Resolution of the VEGP Internal Events PRA Peer Review Findings F&O Review cc F&O Description F&O Resolution Reference Number Element updated with the 2014 initiating event update and V-RIE-IEIF-U00-001 Initiating Events Notebook was updated to reflect the changes. The comparison showed that all plants have loss of offsite power (LOSP)/station blackout (SBO) as the most dominant contributors which indicated that the VEGP PRA results are not an outlier, as compared to similar PWRs. Differences in dominant CDF contributors were investigated, and it was found that those differences are due to differences in details of system configuration/ operation and physical barriers for internal flooding and in the sources for generic initiating event frequency data (VEGP PRA used the latest generic (2010 - 2013) initiating frequency and failure data along with VEGP specific experience data for its data update). This F&O is resolved. QU-F5- QU-F5 cc In Chapter 10, there is insufficient documentation for SR QU-F5 is met at Capability Category PRA-BC-V-01: 1/11/111 the quantification process, which would impact II or equivalent level per Peer review 007-003, Met application (only EOOS). Reviews conclude that the Report. VEGP Internal documentation currently in Chapter 10 is not sufficient Event PRA, to meet this SR fully. Chapter 10 of PRA-BC-07-003 (VEGP Rev4.0, internal PRA model rev 4.) contains Chapter 10, detailed information for quantification VEGP PRA process and also identifies limitations Level 1 and which would affect application. The Level2 identified limitation in quantification Evaluation, process is that the average model Recovery assumed a specific plant system Analysis and alignment configuration. This assumed Uncertainty system alignment configuration is Analysis. chanQed to reflect actual system E1-32

Enclosure 1 Attachment 1 - Resolution of the VEGP Internal Events PRA Peer Review Findings F&O Review cc F&O Description F&O Resolution Reference Number Element alignment configuration when configuration specific risk is evaluated. There is no other limitation in quantification process. This F&O is resolved. LE-GS- LE-GS Not Limitations in the LERF analysis that would impact A comparison of VEGP LERF scenarios PRA-BC-V 01 Met applications are not identified. The LERF analysis with those in Table 4.S.9.3 of the ASME 003, VEGP documentation is incomplete because limitations in PRA standard revealed that the VEGP Internal Event the LERF analysis that would impact applications, as PRA included more potential LERF PRA, Rev 4.0, required by SR LE-GS, are not identified. scenarios than as required for a large dry Chapter 9, containment plant in ASME PRA VEGP Level 2 standard. PRA Modeling. The LERF scenarios modeled in VEGP PRA include containment bypass core damage scenarios (steam generator tube rupture and Interfacing systems LOCA), thermally or pressure induced steam generator tube rupture after core damage, containment isolation failure with core damage and various early containment failure modes. This F&O is resolved. MU MU-84 cc The VEGP plant procedures do not specifically call for Procedure RIE-014, Configuration See procedure 01 11111111 a peer review after a PRA upgrade has been Management of PRA Models, Qualitative RIE-014, Met completed. But the plant has had this peer review and Models and Software outlining Configuration other peer reviews in the past. This change is requirements dealing with PRA Management required by the SR. configuration control, as referenced in of PRA ASME/ANS RA-Sa-2009, Section 1-S, Models, have been developed to comply with Qualitative requirements of RITS Initiative 4b. Models and Software, Procedure RIE-001, Generation and Section 4.2 Maintenance of PRA Models and Associated Updates, Section 4.2.8 states See procedure the requirement for either a full or RIE-001, focused peer review. Generation E1-33

Enclosure 1 Attachment 1 - Resolution of the VEGP Internal Events PRA Peer Review Findings F&O Review cc F&O Description F&O Resolution Reference Number Element and The PRA model update process is Maintenance discussed in Section 5 of this licensing of PRA submittal. Models and Associated This F&O is resolved Updates, Section 4.2.8. AS ME/ANS RA-Sa-2009, Section 2-3 and Appendix 1-A. NEI 00-02, PRA Peer Review Process Guidance. E1-34

Enclosure 1 Introduction and Overall Summary Vogtle Electric Generating Plant Supplemental Response to NRC Generic Letter 2004-02 Enclosure 1 Introduction and Overall Summary Attachment 2 Comparison of RG 1.200 Revision 1 and Revision 2 Supporting Requirements E1-35

Enclosure 1 Attachment 2 - Comparison of RG 1.200 Revision 1 and Revision 2 Supporting Requirements SR in 2007 PRA Standard as SR in 2009 PRA Standard as Description of Change Resolution Amended by RG 1.200, Revision 1 Amended by RG 1.200, Revision 2 IE-C10: IE-C12: The sentences were NUREG/CR-6928 is used as clarifications provided in RG the source for generic data CC-1/11/111: CC-1/11/111: 1.200 Revision 1 and Revision 2, priors in Revision 5 of the ... ... respectively . VEGP internal events PRA. An example of an acceptable generic An example of an acceptable generic data sources is NUREG/CR-5750 data sources is NUREG/CR-6928 The updated SR cites a more [Note 1]. [Note 1]. recent example of an acceptable Qeneric data source. SY-815: SY-814: The sentences were As noted in Table 9.2-1 of the clarifications provided in RG internal events PRA CC-1/11/111: CC-1/111111: 1.200 Revision 1 and Revision 2, calculation, failure of the ... ... respectively . containment boundary due to (h) harsh environments induced by (h) harsh environments induced by venting is not applicable to the containment venting, or failure that containment venting, failure of the The updated SR explicitly VEGP large, dry, sub-may occur prior to the onset of core containment venting ducts, or failure requires consideration of atmospheric containment. damage. of the containment boundary that containment venting ducts and may occur prior to the onset of core failure of the containment damage boundary prior to core damage. DA-C1: DA-C1: Reference NUREG-1715 was NUREG/CR-6928 is used as added by RG 1.200 Revision 1; the source for generic data CC-1/11/111: CC-1/11/111: References NUREG-1715 and priors in Revision 5 of the ... . .. NUREG/CR-6928 were included VEGP internal events PRA. Examples of parameter estimates Examples of parameter estimates in the 2009 version of the PRA and associated sources include: and associated sources include Standard. (a) component failure rates and (a) component failure rates and probabilities: NUREG/CR-4639 [Note probabilities: NUREG/CR-4639 [2-7], The updated SR cites more (1)], NUREG/CR-4550 [Note (2)], NUREG/CR-4550 [2-3], NUREG- recent examples of acceptable NUREG-1715 [Note 71 1715 [2-21], NUREG/CR-6928 [2-20] generic data sources. QU-A2a: QU-A2: The LERF requirement was Section 10.3.2 of the internal added by RG 1.200 Revision 2. events PRA calculation CC-1/11/111: CC-1/11/111: presents estimates for PROVIDE estimates of the individual PROVIDE estimates of the individual The updated SR explicitly individual LERF sequence sequences in a manner consistent sequences in a manner consistent requires consideration of LERF. cutsets. with the estimation of total CDF ... with the estimation of total CDF (and LERF) ... E1-36

Enclosure 1 Attachment 2 - Comparison of RG 1.200 Revision 1 and Revision 2 Supporting Requirements SR in 2007 PRA Standard as SR in 2009 PRA Standard as Description of Change Resolution Amended by RG 1.200, Revision 1 Amended by RG 1.200, Revision 2 aU-A2b: aU-A3: The phrase, "from internal The peer review based on the events", was deleted from the 2007 version of the PRA CC-I: CC-I: 2009 version of the PRA Standard addressed these ESTIMATE the point estimate CDF ESTIMATE the point estimate CDF Standard. The LERF LERF requirements. Section from internal events. (and LERF). requirement was added by RG 10.3.2 of the internal events 1.200 Revision 2. PRA calculation presents the CC-II: mean CDF LERF results. ESTIMATE the mean CDF from CC-II: The SR explicitly requires internal events, accounting for the ESTIMATE the mean CDF (and consideration of LERF. "state-of-knowledge" correlation LERF), accounting for the state-of- However, per the note in 2007 between event probabilities [Note knowledge correlation between SR LE-E4 and LE-F3, LERF was (1 )]. event probabilities [Note (1)]. addressed in applicable requirements of Table 4.5.8, CC-Ill: which includes all au SRs. CALCULATE the mean CDF from CC-Ill: Thus, the peer review using the internal events by propagating the CALCULATE the mean CDF (and 2007 version of the PRA uncertainty distributions, ensuring LERF) by propagating the Standard addressed these LERF that the "state-of-knowledge" uncertainty distributions, ensuring requirements. correlation between event that the state-of-knowledge probabilities is taken into account. correlation between event probabilities is taken into account. aU-B6: aU-B6: The LERF requirement was The peer review based on the added by RG 1.200 Revision 2. 2007 version of the PRA CC-1111/111: CC-1/111111: Standard addressed these ACCOUNT for system successes in ACCOUNT for system successes in The SR explicitly requires LERF requirements. The addition to system failures in the addition to system failures in the consideration of LERF. Level 2 PRA event trees evaluation of accident sequences to evaluation of accident sequences to However, per the note in 2007 presented in Section 9.2 of the the extent needed for realistic the extent needed for realistic SR LE-E4 and LE-F3, LERF was internal events PRA calculation estimation of CDF. This accounting estimation of CDF or LERF. This addressed in applicable explicitly account for system may be accomplished by using accounting may be accomplished by requirements of Table 4.5.8, successes. numerical quantification of success using numerical quantification of which includes all au SRs. probability, complementary logic, or a success probability, complementary Thus, the peer review using the delete term approximation and logic, or a delete .term approximation 2007 version of the PRA includes the treatment of transfers and includes the treatment of Standard addressed these LERF among event trees where the transfers among event trees where requirements. "successes" may not be transferred the "successes" may not be between event trees. transferred between event trees. E1-37

Enclosure 1 Attachment 2 - Comparison of RG 1.200 Revision 1 and Revision 2 Supporting Requirements SR in 2007 PRA Standard as SR in 2009 PRA Standard as Description of Change Resolution Amended by RG 1.200, Revision 1 Amended by RG 1.200, Revision 2 QU-E3: QU-E3: The LERF ~equirement was The peer review based on the added by RG 1.200 Revision 2. 2007 version of the PRA CC-I: CC-I: Standard addressed these ESTIMATE the uncertainty interval of ESTIMATE the uncertainty interval of The SR explicitly requires LERF requirements. Section the CDF results. Provide a basis for the CDF (and LERF) results. consideration of LERF. 10.4 of the internal events PRA the estimate consistent with the Provide a basis for the estimate However, per the Note in 2007 calculation presents the characterization parameter consistent with the SR LE-E4 and LE-F3, LERF was uncertainty intervals for both uncertainties (DA-D3, HR-D6, HR- characterization parameter addressed in applicable CDF and LERF, with GB, IE-C15). uncertainties (DA-D3, HR-06, HR- requirements of Table 4.5.8, consideration of the state-of-GB, IE-C15). which includes all QU SRs. knowledge correlation. CC-II: Thus, the peer review using the ESTIMATE the uncertainty interval of CC-II: 2007 version of the PRA the CDF results. ESTIMATE the ESTIMATE the uncertainty interval of Standard addressed these LERF uncertainty intervals associated the CDF (and LERF) results. requirements. with parameter uncertainties (DA- ESTIMATE the uncertainty D3, HR-D6, HR-GB, IE-C15), taking intervals associated with into account the state-of- parameter uncertainties (DA-D3, knowledge correlation. HR-D6, HR-GB, IE-C15), taking into account the state-of-knowledge CC-Ill: correlation. PROPAGATE parameter uncertainties (DA-D3, HR-D6, HR- CC-111: GB, IE-C15) .... (no change) PROPAGATE parameter uncertainties (OA-03, HR-D6, HR-GB, IE-C15) .... (no change) E1-38

Enclosure 1 Attachment 2 - Comparison of RG 1.200 Revision 1 and Revision 2 Supporting Requirements SR in 2007 PRA Standard as SR in 2009 PRA Standard as Description of Change Resolution Amended by RG 1.200, Revision 1 Amended by RG 1.200, Revision 2 QU-E4: QU-E4: Separate requirements for CC-I, No action, CC-II met for 2007 II, and Ill were collapsed into a version of the PRA Standard. CC-I: CC-1/11/111: single requirement for CC-1/11/111 PROVIDE an assessment of the For each source of model in the 2009 version of the PRA impact of the model uncertainties uncertainty and related assumption Standard. The reference to Note and assumptions on the results of identified in QU-E1 and QU-E2, 1 was deleted by RG 1.200 the PRA. respectively, IDENTIFY how the Revision 2. PRA model is affected (e.g., CC-II: introduction of a new basic event, The updated SR assigns the EVALUATE the sensitivity of the changes to basic event probabilities, same requirement to all three results to model uncertainties and change in success criterion, CCs. Meeting CC-II: in the 2007 key assumptions using sensitivity introduction of a new initiating version of the PRA Standard analyses [Note (1)]. event). assures that the new SR is met. CC-Ill: EVALUATE the sensitivity of the results to uncertain model boundary conditions and other assumptions using sensitivity analyses except where such sources of uncertainty have been adequately treated in the quantitative uncertainty analysis [Note (1 )]. E1-39

Enclosure 1 Attachment 2 - Comparison of RG 1.200 Revision 1 and Revision 2 Supporting Requirements SR in 2007 PRA Standard as SR in 2009 PRA Standard as Description of Change Resolution Amended by RG 1.200, Revision 1 Amended by RG 1.200, Revision 2 LE-F2: LE-F3: Separate requirements for CC-I, No action, CC-II met for 2007 11, and Ill were collapsed into a version of the PRA Standard. CC-I: CC-1/11/111: single requirement for CC-1/11/111 PROVIDE a qualitative assessment IDENTIFY and CHARACTERIZE the in the 2009 version of the PRA of the key sources of uncertainty. LERF sources of model uncertainty Standard. Examples: and related assumptions, in a (a) Identify bounding manner consistent with the The updated SR assigns the assumptions. applicable requirements of Tables 2- same requirement to all three (b) Identify conservative treatment 2.7-2(d) and 2-2.7-2(e). CCs. Meeting CC-II: in the 2007 of phenomena. version of the PRA Standard assures that the new SR is met. CC-II: PROVIDE uncertainty analysis that identifies the key sources of uncertainty and includes sensitivity studies for the significant contributors to LERF. CC-Ill: PROVIDE uncertainty analysis that identifies the key sources of uncertainty and includes sensitivity studies. E1-40

Enclosure 1 Attachment 2 - Comparison of RG 1.200 Revision 1 and Revision 2 Supporting Requirements SR in 2007 PRA Standard as SR in 2009 PRA Standard as Description of Change Resolution Amended by RG 1.200, Revision 1 Amended by RG 1.200, Revision 2 IF-F2: IFPP-82: The requirement to document Section 5 and Appendix A of walkdowns performed in support the internal flooding PRA CC-11111111: CC-11111111: of plant partitioning was added document the walkdowns DOCUMENT the process used to DOCUMENT the process used to to the 2009 version of the PRA performed to validate identify ... flood areas, ... For identify flood areas. For example, Standard. information related to flood example, this documentation typically this documentation typically includes areas, flood sources, SSCs, includes , The updated SR cites examples mitigation and other flood ... of acceptable documentation of related features in the flood (b) flood areas used in the analysis (a) flood areas used in the analysis the process to identify flood areas. and the reason for eliminating areas and the reason for eliminating areas sources. from further analysis from further analysis ... (b) any walkdowns performed in Since documentation of support of the plant partitioning walkdowns was not in the 2007 version of the PRA Standard, it was not reviewed as part of the peer review conducted using that version of the PRA Standard. E1-41

Enclosure 1 Attachment 2 - Comparison of RG 1.200 Revision 1 and Revision 2 Supporting Requirements SR in 2007 PRA Standard as SR in 2009 PRA Standard as Description of Change Resolution Amended by RG 1.200, Revision 1 Amended by RG 1.200, Revision 2 IF-81: IFSO-A1: The requirement to include the Potential flood sources fire protection system in Item (a) identified in Section 5 of the CC-1111/111: CC-11111111: as a potential flooding source internal flooding PRA reviewed For each flood area, IDENTIFY the For each flood area, IDENTIFY the was added by RG 1.200 as part of 2009 peer review potential sources of flooding [Note potential sources of flooding [Note Revision 1. This requirement against 2007 version of the (1)]. INCLUDE: (1)]. INCLUDE was addressed in the peer PRA standard amended by RG (a) equipment (e.g., piping, valves, (a) equipment (e.g., piping, valves, review, which used the 2007 1.200, Revision 1 include pumps) located in the area that are pumps) located in the area that are version of the PRA Standard RCS-connected systems - connected to fluid systems (e.g., connected to fluid systems (e.g., amended by RG 1.200 Revision chemical and volume control circulating water system, service circulating water system, service 1. system (CVCS), containment water system, fire protection system, water system, fire protection system, spray (CS), residual heat component cooling water system, component cooling water system, The requirement to include the removal (RHR), reactor coolant feedwater system, condensate and feedwater system, condensate and reactor coolant system in Item system drain tank (RCSDT), steam systems) steam systems, and reactor coolant (a) as a potential flooding source safety injection (SI), and ... system) was added to the 2009 version reactor water makeup system

                                     ...                                    of the PRA Standard. Thus, it      (RMWS). The Containment was not reviewed as part of the    Building (and RCS peer review conducted using        components therein) is not that version of the PRA            included in the scope of the Standard.                          internal flooding analysis.

E1-42

Enclosure 1 Attachment 2 - Comparison of RG 1.200 Revision 1 and Revision 2 Supporting Requirements SR in 2007 PRA Standard as SR in 2009 PRA Standard as Description of Change Resolution Amended by RG 1.200, Revision 1 Amended by RG 1.200, Revision 2 IF-F2: IFS0-82: The requirement to document Section 5 and Appendix A of walkdowns performed in support the internal flooding PRA CC-1/11/111: CC-11111111: of the identification or screening document the walkdowns DOCUMENT the process used to DOCUMENT the process used to of flood sources was added to performed to validate identify applicable flood sources. For identify applicable flood sources. For 2009 version of the PRA information related to flood example, this documentation typically example, this documentation Standard. areas, flood sources, SSCs, includes typically includes mitigation and other flood (a) flood sources identified in the (a) flood sources identified in the The updated SR cites examples related features in the flood analysis, rules used to screen out analysis, rules used to screen out of acceptable documentation of areas. these sources, and the resulting list these sources, and the resulting list the process to identify flood of sources to be further examined of sources to be further examined sources . (f) screening criteria used in the Since documentation of analysis (b)screening criteria used in the walkdowns was not in the 2007 ... analysis version of the PRA Standard, it (j) calculations or other analyses was not reviewed as part of the used to support or refine the flooding (c) calculations or other analyses peer review conducted using evaluation used to support or refine the flooding that version of the PRA ... evaluation Standard. (d) any walkdowns performed in support of the identification or screeninQ of flood sources E1-43

Enclosure 1 Attachment 2 - Comparison of RG 1.200 Revision 1 and Revision 2 Supporting Requirements SR in 2007 PRA Standard as SR in 2009 PRA Standard as Description of Change Resolution Amended by RG 1.200, Revision 1 Amended by RG 1.200, Revision 2 IF-F2: IFSN-82: The requirement to document Section 5 and Appendix A of walkdowns performed in support the internal flooding PRA CC-1/11/111: CC-1/11/ll I: of the identification or screening document the walkdowns DOCUMENT the process used to DOCUMENT the process used to of flood scenarios was added to performed to validate identify applicable flood scenarios. identify applicable flood scenarios. 2009 version of the PRA information related to flood For example, this documentation For example, this documentation Standard. areas, flood sources, SSCs, typically includes typically includes mitigation and other flood ... The updated SR cites examples related features in the flood (c) propagation pathways ... (a) propagation pathways ... of acceptable documentation of areas. ... the process to identify flood (d) accident mitigating features and (b) accident mitigating features and scenarios. barriers credited ... barriers credited ... ... Since documentation of (e) assumptions or calculations used (c) assumptions or calculations used walkdowns was not in the 2007 in the determination of ... flood- in the determination of ... flood- version of the PRA Standard, it induced effects on equipment induced effects on equipment was not reviewed as part of the operability operability peer review conducted using ... that version of the PRA (f) screening criteria used in the (d) screening criteria used in the Standard. analysis analysis (g) flooding scenarios considered, (e) flooding scenarios considered, screened, and retained screened, and retained (h) description of how the internal (f) description of how the internal event analysis models were modified event analysis models were modified (j) calculations or other analyses (g) calculations or other analyses used to support or refine the flooding used to support or refine the flooding evaluation evaluation ... (h) any walkdowns performed in support of the identification or screening of flood scenarios E1-44

Enclosure 1 Attachment 2 - Comparison of RG 1.200 Revision 1 and Revision 2 Supporting Requirements SR in 2007 PRA Standard as SR in 2009 PRA Standard as Description of Change Resolution Amended by RG 1.200, Revision 1 Amended by RG 1.200, Revision 2 IF-F2: IFQU-82: The requirement to document Section 5 and Appendix A of walkdowns performed in support the internal flooding PRA CC-1/11/111: CC-1/111111: of internal flood accident document the walkdowns DOCUMENT the process used to DOCUMENT the process used to sequence quantification was performed to validate define the applicable internal flood define the applicable internal flood added in 2009 version of the information related to flood accident sequences and their accident sequences and their PRA Standard. areas, flood sources, SSCs, associated quantification. For associated quantification. For mitigation and other flood example, this documentation typically example, this documentation The updated SR cites examples related features in the flood includes typically includes of acceptable documentation of areas that are considered in ... (a) calculations or other analyses the process to identify flood flood sequence definition. (j) calculations or other analyses used to support or refine the flooding related features considered in used to support or refine the flooding evaluation flood sequence quantification. evaluation ... (b) screening criteria used in the Since documentation of (f) screening criteria used in the analysis walkdowns was not in the 2007 analysis version of the PRA Standard, it ... was not reviewed as part of the (i) flooding scenarios considered, (c) flooding scenarios considered, peer review conducted using screened, and retained screened, and retained that version of the PRA ... Standard . (k) results of the internal flood (d) results of the internal flood analysis, consistent with the analysis, consistent with the quantification requirements provided quantification requirements provided in HLR-QU-D in HLR-QU-D (e) any walkdowns performed in support of internal flood accident sequence quantification E1-45

Enclosure 1 Introduction and Overall Summary Vogtle Electric Generating Plant Supplemental Response to NRC Generic Letter 2004-02 Enclosure 1 Introduction and Overall Summary Attachment 3 Resolution of the VEGP Seismic PRA Peer Review Findings E1-46

Enclosure 1 Attachment 3 - Resolution of the VEGP Seismic PRA Peer Review Findings F&O Review cc F&O F&O F&O Disposition for GSl-191 Number Element Description Basis Resolution and Other Applications 11-3 SHA-E2 111111 While variability in To maintain hazard-consistent Expand documentation to There is an abundance of the mean base-case ground motion hazard at the demonstrate that a single site-specific Vs data from Vs profile is control point, the site response base-case Vs profile VEGP Units 3&4, which incorporated in the analysis needs to incorporate adequately represents reduces epistemic site response appropriate epistemic uncertainty the Units 1&2 site. Or if uncertainty to an analysis, no and aleatory variability in its that is not the case, insignificant level. epistemic inputs. The Vs profile for the include epistemic Additional discussion of uncertainty in the Vogtle Units 1&2 site is uncertainty in the the rationale for use of a base-case profile is represented by a single Vs profile, characterization of Vs single base-case Vs profile represented. indicating there is no epistemic profile and evaluate the for the site has been Documentation of uncertainty in the mean base- impact on control point included in the the justification for case profile. Documentation of ground motions. documentation. The added this assessment this assessment needs to be discussion demonstrates should be expanded. that a single base-case expanded. shear-wave velocity (Vs) Discussion with staff indicates profile adequately (This F&O originated that consideration of the represents the Vogtle site, from SR SHA-E2) combined data for the Vogtle site based on the availability of (Units 1&2, Units 3&4, ISFSI) Vs data, which reduces the provides sufficient confidence that epistemic uncertainty for a single mean base-case profile this particular parameter. characterizes the site. This This finding has been conclusion is based on the resolved with no significant quantity and quality of the impact to the SPRA results combined data and an evaluation or conclusions. showing the site is relatively uniform with respect to Vs. For some depth ranges, data from the nearby Savannah River Site (SRS) are used to support the profile interpretation. The documentation presents summaries of velocity data, but does not provide sufficient information to suooort the lack of E1-47

Enclosure 1 Attachment 3 - Resolution of the VEGP Seismic PRA Peer Review Findings F&O Review cc F&O F&O F&O Disposition for GSl-191 Number Element Description Basis Resolution and Other Applications I epistemic uncertainty at the Units 1&2 site over the complete depth range of the Vs profile. This would typically require multiple measurements throughout the depth range that provide a consistent picture of natural variability about a single mean base-case profile. The technical basis and justification that a single base-case profile is appropriate should be provided in more detail. This should include the basis for applying conclusions from other Vogtle locations to the Units 1&2 site. [A related Suggestion 11-2 addresses specifically potential epistemic uncertainty in the Blue Bluff Marl stratum.] 11-8 SHA-E2 11/111 Upper crustal site The Vogtle Site Response Provide a basis in the A discussion of the range attenuation of Analysis notes that the damping. documentation for of possible values of deep ground motion associated with the base-case representing base-case soil damping has been (kappa) is, profile corresponds to a total kappa at the site by a included in the generally, an kappa value for the soil column of single value. The basis documentation. uncertain parameter. 0.01 sec. The report does not might include sensitivity A sensitivity study on the Thus, to maintain address epistemic uncertainty in analyses to show the epistemic uncertainty of hazard-consistent kappa. impact of epistemic deep soil damping has ground motion at the uncertainty in kappa. been performed using control point, this In discussion with staff during the median, lower range, and uncertainty should peer review, it was noted that upper range alternatives be incorporated in randomization of the damping for deep rock damping. the site response associated with the profile layers Site response analysis analysis, or the represents both random variability was performed using 1E-4 basis for not and epistemic uncertainty. It was HF and LF rock input includinq it should also noted that kappa was motion. The resultinq E1-48

Enclosure 1 Attachment 3 - Resolution of the VEGP Seismic PRA Peer Review Findings F&O Review cc F&O F&O F&O Disposition for GSl-191 Number Element Description Basis Resolution and Other Applications be provided. In expected to be small for the amplification functions and either case, the Vogtle site and uncertainties in log-standard deviation technical basis and that small value would not be were weight-averaged and justification should expected to have a significant compared to the original be documented. impact on site amplification. Staff base case for each of BBM also noted that the approach High Pl and BBM Low Pl used had been reviewed by the soil columns. It was NRG for the Vogtle ESP and concluded that the (This F&O originated COLA. inclusion of alternative from SR SHA-E2) base cases for deep soil The SPID provides guidance damping to account accepted by the NRG for explicitly for the epistemic response to NTTF 2.1 uncertainty associated with Recommendation: Seismic that site kappa does not have indicates kappa is difficult to any significant effects on measure and thus subject to large the resulting seismic uncertainty (SPID Section B- hazard curves and UHRS. 5.1.3.2). The sensitivity study has been added to the SPRA Documentation of the technical documentation. basis for kappa characterization This finding has been should be expanded. resolved with no significant impact to the SPRA results or conclusions. 12-1 SHA-J1 Not As part of the PSHA The approach that was taken to Documentation should be A PSHA report has been Met implementation, the model earthquakes in the PSHA provided that describes prepared that describes analyst has different calculation was not identified. how seismic sources are how earthquake events alternatives for There are two basic alternatives modeled in the PSHA were modeled for area modeling the that can be used to model (i.e., how the SSC and sources in the PSHA earthquake earthquake events; as extended GMMs) were calculations. This was by occurrences in the fault ruptures, or as point sources. implemented in the modeling each earthquake calculations. The The approach that is used Vogtle PSHA. as a point source, and PSHA influences how the CEUS ground using correction factors for documentation does motion model is implemented. distance and ground not describe the motion uncertainty that approach that was No documentation is provided on modify the ground motion used to model either of these subjects estimate to include the E1-49

Enclosure 1 Attachment 3 - Resolution of the VEGP Seismic PRA Peer Review Findings F&O Review cc F&O F&O F&O Disposition for GSl-191 Number Element Description Basis Resolution and Other Applications earthquakes. (earthquake source modeling and effect of a closer distance use of the ground motion to a fault rupture (because attenuation models). From the rupture may be closer questions posed to the PSHA to the site than the single (This F&O originated analysts, it is our understanding point used to represent from SR SHA-J1) that earthquakes were modeled that event) and the as point sources and the uncertainty in ground appropriate ground motion motion because the aleatory uncertainty was used in azimuth of the rupture is the calculation. unknown. These correction factors were published by EPRI. This finding has been resolved with no significant impact to the SPRA results or conclusions. 12-11 SHA-J1 Not As part of the PSHA The PSHA analysts were asked Provide a description of A PSHA report has been Met implementation, the to describe the approach that was the earthquake modeling prepared that describes analyst has usedtomodelearthquakesinthe approach that was used how pseudo-faults were alternatives for Charleston RLME seismic source. to model the Charleston implemented to represent modeling the The response indicated that RLME seismic source the Charleston RLME earthquake earthquakes in the Charleston and how the approach source. This includes: 1. A occurrences in the RLME source were modeled was implemented. description of the pseudo-calculations. The using 'pseudo faults'. faults. 2. A definition of PSHA pseudo-faults as documentation does The PSHA report does not: constructed faults that not describe the 1. Describe that a 'pseudo fault' represent possible sources approach that was approach was used to model of future large used to model earthquakes in the Charleston earthquakes. 3. earthquakes in RLME source. Implementation of the RLME sources. 2. Provide a definition of 'pseudo pseudo-faults including faults'. spacing and limits at the

3. Describe how the 'pseudo fault' borders of the Charleston approach was implemented for source. 4. Documentation (This F&O originated the Charleston RLME seismic of the rupture area, length, from SR SHA-J1) source (e.g., what was the fault and width that were spacinQ that was used; how was estimated for possible E1-50

Enclosure 1 Attachment 3 - Resolution of the VEGP Seismic PRA Peer Review Findings F&O Review cc F&O F&O F&O Disposition for GSl-191 Number Element Description Basis Resolution and Other Applications the earthquake rate distributed to future earthquakes. 5. A the faults, etc.). description of how

4. Document the fault rupture earthquake ruptures are model that was used. distributed on the faults.

5. Describe how earthquake This finding has been events are distributed on the resolved with no significant faults. impact to the SPRA results or conclusions.

12-15 SHA-11, Not A screening A screening analysis was not A screening analysis for This evaluation was done SHA-12 Met assessment was performed for hazards such as other seismic hazards for the Vogtle 3&4 COLA performed for soil settlement, fault displacement, should be performed and and is noted in the ESP liquefaction and is tsunami, seiche, etc. documented as part of SAR. The Vogtle 3&4 described in seismic the PSHA and SPRA. evaluation is applicable to, fragility calculation. It is anticipated these other and has been cited in, the seismic hazards will be screened It is expected that Vogtle 1&2 SPRA Fragility A screening out. information in the FSAR report. assessment was not for Vogtle 1 & 2 and in performed for other the COLA for Units 3 & 4 This finding has been potential seismic can be used to support resolved with no significant hazards. this requirement. impact to the SPRA results or conclusions. (This F&O originated from SR SHA-11) 12-16 SHA-J1 Not The Vogtle PSHA The documentation of the PSHA- Prepare a complete and A PSHA report has been Met has gone through a is provided in a collection of up-to-date PSHA prepared that includes number of changes documents that were prepared in document that includes hazard results, and revisions since the 2012-2014 time frame. There all results, sensitivity uncertainties in hazard, 2012 due to does not exist a single document calculations, and sensitivities to input changes in models, that contains a set of results that deaggregation results, uncertainties; this input data, etc. As is based on the current PSHA etc. that is based on the summarizes hazard results new calculations model. current model. for the Vogtle site. were performed and This finding has been reports generated, resolved with no significant sensitivity results, impact to the SPRA results were not carried or conclusions. forward. As a result, there does not exist E1-51

Enclosure 1 Attachment 3 - Resolution of the VEGP Seismic PRA Peer Review Findings F&O Review cc F&O F&O F&O Disposition for GSl-191 Number Element Description Basis Resolution and Other Applications a current report that includes all PSHA results, deaggregations, etc. that is based on the current PSHA model. (This F&O originated from SR SHA-J1) 12-18 SHA-82, I/II The Vogtle PSHA is As part of a site-specific PSHA, A data gathering effort A detailed study of new SHA-C4, Not based on the CEUS there is a need to gather, review should be undertaken to geological, seismological, SHA-H1 Met SSC seismic source and evaluate new geological, identify new information and geophysical Not model which was seismological, or geophysical that post-dates the CEUS information was Met completed in 2012. information or information that is SSC data collection conducted, to determine if The SSC model was defined at a scale that was not effort. The data gathering any information a developed at a considered in the development of effort should also look for subsequent to the EPRI regional scale that the CEUS SSC model. As part of information local to the SSC model is available was based on data the Vogtle SPRA, no effort was Vogtle site region that that should be gathered up until made to gather up-to-date and was not considered, or at incorporated into the about 2010. (Note, local (local to the Vogtle site) a scale that was not seismic hazard results for the date when data information to evaluate whether addressed as part of the Vogtle. This study is was gathered any new information has become CEUS SSC regional described in the SPRA varied; for example available on active faulting and/or evaluation. documentation. While the the earthquake the development new seismic area around the site catalog was sources or the revision of sources Some of this information continues to be studied by complete through in the CEUS SSC model in the may be available in the many earth scientists, 2008.) In the sense vicinity of the Vogtle plant. COLA for Vogtle Units 3 there was no new that the CEUS SSC &4. information identified that model was not Since up-to-data was not would change the estimate specifically gathered, consideration of of seismic hazard for performed as a site- alternatives could not be Vogt le. specific PSHA for addressed. This finding has been the Vogtle site. resolved with no significant impact to the SPRA results (This F&O originated or conclusions. from SR SHA-82) E1-52

Enclosure 1 Attachment 3 - Resolution of the VEGP Seismic PRA Peer Review Findings F&O Review cc F&O F&O F&O Disposition for GSl-191 Number Element Description Basis Resolution and Other Applications 12-2 SHA-J1 Not The method that is For soil sites, the soil hazard is The documentation The methodology used for Met used in the Vogtle generally (though not exclusively, should include a the surface hazard PSHA to estimate since other methods could be description of the calculation has been the soil site hazard used) determined in two steps; methodology that is used described in detail, and a is not described or probabilistic rock hazard results to combine the rock comparison made between referenced. are estimated which are then hazard results and the the GMRS using the two combined with probabilistic site amplification factors approaches 2A and 3. estimates of the site response. to determine the soil Approach 2A was used for The method used in the Vogtle hazard at the Vogtle site. the calculation of SSI input (This F&O originated PSHA to estimate the soil hazard motions at foundation from SR SHA-J1) is not described. elevations and Approach 3 was used for the calculation of surface hazard and GMRS at the ground surface, as defined in NUREG/CR-6728. It was concluded that the use of Approach 2A USHRS as input to the SSI analysis of the Vogtle plant is considered acceptable and does not present any significant inconsistency with the seismic hazard curve and GMRS at the ground surface, which were calculated using Approach 3. This finding has been resolved with no significant impact to the SPRA results or conclusions. E1-53

Enclosure 1 Attachment 3 - Resolution of the VEGP Seismic PRA Peer Review Findings F&O Review cc F&O F&O F&O Disposition for GSl-191 Number Element Description Basis Resolution and Other Applications 12-22 SHA-E2 11/111 Versions 1 and 2 of The site response calculation A framework and A description of the the site response does not present a clear approach for evaluating methodology used to Calculation do not description of how aleatory and and modeling account for epistemic and describe a epistemic uncertainties are uncertainties in the site aleatory uncertainties in framework for identified and evaluated. As a response should be soil hazard has been evaluating and result it is difficult to track the developed and added to the characterizing propagation of uncertainties is implemented. The site documentation. sources of aleatory carried out in the site response response calculation This finding has been and epistemic analysis. documentation should resolved with no significant uncertainty and how fully describe the impact to the SPRA results the approach was It is worth noting that there is methodology and its or conclusions. implemented. some epistemic site response implementation. uncertainty that is accounted for (This F&O originated in the rock GMPEs. from SR SHA-E2) 12-23 SPR-E5 II The quantification The documentation presents the Develop and document Additional detail has been process has results of three different an understanding of the added to the SPRA included the uncertainty calculations for CDF earlier point estimate Quantification report to uncertainties in the and LERF. In addition, point results for CDF and LERF document the uncertainty, seismic hazard, estimates for CDF and LERF are and of uncertainty results. importance, and sensitivity fragility and calculated and reported in other analyses and relate the systems-analysis sections. Thus the documentation uncertainty analysis mean elements of the reports two estimates of the mean CDF and LERF to the SPRA. The results CDF and LERF respectively from point estimate values. presented are different uncertainty calculations internally and a 'Point Estimates' result for This finding has been inconsistent and are each. All of these results are resolved with no significant inconsistent with the different than the point estimate impact to the SPRA results results reported in (approximate mean) reported in or conclusions. other sections for other sections for CDF and LERF, CDF and LERF, respectively. The documentation . respectively. in the report does not describe the basis (inputs) for these (This F&O originated calculations, or offer an from SR SPR-E5) interpretation of the results. 12-24 SPR-E5 II The Quantification The uncertainty analysis is Provide documentation of Additional detail has been report does not presented with the results the uncertaintv analvsis added to the E1-54

Enclosure 1 Attachment 3 - Resolution of the VEGP Seismic PRA Peer Review Findings F&O Review cc F&O F&O F&O Disposition for GSl-191 Number Element Description Basis Resolution and Other Applications provide reported. The report provides that describes the results, documentation of the documentation of limited discussion of the results how they are being seismic plant response the uncertainty and the insights that might be interpreted and the model, model analysis results. gained from them. insights that are derived implementation, and from them. quantification in the QU The two sets of results that are report. In addition, the reported are not discussed in uncertainty, importance, (This F&O originated terms of their relationship to each and sensitivity analyses from SR SPR-ES) other. For instance the mean are described in more values should be the same (but detail. are not). The uncertainty estimates provide insight to the This finding has been total uncertainty and the resolved with no significant contribution of the basic event impact to the SPRA results uncertainty to the total. or conclusions. In addition, neither the table of results or the discussion identifies what is the 'final' uncertainty result that includes the propagation of uncertainties of all elements of the SPRA to the estimates of CDF and LERF. 12-26 SPR-ES II There are The report does not present the Document the results of Updated Monte Carlo differences in the results of sensitivity calculations sensitivity calculations on uncertainty runs have results for CDF and with regard to the number of the number of Monte been performed with LERF that are Monte Carlo simulations that are Carlo simulations 20,000 iterations for SCDF reported. A possible needed to produce stable results. required to produce and SLERF. This is a contributor to these stable results. sufficiently high number of differences may be It is our understanding from simulations to produce a due to the number of discussion with the PRA staff that stable result. The SPRA Monte Carlo these types of sensitivity documentation has been simulations that calculations were performed. updated to clearly were performed. indicates the results. (This F&O originated This finding has been from SR SPR-ES) resolved with no siqnificant E1-55

Enclosure 1 Attachment 3 - Resolution of the VEGP Seismic PRA Peer Review Findings F&O Review cc F&O F&O F&O Disposition for GSl-191 Number Element Description Basis Resolution and Other Applications impact to the SPRA results or conclusions. 12-27 SPR-F2 Met Documentation The current quantification Provide clear and The QU report should be provided document does not provide a complete documentation documentation has been that describes how clear description of the how the of the approach used to updated to describe the the plant model plant model is quantified. For quantify the seismic plant quantification process, analysis is example the discussion does not response model, to including the technique for quantified. identify how calculations are perform the risk combining cutsets over the performed, what the limitations of quantification, uncertainty 14 acceleration intervals, these quantifications are and how analysis, and importance and obtaining the they affect the results. analysis. importance measures. (This F&O originated from SR SPR-F2) This finding has been resolved with no significant impact to the SPRA results or conclusions. 12-29 SPR-E2 Met The Quantification There is limited documentation of Document the process Additional detail has been report provides the process and the numerical and methods that were added to the QU report to limited methods that were used to used to perform the document the uncertainty, documentation of perform the uncertainty analysis. uncertainty analysis. importance, and sensitivity the process and Based on the documentation that Where appropriate analyses and relate the methods that were is provided and discussions with document where uncertainty analysis mean used to perform the the PRA staff there is limited but consistencies and SCDF and SLERF to the uncertainty analysis. not complete understanding of potential inconsistencies point estimate values. the methods that were used and in results might be the relationship of these methods expected. This finding has been to the results were obtained. resolved with no significant (This F&O originated impact to the SPRA results from SR SPR-E2) In some cases (as described in or conclusions. the documentation) the results from the uncertainty analysis are not the same as the results reported in other sections of the documentation for CDF and LERF (though this connection is not E1-56

Enclosure 1 Attachment 3 - Resolution of the VEGP Seismic PRA Peer Review Findings F&O Review cc F&O F&O F&O Disposition for GSl-191 Number Element Description Basis Resolution and Other Applications clearly stated in the report). However, it would seem the results should be internally consistent. 12-31 SPR-F1 Not The standard There is limited documentation Documentation should be Additional detail has been Met requires a level of that describes the seismic plant provided in sufficient added to the documentation that response analysis and detail that describes the documentation of the provides an quantification; how the model was seismic plant model, how seismic plant response understanding of the implemented, how the it is implement and model, model seismic plant quantification was performed and quantified. implementation, and response model and a discussion of the analysis quantification in the QU the quantification. results. report. In addition, the This requirement is uncertainty, importance, not met. To meet this requirement, the and sensitivity analyses documentation must be in are described in more considerable detail in order to detail. support the review process and (This F&O originated future updates. Part of the This finding has been from SR SPR-F1) documentation should include a resolved with no significant detailed discussion of the results, impact to the SPRA results sensitivity calculations, and the or conclusions. uncertainty analysis. 12-32 SPR-F3 Met The documentation The purpose of this supporting Document and discuss The documentation of the of the sources of requirement is that documentation the contribution of the uncertainty analysis has model uncertainty should be presented that different sources of been expanded in the and a description of addresses the sources of uncertainty that are Quantification report. A the analysis epistemic (knowledge) uncertainty modeled in the SPRA. discussion of sources of assumptions is not that are modeled and their model uncertainty has complete in the contribution to the total been added to the report, SPRA quantification uncertainty in CDF and LERF. and potentially important report. In addition, sources have been there is not a clear In addition, the documentation addressed in the sensitivity description of the should discuss elements of the analysis. uncertainty analysis seismic plant model where there and the contributors may be latent sources of This finding has been to the total uncertainty that are not modeled resolved with no significant uncertainty bevond and assumptions that are made in impact to the SPRA results E1-57

Enclosure 1 Attachment 3 - Resolution of the VEGP Seismic PRA Peer Review Findings F&O Review cc F&O F&O F&O Disposition for GSl-191 Number Element Description Basis Resolution and Other Applications a simple report from performing the analysis. or conclusions. UN CERT. (This F&O originated from SR SPR-F3) 12-36 SHA-83, 1111, As part of a site- As part of the Vogtle PSHA an An up-to-date earthquake An update to the SHA-C4, Not specific PSHA, an effort was not made to gather catalog for the Vogtle site earthquake catalog was SHA-H1 Met up-to-date data on earthquakes that region should be prepared from the time of Not earthquake catalog occurred since 2008. As such, the developed to assess the CEUS SSC catalog Met should be used. The analysts did not assess whether whether modifications to (through 2008) through CEUS SSC study more recent seismicity is the seismic source February 2016. The rate involved the consistent with the recurrence parameters or of occurrence of development of a characterization parameters required. The updated earthquakes within 320 km comprehensive estimated as part of the CEUS catalog, resources used of the Vogtle site was earthquake catalog SSC study (NRC, 2012). in compiling the update compared to the rate of based on data and the results of the earthquakes represented through 2008. The We note that as part of the Vogtle evaluation should be by the CEUS SSC seismic Vogtle site-specific PSHA, calculations were documented as part of source model for that PSHA should performed to recompute the the PSHA. If more recent same area, this consider the impact seismic hazard at the site to take seismicity is not comparison being made SSC of any into account changes in the consistent with the for M>2.9. It was found additional seismicity CEUS SSC earthquake catalog existing CEUS SSC that the updated catalog since 2008 up to the through 2008 that were made seismic source implied a rate of time the study following the completion of the parameters, the earthquakes that is lower started. CEUS SSC study. These parameters should be than the mean rate from changes reflect the identification updated and the PSHA the CEUS SSC seismic (This F&O originated of reservoir induced seismicity should be updated. sources. Therefore, from SR SHA-C4) earthquakes and the re- incorporating the effects of interpretation of the location of a updated catalog on the some earthquakes in the hazard at Vogtle would Charleston, SC area that decrease the hazard occurred in the 1880's (EPRI, slightly, and was not 2014). undertaken. This comparison is documented References in the SPRA documentation. EPRI (2014). Review of EPRI This findinQ has been E1-58

Enclosure 1 Attachment 3 - Resolution of the VEGP Seismic PRA Peer Review Findings F&O Review cc F&O F&O F&O Disposition for GSl-191 Number Element Description Basis Resolution and Other Applications 1021097 Earthquake Catalog for resolved with no significant RIS Earthquakes in the impact to the SPRA results Southeastern U. S. and or conclusions. Earthquakes in South Carolina Near the Time of the 1886 Charleston Earthquake Sequence, transmitted by letter from J. Richards to R. McGuire on March 5, 2014. 12-8 SHA-J3 Not A foundational The documentation of the The resolution to this Sources of uncertainty in Met element of PSHA as sources of model uncertainty finding could involve: the seismic hazard it has evolved over analysis and a description of the analysis for Vogtle are the past 30 years is analysis assumptions is not 1. Documentation and discussed in the updated the development complete in the PSHA report in its discussion of the SPRA documentation. and implementation current form such that a clear contribution of different These include uncertainty of methods to understanding of the contribution sources of uncertainty in seismic source model identify, evaluate, of individual sources of that are modeled in the (for background and model sources uncertainty to the estimate of PSHA. The earthquake sources and of epistemic (model hazard are understood. Limited documentation of the for the Charleston RLME), and parametric). information on the contribution of contribution of different in maximum magnitude for uncertainty in the seismic sources to the total mean sources of uncertainty background seismic estimate of ground hazard is presented, but can be shown by means sources and for the motion hazards. As information on the contributors to of 'tornado plots' that Charleston RLME, in such fairly rigorous the uncertainty is not provided. quantify the sensitivity of ground motion prediction analyses are carried the hazard at different equation, in smoothing out (SSHAC studies) With respect to addressing model ground motion levels to assumptions for seismicity to quantitatively uncertainties and associated the various branches in parameters in background address model assumptions there are some the logic tree. These sources, and in site uncertainties. examples that can be identified in plots show which sources amplification model. the Vogtle PSHA. For example, in of epistemic uncertainty "Tornado plots" are At the same time the site response analysis the are most important. It included in the updated there is within any assumption is made that the 1D should include the source SPRA documentation that analysis sources of equivalent linear model (SHAKE model uncertainty, show the contribution to uncertainty that are type) to estimate the site ground motion model total uncertainty in seismic not directly modeled amplification and ground motion uncertainty, and site hazard from source model and assumptions input to plant structures is response uncertainty. uncertainty, maximum that are made for appropriate. Currently, the total magnitude uncertainty, E1-59

Enclosure 1 Attachment 3 - Resolution of the VEGP Seismic PRA Peer Review Findings F&O Review cc F&O F&O F&O Disposition for GSl-191 Number Element Description Basis Resolution and Other Applications pragmatic or other uncertainty is shown by ground motion prediction reasons. There are the hazard fractiles, but it equation uncertainty, also sources of is not broken down to smoothing assumptions for model uncertainty provide understanding as seismicity parameters in that are embedded to what is most important. background sources, and in the context of site response uncertainty. current practice that 2. Identification and These plots are presented are 'accepted' and discussion of model for 10 Hz and 1 Hz typically not subject assumptions that are spectral acceleration, for to critical review. For made. ground motion amplitudes instance, in the corresponding to mean PSHA it is standard annual frequencies of practice to assume exceedance of 1E-4 and that the temporal 1E-5. These "tornado occurrence of plots" show that ground earthquakes is motion prediction equation defined by a is the major contributor to Poisson process. seismic hazard uncertainty This assumption is for both 10 Hz and 1 hz well accepted spectral acceleration, and despite the fact that maximum magnitude of it violates certain the Charleston RLME fundamentally source is an important understanding of contributor for 1 Hz tectonic processes spectral acceleration. (strain The use of equivalent accumulation). A linear one-dimensional second practice is site response analysis, the fact that and its associated earthquake assumptions, and its aftershocks are not adequacy for the Vogtle modeled in the site are documented in the PSHA, even though hazard calculation. they may be This finding has been significant events resolved with no significant (depending on the impact to the SPRA results size of the main or conclusions. E1-60

Enclosure 1 Attachment 3 - Resolution of the VEGP Seismic PRA Peer Review Findings F&O Review cc F&O F&O F&O Disposition for GSl-191 Number Element Description Basis Resolution and other Applications event). In the spirit of the standard it seems appropriate that sources of model uncertainty that are modeled as well as sources of uncertainty and associated assumptions as they relate to the site-specific analysis should be identified/ discussed and their influence on the results discussed. As SPRA reviews and the use of the standard has evolved, it would seem the former interpretation is reasonable, but potentially incomplete. It is reasonable from the perspective that documentation of the sources of model uncertainty and their contribution to the site-specific hazard results is a valuable E1-61

Enclosure 1 Attachment 3 - Resolution of the VEGP Seismic PRA Peer Review Findings F&O Review cc F&O F&O F&O Disposition for GSl-191 Number Element Description Basis Resolution and Other Applications product that supports the peer review process and assessments in the future as new information becomes available). Similarly, documenting assumptions provides similar support for peer reviews and future updates. The notion that model uncertainties and related assumptions that are not addressed in the PSHA is at a certain level an extreme requirement that may not be readily met and may not be particularly supportive of the analysis that is performed. For purposes of this review, the following approach is taken with regard to this supporting requirement:

1. The E1-62

Enclosure 1 Attachment 3 - Resolution of the VEGP Seismic PRA Peer Review Findings F&O Review cc F&O F&O F&O Disposition for GSl-191 Number Element Description Basis Resolution and Other Applications documentation should present quantitative results and discussion the sources of epistemic uncertainty that are modeled and their contribution to the total uncertainty in the seismic hazard.

2. The documentation should discuss elements of the PSHA model where their may be latent sources of model uncertainty that are not modeled and assumptions that are made in performing the analysis.

(This F&O originated from SR SHA-J3) 14-1 SFR-A2 I The conservatisms SFR-A2 requires that seismic Account for conservatism Evaluation of anchorage that exist in fragilities be based on plant- in the building response has been updated to structural demand specific data and that they are analyses in the structure include clipping of in-were not properly realistic and median centered with response factor for structure response accounted for in the reasonable estimates of component fragility spectra, and the estimation of uncertainty. evaluations. methodology is component and documented in the fragility structure fragilities. The structural response factor Use clipped spectra for notebook. used in all component fragilities assessing anchorage Structure response is reviewed is reported as 1.0. This capacities. dominated by the soft soil factor will be greater than 1.0 on which Vogtle 1 and 2 (This F&O originated because of the conservatism structures are founded. E1-63

Enclosure 1 Attachment 3 - Resolution of the VEGP Seismic PRA Peer Review Findings F&O Review cc F&O F&O F&O Disposition for GSl-191 Number Element Description Basis Resolution and Other Applications from SR SFR-A2) . introduced in the demand through This would cause higher the structural analysis. Because damping at lower hazard of this, the component and frequency levels and lead structural fragilities are biased to stress similar to the low. stress calculated for the buildings at 1E-4. As a The fragilities developed for result the structural structures and components that response factor is close to are mounted in those structures 1 and is accounted for will be biased low because the appropriately in the fragility input structural demands include evaluations. conservatisms. Time histories The input motion at the used for the SSI analysis have control point in the SSI been processed such that each analysis has been record envelopes the target modified to reasonably UHRS. This will introduce some match the corresponding level of conservatism. The input 1E-4 UHRS from the site-motion at the control point has consistent input motion been scaled to produce resultant analysis. FIRS that envelopes the FIRS This finding has been coming out of the site-consistent resolved. input motion analysis. In structure response spectra coming out of the SSI analyses were not peak clipped when computing anchorage demands. Structure response at the calculated equipment fragility levels is considerably higher than the 1E-4 UHRS considered in the building response analyses. The structure will have additional cracked shear walls and higher associated levels of damping at these higher ground motions. 14-10 SFR-A2 I Significant In the fragility calculations of heat Realistic nozzle loads The CCW and ACCW heat conservatisms were exchangers (PRA-BC-V-14-009 should be determined for exchanger capacities have E1-64

Enclosure 1 Attachment 3 - Resolution of the VEGP Seismic PRA Peer Review Findings F&O Review cc F&O F&O F&O Disposition for GSl-191 Number Element Description Basis Resolution and Other Applications noted in several Appendix A}, nozzle loads fragility evaluation of heat been updated to reflect sampled fragility significantly contribute to the exchangers. realistic nozzle loads. The calculations. seismic demands which form the equipment fragilities have basis for the median capacities. been updated to account Based on in-plant walkdowns by for appropriate frequency, the peer review teams and also The equipment capacity and uncertainty has been (This F&O originated noted in the walkdown report, the factor should be based on considered in these from SR SFR-A2) piping is well supported in all the frequency range of updates. directions and will not impose interest. That frequency This finding has been significant nozzle loads during a range of interest is resolved. seismic event. The CCW and centered at the ACCW capacities are below the fundamental frequency of 2.Sg screening level and are the pump, and considers significant contributors to risk so some uncertainty in that more realistic fragilities are frequency. required. Battery rack 11806B3BN3 in calculation PRA-BC-V-14-010 Appendix J2 is governed by GERS capacity. The GERS capacity is taken to be 1g, which corresponds to a frequency of 1 Hz. This is not realistic. The actual capacity is about 4g. The median capacity reported in the calculation is well below the 2.Sg screening level and is not realistic. The median capacity reported for the Turbine Driven Auxiliary Feedwater Pump is reported in Calculation PRA-BC-V-14-008 as 1.56g. This fragility is based on the seismic qualification document. The frequency range E1-65

Enclosure 1 Attachment 3 - Resolution of the VEGP Seismic PRA Peer Review Findings F&O Review cc F&O F&O F&O Disposition for GSl-191 Number Element Description Basis Resolution and Other Applications of interest for the fragility evaluation should be centered around the fundamental frequency of the assembly and not consider the entire frequency ranQe. 14-14 SFR-G2 Met The iterative In review of the seismic fragility Add a description of the The description of the process used for calculation for the safety features iterative process for iterative process for developing realistic sequencer (11821U3001), it was computing the component computing fragilities has fragilities is not well discovered that an iterative fragilities in the SPRA been documented. documented. process was used. The initial documentation fragility is based on EPRI 6041 This finding has been screening methodology and an resolved with no significant equipment capacity factor that is impact to the SPRA results (This F&O originated equal to the EPRI 6041 median or conclusions. from SR SFR-G2) capacity divided by the peak in structure demand. If this value is less than the screening capacity (2.5g), then the fragility may be refined by examining the component fundamental frequency. The fragility may be further refined by examining component specific qualification test reports. However, the fragility used in the logic tree by the systems analyst is generally the highest of these computed. This is reasonable and appropriate, however, this process is not described in the fragility notebook or fragility calculations. 14-17 SFR-02 Met Inconsistencies and Fragilities for the Vogtle 1&2 Update SNC calculation The following changes errors in NSSS Nuclear Steam Supply System no. PRA-BC-V-14-015 to have been made: NSSS fragility (NSSS) are based on the results incorporate corrections fragility calculations have development. of the WestinQhouse analysis of and enhancements. been updated to reflect E1-66

Enclosure 1 Attachment 3 - Resolution of the VEGP Seismic PRA Peer Review Findings F&O Review cc F&O F&O F&O Disposition for GSl-191 Number Element Description Basis Resolution and Other Applications record (AOR) associated with the Westinghouse-provided safe shutdown earthquake (SSE). critical loads and support In general, fragilities are capacities represented in (This F&O originated developed through scaling of the the critical failure modes; from SR SFR-02) SSE demands to the RLE and the effect of inelastic using the AOR seismic margins. energy absorption is Various deficiencies were noted factored in and in the development of the documented in fragility fragilities associated with these calculation as appropriate; components. the Reactor Coolant Pump Basis: The NSSS Seismic fragility has been updated fragility evaluation (SNC to reflect the failure of the calculation no. PRA-BC-V pump associated with 015) includes detail calculations LOCA; the reactor for each of the major NSSS internals fragility has been components. It indicates that the updated in the calculation; critical failure modes for the and the new fragilities components are controlled by the have been reflected in the support capacities. updated SPRA model. During the Peer Review, the team members discussed these issues This finding has been with SNC staff to obtain insights resolved. and develop potential resolution paths. Key issues included: (a) Basis for assumption that the support capacities represented the critical failure mode was not documented. SNC indicated that this was based on input from Westinghouse and NUREG-3360 and will update the fragility evaluation of provide this information. (b) Inelastic energy absorption was not credited to increase the median capacities - this does not result in realistic median E1-67

Enclosure 1 Attachment 3 - Resolution of the VEGP Seismic PRA Peer Review Findings F&O Review cc F&O F&O F&O Disposition for GSl-191 Number Element Description Basis Resolution and Other Applications capacities (overly conservative). (c) Reactor Coolant Pump fragility was based on consideration of the failure of the attached CCW piping, due to an assumption that a small-break/RCP seal LOCA was critical. It was learned during the Peer Review that failure in the system model was linked to a large-break LOCA, so the failure mode considered in the fragility evaluation is not consistent with the system model - SNC indicated that they will revise the fragility evaluation. (d) Reactor Internal fragility evaluation determined the demand based an average spectral acceleration over the range of 2 to 3 Hz, rather than using the peak acceleration in this range of the ISRS, and did not consider the contribution of higher modes. SNC indicated that this was done to avoid an overly conservative capacity, but agreed that the contribution of higher modes should be addressed, and will revise the calculation. (f) Control Rod Drive Mechanism fragility evaluation assumed that material stresses were the critical failure mode, and did not address the potential impact of deflections on rod drop. SNC indicated that information provided by WestinQhouse (based on a E1-68

Enclosure 1 Attachment 3 - Resolution of the VEGP Seismic PRA Peer Review Findings F&O Review cc F&O F&O F&O Disposition for GSl-191 Number Element Description Basis Resolution and Other Applications Japanese testing program) indicated that the deflection levels associated with seismic loading does not impact rod drop, and agree to add this discussion to the calculation. 14-20 SFR-E4, Met Seismic induced fire The only mention for seismic Seismic induced fire is an The seismic-induced fire SPR-89 Met evaluations are not induced fire evaluation is important element of the and flood evaluations have documented in the contained in the quantification fragility evaluation been updated, and walkdown report or notebook. Based on discussions process and this should documented in the fragility fragility calculations. during the peer review, it is be clearly documented. and quantification report. understood that seismic induced This includes the details of fire was a key consideration the walkdown procedure during the walkdowns. However, used to evaluate the (This F&O originated detail of the walkdown procedure potential for seismically from SR SFR-E4) for fire following earthquake is induced fires, including the missing. The write up should methodology, screening include team composition, criteria and results. methodology, screening criteria, and results, This finding has been resolved with no significant impact to the SPRA results or conclusions. 14-4 SFR-01 Met A potential for SFR-01 requires that realistic Evaluate the potential for The evaluation for sloshing induced failure modes of structures and flood induced failure of potential flood induced inundation of the equipment that interfere with the the NSCW Pumps or failure of the NSCW NSCW Pumps operation of that equipment be NSCW discharge MOVs. pumps or the NSCW (11202P4007, identified. discharge MOVs has been 11202P408) and performed and associated The potential for earthquake documented in the fragility discharge motor induced sloshing of the water calculation for the NSCW operated valves within the NSCW tower exists. tower. There was no (1HV11600, 11606, From field walkdowns of the significant impact on the 11607, 11613) inthe NSCW it was observed that there pump or MOV fragilities. NSCW exists and is a potential for sloshing of was not identified contents to potentially splash onto This finding has been either in the or flood the pumps and or motor resolved with no significant E1-69

Enclosure 1 Attachment 3 - Resolution of the VEGP Seismic PRA Peer Review Findings F&O Review cc F&O F&O F&O Disposition for GSl-191 Number Element Description Basis Resolution and Other Applications walkdowns or operated valves on the attached impact to the SPRA results subsequent discharge piping. or conclusions. analysis. (This F&O originated from SR SFR-D1) 14-5 SFR-D1 Met The potential for Vogtle 1&2 is a soil site, with Develop estimates of the Documentation has been seismically-induced engineered fill from the rock differential settlements updated to include the differential interface to the finished grade. between adjacent effects of earthquake settlements between The in-scope Seismic Category I structures and assess the induced settlement; no structures was not structures have foundations with fragility of commodities significant differential addressed. varying embedment depths, based on their ability to settlements were ranging from surface founded accommodate the computed between the (elev. 220 ft.) to a foundation associated differential structures. embedment of 110 ft. (elev. 110 displacements. This finding has been (This F&O originated ft.). Since soils, including resolved with no significant from SR SFR-D1) engineered fill, will impact to the SPRA results consolidate/settle to some extent or conclusions. when subjected to high level earthquake ground motion, and the amount of settlement is proportional to the thickness of the soil layer under the foundation, the settlement of one structure relative to another structure is dependent on the depth of the foundation embedment. The Fragility Notebook (PRA-BC-V-14-025) does not address the potential differential settlement between buildings, or the potential effect on commodities (e.g., piping, electrical raceways, HVAC ducts, etc.) that cross the separation between adjacent E1-70

Enclosure 1 Attachment 3 - Resolution of the VEGP Seismic PRA Peer Review Findings F&O Review cc F&O F&O F&O Disposition for GSl-191 Number Element Description Basis Resolution and Other Applications structures. During the performance of the Peer Review, SNC personnel indicated that the consideration of differential settlements was not required, since the structures were founded on engineered fill. 14-6 SFR-G2 Met The results of the The walkdown guidance provided Provide documentation of As noted in the Finding seismic gap/shake in Appendix F (Checklists and the results of the seismic basis, inspection of the space walkdowns Walkdown Data Sheets) of EPRI gap walkdowns. seismic gaps was included are not documented. NP-6041 includes attributes of in the seismic walkdowns. seismic gaps between structures Piping across seismic which should be addressed in the gaps is designed with performance of the walkdowns. adequate flexibility to (This F&O originated These include the clearance accommodate building from SR SFR-G2) between adjacent structures and motions, and pipe sleeves the ability of any subsystems provide adequate gaps for (e.g., piping, cable trays, HVAC piping movement. The ducts) spanning the gap to documentation has been accommodate the differential updated to reflect the seismic displacements. inspections performed during the walkdowns. The Seismic Walkdown Report This finding has been (PRA-BC-V-14-005) does not resolved with no significant include documentation of the impact to the SPRA results results/findings/observations or conclusions. associated with the inspection of the seismic gaps between structures or the subsystems spanning the gap. During the performance of the Peer Review, SNC personal indicated that inspection of the seismic gaps was included in the seismic walkdowns, but not explicitly described in the report. The ability of components to E1-71

Enclosure 1 Attachment 3 - Resolution of the VEGP Seismic PRA Peer Review Findings F&O Review cc F&O F&O F&O Disposition for GSl-191 Number Element Description Basis Resolution and Other Applications accommodate potential differential movement at the building separations is implied in the discussion of rugged components (piping, cable trays, and HVAC ducts) in the section on Rationale for Screening in the report. In addition, information from the Vogtle IPEEE Report indicated that the seismic gaps had been inspected during the IPEEE. 14-7 SFR-A2, I, The fragility The determination of the Update the fragility The fragility evaluation of SFR-F4 Met evaluation for the fundamental frequency of evaluation for the polar the polar crane has been Containment Polar structures and components crane to address potential updated to address Crane (in fragility involves a certain degree of uncertainty in the potential uncertainty in the notebook) did not uncertainty. This uncertainty fundamental frequency fundamental frequency address the impact must be accounted for in the and the contribution of and contribution of higher of variation in the determination of the seismic higher modes. modes. fundamental accelerations from the applicable This finding has been frequency on the in-structure response spectra resolved. applicable seismic (ISRS). demand. The section of the Polar Crane of the Fragility Notebook evaluates the polar crane as a potential (This F&O originated seismic interaction source relative from SR SFR-A2) to the reactor vessel and other NSSS components inside the containment structure. In the determination of the vertical spectral acceleration applicable to the polar crane, the computed fundamental frequency falls within a valley in the applicable ISRS, on the low frequency side of the primary spectral peak. E1-72

Enclosure 1 Attachment 3 - Resolution of the VEGP Seismic PRA Peer Review Findings F&O Review cc F&O F&O F&O Disposition for GSl-191 Number Element Description Basis Resolution and Other Applications Uncertainty in the calculated frequency, and the contribution of high modes, could result in an increase in the applied vertical acceleration. During the performance of the Peer Review, SNC personnel provided a written response indicating that it is appropriate to increase the applied acceleration by 50%, which will result in a 20% decrease in the median capacity of the polar crane. 14-8 SFR-F3 111111 Relay fragility The relay evaluation for the Perform more realistic The relay fragilities have calculations include turbine driven auxiliary feedwater relay fragility evaluations. been updated using the conservative pump control panel in calculation appropriate response and assumptions. PRA-BC-V-14-008 is based on a in-cabinet amplification generic capacity for motor starters factors, and are realistic. and contactors (intended for This finding has been motor control centers) and an resolved. (This F&O originated amplification factor associated from SR SFR-F3) with center of door panel response. Based on walkdown observations the relay is not mounted on the door panel so is likely on an internal bracket. The median capacity of 0.627g is well below the screening level and is not realistic. The relay evaluations in calculation PRA-BC-V-14-009 are governed by response in the vertical direction, and the in-cabinet amplification factors used in the calculation are associated with horizontal response. The E1-73

Enclosure 1 Attachment 3 - Resolution of the VEGP Seismic PRA Peer Review Findings F&O Review cc F&O F&O F&O Disposition for GSl-191 Number Element Description Basis Resolution and Other Applications resulting median capacities of 0.762g (Appendix M1) and 1.026g (Appendix M2) are well below the screening level and are not realistic. 14-9 SFR-D2 Met The seismic The summary of the seismic Perform resolution of The noted walkdown walkdown report walkdowns documents a number open items and provide issues have been includes a number of issues identified during the documentation of the evaluated and reflected in of open items that performance of the walkdowns resolution associated with the revised documentation: are not are not that required follow-up actions each of the issues, either - potential piping traceable to a (31). These include spatial in the Fragility Notebook interaction; resolution interaction issues, housekeeping or the SPRA Database. - the difference in inverter issues, anchorage issues, valves anchorage configuration; having configurations that do not - potential interaction meet the EPRI guidelines, concerns with the (This F&O originated configuration issues, installation overhead heater; this from SR SFR-D2) errors, etc. evaluation is in the fragility notebook. The Seismic Walkdown Report Valve operator heights & does not document how the weights that were outside issues identified during the EPRI guidelines have walkdowns have been addressed, been taken into account in either in the field (e.g., correction the fragility analysis for of installation errors, resolution of these components. housekeeping issues) or in the The Diesel Generator fragility evaluations (e.g., valve Exhaust Silencer was re-configurations, anchorage evaluated to the as-issues). During the performance operated condition. of the Peer Review, the Peer The fragility analysis for Review Team provided a list of these components has the walkdown issues to SNC been completed for the as personnel, and SNC provided a built condition. summary of how they were This finding has been addressed. Most issues had resolved. been adequately addressed during the development of the SPRA, but it was determined that E1-74

Enclosure 1 Attachment 3 - Resolution of the VEGP Seismic PRA Peer Review Findings F&O Review cc F&O F&O F&O Disposition for GSl-191 Number Element Description Basis Resolution and Other Applications the following would require further effort for resolution: (a) Potential interaction between piping and deluge valve - follow-up walkdowns required. (b) Anchorage configuration on inverter - follow-up revision to ' fragility evaluation required (c) Overhead heater poses potential interaction issue - follow-up walkdown required. (d) Valve operator heights/weights outside of EPRI guidelines - follow-up walkdown required. (e) Diesel Generator exhaust silencer anchor bolt nuts - not addressed in fragility evaluation, further evaluation required. (f) Valve operator heights outside of EPRI guidelines and potential lack of yoke support - these valves are part of the unfinished scope described in the Fragility Notebook, which will be completed in the future. (g) Valve operator heights outside of EPRI guidelines - further evaluation required. 16-1 SFR-F3, 11/111, The model Relay chatter is consistently being Complete the analysis The approach to screening SPR-84, Not presented for peer observed as a significant and incorporate the and modeling of SPR-E5 Met, review did not contributor to risk profile in effects of relay chatter seismically-induced relay II incorporate the recently peer reviewed S-PRAs and similar devices in the failures and chatter was effects of relay and it is therefore realistic to PRA logic model. provided to the peer chatter as the expect that relay chatter is a review team and analysis was not yet potential significant contributor. determined to have been complete. DurinQ the peer review it was performed aooropriately; E1-75

Enclosure 1 Attachment 3 - Resolution of the VEGP Seismic PRA Peer Review Findings F&O Review cc F&O F&O F&O Disposition for GSl-191 Number Element Description Basis Resolution and Other Applications discussed that the SPRA team only the incorporation into does not believe relays will be a the model of the impacts of significant contributors but it was relay chatter from (This F&O originated also said that this conclusion/ unscreened relays was not from SR SPR-84) expectation is based on complete. The final potentially crediting operator screening resulted in only actions. Thus, the effects of relay 2 relays being chatter per se may be significant incorporated into the (and provide some insights) while model, with one having an the combination of relays and a operator action. Relay number of HEP may not be. chatter fragilities and impacts have been incorporated into the seismic model, in a manner consistent with that used for other failures. This finding has been resolved. 16-10 SPR-86 Met The documentation There is only a short sentence More detailed Walkdown documentation about the supporting the discussion on documentation is on accessibility for walkdowns in alternative access pathways. suggested to support the operator actions, including support to seismic conclusion on photos, has been impact on HRA accessibility, alternative improved. Potential failure appear limited. route, availability of of block walls has been tools/keys, clear reviewed and documented. identification of Required tools and equipment manipulated in equipment, such as (This F&O originated each local action. ladders, have been from SR SPR-86) identified with locations Obviously, the goal of the when needed. The enhanced documentation documentation supports is not to convince the the seismic HRA peer reviewer that the assumptions and walkdowns were modeling. performed but rather to ensure that the analyst is This finding has been E1-76

Enclosure 1 Attachment 3 - Resolution of the VEGP Seismic PRA Peer Review Findings F&O Review cc F&O F&O F&O Disposition for GSl-191 Number Element Description Basis Resolution and Other Applications fully convinced of the resolved with no significant conclusions. impact to the SPRA results or conclusions. Past SPRAs have shown examples of equipment needed for the HFE that was not in the SEL, or that has different actuators when manually actuated, or that needed ladders that were not easily accessible or that were close to block walls (or under ceiling that could collapse) that were not considered an issue because the block walls were not near safety related equipment (and therefore not addressed in the rest of the SPRA work). In this perspective, a more systematic documentation of the feasibility and accessibility analysis for each of the HFE credited in the SPRA is suggested. E1-77

Enclosure 1 Attachment 3 - Resolution of the VEGP Seismic PRA Peer Review Findings F&O Review cc F&O F&O F&O Disposition for GSl-191 Number Element Description Basis Resolution and Other Applications 16-11 SPR-E2 Met Missing review of It is understood that the As this exercise was A detailed quantitative the potential for investigation performed in internal apparently performed for HRA dependency analysis additional events to identify potential HFE the Fire PRA (as based on using the HRA dependencies dependency has been relied upon discussed during the peer calculator was performed introduced by the in the Vogtle SPRA. review), it is suggested and documented. There SPRA models (QU- that a review of the was no significant impact C1&2) The SPRA logic may identify potential for unforeseen on results since human additional dependencies trends dependencies trends is actions are not significant that were not identified in the performed. contributors in the Vogtle internal events. SPRA. (This F&O originated As it is understood that from SR SPR-E2) the plan is to transition to This finding has been a different dependency resolved. analysis method (based on HRA calculator), this may be addressed within the same transition as it is realistic to expect that not too many (if any) new dependencies would be identified. 16-12 SPR-E2 Met Missing It is an industry expectation (as It is understood that the The QU report has been documentation of discussed in NEI peer review task SPRA documentation will updated to document the the review of non force meetings) that review of the be revised to incorporate review of both dominant significant cutsets non significant cutsets is explicitly explicitly the two reviews cutsets and non-significant QU-05. documented. discussed in the basis for cutsets for both CDF and this F&O. It is also LERF. Based on discussion during the recommended to peer review, two reviews were document the review of This finding has been (This F&O originated performed to validate the overall cutsets following resolved with no significant from SR SPR-E2) model and cutsets. The first was guidance from the NEI impact to the SPRA results a random review of cutsets at peer review task force. or conclusions. midpoints and low significance for each of the %Gxx initiators to verify that the cutsets are valid cutsets, and that the patterns are appropriate. That is, if one cutset E1-78

Enclosure 1 Attachment 3 - Resolution of the VEGP Seismic PRA Peer Review Findings F&O Review cc F&O F&O F&O Disposition for GSl-191 Number Element Description Basis Resolution and Other Applications is valid, then another cutsetwith slightly different seismic failures (or random failures) should also be nearby. The second review, more importantly, lowered the median seismic capacity for each of the seismic initiators and some of the other seismic failures to ensure that the model would properly generate valid cutsets. For example, the LLOCA fragility was reduced to 0.5g to generate LLOCA cutsets. For ATWT, the fragility of the CRDs and RV internals were reduced to 0.5g to verify that valid A TWT cutsets were generated. 16-15 SPR-E6, Met, Documentation of The current documentation does Expand the The LERF documentation SPR-F2 Met LERF model not explain what are the basis for documentation to ensure in the QU report was applicability review. retaining the LERF logic and that the criteria used to expanded to describe the analysis unchanged within the retain the LERF analysis review of applicability of SPRA logic. in the SPRA is explained the internal events PRA so that the same LERF analysis to the (This F&O originated During the peer review the applicability review can seismic PRA. from SR SPR-F2) following explanation was be performed following provided by the SPRA team: future potential revisions This finding has been of the LERF modeling. resolved with no significant "The internal events Level 2 impact to the SPRA results notebook was reviewed to ensure or conclusions. that the definition of LERF would be appropriate for seismic events. The LERF definition includes the use of a 12 hour time period for release after event initiation, to allow for evacuation. This time E1-79

Enclosure 1 Attachment 3 - Resolution of the VEGP Seismic PRA Peer Review Findings F&O Review cc F&O F&O F&O Disposition for GSl-191 Number Element Description Basis Resolution and Other Applications period is considered to be valid for Vogtle seismic events, particularly due to the very low population density in the area. Other characteristics, such as bypass and scrubbing, are the same for seismic as for internal events. The logic for the internal events LERF model is very straightforward, with sequences from the GDF model ANDed with the appropriate LERF fault tree. This logic is also appropriate for seismic events." 16-18 SPR-88 Ill Very small LOCA The DB has a specific entry for To the peer review team Additional information on have been screened the incore thermocouples and knowledge Vogtle is the the walkdown for very from the analysis provides pictures of them. Still, in- only plant that has small LOCA has been based on core thermocouple tubing is not elected to perform added to fragility report to walkdowns but little the only possible source of very dedicated walkdowns in provide the basis for the documentation small LOCA that is envisioned support of not modeling VSLOCA screening. exists of such and the only documentation of very small LOCA. This walkdowns. addressing the other potential would be a best practice This finding has been sources is in the quantification but it also behooves to resolved with no significant notebook: the SPRA team to impact to the SPRA results provide detailed or conclusions. (This F&O originated "For Vogtle 1&2, the seismic documentation of such from SR SPR-88) walkdowns inspected and walkdowns and how they photographed a large sample of supported a systematic the small piping and tubing lines evaluation of the potential connected to the primary system sources of very small in order to identify any LOCA. weaknesses. The piping was iudQed to be ruQQed." E1-80

Enclosure 1 Attachment 3 - Resolution of the VEGP Seismic PRA Peer Review Findings F&O Review cc F&O F&O F&O Disposition for GSl-191 Number Element Description Basis Resolution and Other Applications 16-2 SFR-C1, I/II, Fragilities were not The 2014 hazard was only used During the peer review The fragilities have been SPR-E1 Met corrected to reflect as input to FRANX for the final the SNC staff answered a recalculated based on the the 2014 hazard quantification. It is understood question on this topic by 2014 hazard and the new used for that the fragility estimates have performing an initial values incorporated into quantification. (This been performed based on the limited investigation of the the SPRA model and F&O originated from 2012 hazard. While it is not effect on fragilities quantification. SR SPR-E1) expected nor recommended to correction to reflect the regenerate all the fragility work 2014 hazard and This finding has been with the new hazard, some concluded that the effect resolved. consideration on the possible of this scaling is not change in fragility due to the use insignificant (especially of the newer hazard should be for LERF). It is made. recommended to continue and expand this investigation to make the quantification fully consistent with the fragility values. 16-4 SPR-82 Not The effect of seismic There is no assessment of the While it is recognized that The methodology used for Met impact on effect of changing the breaking the industry is still the seismic HRA analysis performance points in the Surry method. The developing methods in is based on defining PSFs shaping factors is Surry method is based on support to this particular as a function of seismic considered in the methods used in the past at topic (e.g., recently hazard level (bins), which analysis by the SONGS and Diablo Canyon and published EPRI HRA is consistent with the EPRI usage of the Surry the 0.8g breaking point was method for external seismic HRA guidance in method. developed for California events), some additional EPRI 3002008093. The earthquakes. In the Vogtle considerations should be Integrated PSFs and bins analysis there is no indications on done to understand the (breaking points) have whether the breaking point at effect of HEPs in the been updated with (This F&O originated 0.8g is.also applicable to Vogtle. model rather than simply additional breaking points from SR SPR-82) There are also no sensitivity implementing the Surry and integrated PSFs to analyses that would support method as is. reflect seismic binning whether a change in the breaking applicable to Vogtle, in points is significant or not. Three examples for accordance with this addressing this finding finding and consistent with may be the following: the EPRI guidance. The

1. Perform sensitivities on updated values have been E1-81

Enclosure 1 Attachment 3 - Resolution of the VEGP Seismic PRA Peer Review Findings F&O Review cc F&O F&O F&O Disposition for GSl-191 Number Element Description Basis Resolution and Other Applications the values of the applied to both internal multipliers and the g events HFEs and seismic-levels where the breaking unique HFEs within the point happens. plant response model.

2. Use a different There was no significant multipliers method with impact on the SPRA more breaking points. results.

3. Apply the impact of seismic specific PSF at This finding has been the individual PSF level resolved.

(i.e., timing, stress, etc.) in the HRA calculator. 16-5 SPR-81, Met, LOCA modeling and The selection of the fragility data Documentation on the LOCA basis has been re-SPR-F1 Not fragility selection not used for all LOCA is discussed in use of fragility in support evaluated and updated. Met clearly documented. Appendix 8.2 of the quantification to LOCA should be This was partially due to notebook but is confusing in the clarified to better seismic fragility update mapping of selected fragilities represent the rationale and partially a matter of with specific failures. selected and potentially adding amplifying (This F&O originated addresses the modeling information to the LOCA from SR SPR-F1) It appears that the fragility uncertainties associated basis. The quantification selected to represent LOCA with this selection. report includes updated sequences are coming from documentation. Although specific components but then While this finding is LOCAs are a significant they are used to represents sort expected to be addressed contributor to the SPRA of surrogate events for potential via documentation, some results, the VEGP SCDF failures along the piping network. additional suggestions and SLERF are sufficiently are provided, such as: small that further LOCA Using localized events as modeling sensitivity surrogate for pipe network failure 1. Perform a sensitivity to beyond what has been is probably conservative and may show that the modeling provided in the updated not be fully consistent with the approach described is not model quantification is not system success criteria and significantly skew the warranted. modeling in the internal events results for seismic; modeling. For example, the This finding has been seismic-induced MLOCA fragility 2. Modify the logic by resolved. seems to be based on failure of mapping the seismic-the pressurizer surge line, which induced MLOCA to a E1-82

Enclosure 1 Attachment 3 - Resolution of the VEGP Seismic PRA Peer Review Findings F&O Review cc F&O F&O F&O Disposition for GSl-191 Number Element Description Basis Resolution and Other Applications is a localized failure. The seismic- different position in the induced MLOCA initiator is logic (e.g., a dummy mapped to the internal events event can be entered in MLOCA initiator. The internal the model to provide a events logic for MLOCA has a target for the FRANX split fraction that divides MLOCA injection). (and LLOCA) in four 25% contributors impacting all four CL/HL. Since the seismic-induced MLOCA is a localized failure, the internal events logic is not fully applicable (probably slightly conservative). Because the documentation is potentially leading to a misunderstanding of the selected approach (thus impacting ease on update), this F&O is considered a finding against the documentation SR. 16-6 SPR-B2 Not The effect of seismic The Vogtle SPRA elected to use Expand the IPSF The methodology used for Met impact on Integrated Performance Shaping approach to all the the seismic HRA analysis performance Factors (IPSF) multipliers. While operator actions credited is based on defining PSFs shaping factors is this approach was used for the in the SPRA. as a function of seismic not considered for HEPs that were carried over from hazard level (bins), which any action that was internal events, it was is consistent with the EPRI explicitly added for systematically not done for all the seismic HRA guidance in the SPRA (e.g., actions explicitly added for EPRI 3002008093 flood isolation or DG seismic. [9]. The Integrated PSFs output breaker and bins (breaking points) closure). Based on discussion during the have been updated to peer review, the analyst believed reflect seismic binning that having designed these applicable to Vogtle, in actions for specific scenarios accordance with this following a seismic event, the finding and consistent with (This F&O originated impact of seismic specific PSF is the EPRI Quidance. The E1-83

Enclosure 1 Attachment 3 - Resolution of the VEGP Seismic PRA Peer Review Findings F&O Review cc F&O F&O F&O Disposition for GSl-191 Number Element Description Basis Resolution and Other Applications from SR SPR-B2) already included. updated values have been applied to both internal The objection to this conclusion is events HFEs and seismic-that the seismic specific PSF unique HFEs within the should realistically change with plant response model. the magnitude of the event. This There was no significant change addresses the change in impact on the SPRA the overall context of the plant results. when a small seismic event happens as opposed to when a This finding has been very large seismic event happens. resolved. This seems not to be captured by the approach selected for the Vogtle SPRA. One example of this is that an action that has a 30 minute Tsw (S-OA-BKR-LOCAL) maintains an HEP of 1.60E-03 at all g levels, including the %G14 interval (i.e., >2g). It is understood that this is not expected to be quantitatively significant because failure of the recovered equipment is taken care by the logic model. 16-7 SPR-E2 Met Base case seismic Both CDF and LERF are LERF at 1E-11 truncation LERF truncation, which LERF does not meet truncated at 1.0E-09 with 1000 meets the QU-83 was already considered in the truncation cutsets managed by ACUBE. This truncation requirement. sensitivity studies, has requirements from meets the QU-B3 requirement for Rename LERF at 1E-11 been revised appropriately QU-B3. CDF but not for LERF. as the base case for to meet QU-B3. A new LERF. LERF truncation limit has been established consistent with the LERF (This F&O originated results. Quantification is at from SR SPR-E2) 1E-12, which is a suitably low value. E1-84

Enclosure 1 Attachment 3 - Resolution of the VEGP Seismic PRA Peer Review Findings F&O Review cc F&O F&O F&O Disposition for GSl-191 Number Element Description Basis Resolution and Other Applications This finding has been resolved. 16-8 SPR-E2 Met Missing There is a description of the most While it is understood that The QU report has been documentation of important scenarios but there is the Draft. B version of the updated to document the cutsets review (cfr. no cutset-by-cutset review. quantification notebook is review of both dominant QU-D1) still somewhat a work in cutsets and non-significant process, it is expected cutsets for both CDF and that when the model LERF. reaches a more stable (This F&O originated state documentation of This finding has been from SR SPR-E2) the review of the cutsets resolved with no significant is going to be part of the impact to the SPRA results documentation. or conclusions. 16-9 SPR-B1, Met, Screening values At the time when the An appropriate resolution The seismic HRA analysis SPR-B4b Met used for the HEPs documentation was provided for of this F&O is pending the has been revised to be that (at the time of peer review, the most significant current evolution of the consistent with the EPRI the provided operator actions (i.e., flood model and the seismic HRA guidance in documentation) isolation of ACCW HX) were all importance of operator EPRI 3002008093. The were in the most screening values, which would actions in the SPRA. original screening HEPs significant cutsets. only meet CCI for HR-G1 (directly Given the expectation have been updated using called through SPR-B1 ). that operator actions will the HRA Calculator, be needed to mitigate the consistent with the In addition, there is little importance of relay approach used in the (This F&O originated documentation or supporting chatter (not yet included VEGP internal events from SR SPR-B1) evidence to justify screening in the SPRA logic model) PRA. The Documentation values as low as 3.00E-2 this F&O was provided to has been updated. ensure care is used in the Operator response to relay generation of HEPs if chatter has been they appear in important addressed and evaluated cutsets and also to within the same process, provide more justification and not found to be for screening values less important. than 1.00E-1 because a low screening value may This finding has been indeed skew the actual resolved. importance of the newly qenerated HEP. E1-85

Enclosure 1 Attachment 3 - Resolution of the VEGP Seismic PRA Peer Review Findings F&O Review cc F&O F&O F&O Disposition for GSl-191 Number Element Description Basis Resolution and Other Applications 17-1 SPR-81 Met The documentation The modeling approach injected A separate section in the The discussion of accident does not specifically seismic fragilities into fault trees documentation that sequences and success address the that were modified from the specifically addresses criteria has been applicability of the internal events PRA model. It can accident sequences and expanded, and specific internal events be inferred from this approach, success criteria is needed descriptions of the flooding accident sequences and it was verified by discussions to collect the information scenarios has been added. and success criteria with the staff, that the internal in one logical place, and This finding is to the SPRA model, events sequences and success is needed to support documentation only and and does not criteria were considered to be effective peer reviews does not impact Seismic properly document applicable to the SPRA model. and future model PRA model results. the accident This was not specifically stated in updates. sequences created the documentation. This finding has been specifically for the resolved with no significant SPRA model. Further, several additional impact to the SPRA results seismic flooding sequences were or conclusions. added to the fault tree. These sequences are not discussed (This F&O originated from an accident sequence and from SR SPR-81) success criteria perspective. Inspection of the fault tree and discussions with the staff indicate that the sequences were appropriately developed with specific success criteria that is different from other internal events sequences. The development of these sequences needs to be included in the documentation. Including event trees for these sequences would also aid in a reader's understandinQ. E1-86

Enclosure 1 Attachment 3 - Resolution of the VEGP Seismic PRA Peer Review Findings F&O Review cc F&O F&O F&O Disposition for GSl-191 Number Element Description Basis Resolution and Other Applications 17-2 SPR-E2, Met, The processes used Examples include: Expand the Documentation for QU SPR-F2 Met to create the documentation to clearly results has been im-presented The top cutsets shown in the explain the post- proved to describe the quantification results quantification report are produced processing of the results processes used to ag-are not fully by combining the cutsets from all generated by CAFTA and gregate results over the 14 documented. the seismic interval cutsets in a FRANX. Examples hazard intervals. The process that is not documented. include: importance calculations have been re-quantified While the process used to obtain - Explain how the cutsets and the method for (This F&O originated the importance measures is generated by FRANX are presentation documented. from SR SPR-F2) documented, discussions with the combined into g-level-PRA staff indicated that independent cutsets. This finding has been importances for some of the basic resolved with no significant events were obtained in a - Explain the post- impact to the SPRA results different manner (setting to one or processing used to or conclusions. zero and requantifying). This is generate importance not documented in the notebook. measures, especially focusing on the deviation from a normal practice that is currently only mentioned in the notebook. 17-3 SPR-83, I/II, Subdividing To account for similar equipment The impact of the The non-minimal cutsets in SPR-E4 I/II correlation groups that has different fragilities due to retention of these non- the peer reviewed model based on different building locations, minimal cutsets on were identified and weaker/stronger certain correlation groups were CDF/LERF and reviewed for impact, and components subdivided to assign a seismic importance measures determined to be non-resulted in retention capacity to a weaker component should be assessed and significant to risk. The of non-minimal that only failed that component. the results documented, results were very slightly cutsets in some The higher capacity was then or a method to remove conservative due to these cases, which could assigned to both components, the non-minimal cutsets non-minimal cutsets. The impact CDF/LERF and was effectively the correlated should be devised. Each issue has been addressed results as well as failure of both components. This subdivided correlation in the updated model, such model importance can result in the retention of non- group should be non-minimal cutsets no measures. The minimal cutsets in some cases. investigated for similar longer appear. magnitude and For example, for the Containment effects. acceptability of Fan Cooler Units there are This finding has been E1-87

Enclosure 1 Attachment 3 - Resolution of the VEGP Seismic PRA Peer Review Findings F&O Review cc F&O F&O F&O Disposition for GSl-191 Number Element Description Basis Resolution and Other Applications these impacts was cutsets in which, due to other resolved with no significant not documented. failures, only one containment fan impact to the SPRA results cooler needs to seismically fail to or conclusions. cause core damage. Inspection of the cutsets shows that two otherwise identical cutsets, are retained: one in which the 1Fan (This F&O originated 'group' occurs, and one in which from SR SPR-E4) the 4Fans group occurs. The 4Fans cutset is not minimal, and should not be included in the results. Discussions with the staff indicated that these non minimal cutsets were noted during the quantification review process, but were thought to not greatly impact overall results. No formal assessment was done, however, and no record of the informal assessment was included in the documentation. 17-4 SPR-E6 Met No quantitative A quantitative analysis is required Perform the analysis and The quantitative analysis analysis of the to meet CCII for LE-F1 & LE-G3, include the results in the of significant LERF plant relative contribution which are directly called from quantification notebook. damage states and to LERF from Plant SPR-E6. contributors has been Damage States and performed. A table and Significant LERF associated discussion of contributors was plant damage states and presented in the significant contributors has quantification been added to the LERF results. QU documentation to resolve this finding. (This F&O originated from SR SPR-E6) This finding has been resolved. E1-88

Vogtle Electric Generating Plant - Units 1 & 2 Supplemental Response to NRC Generic Letter 2004-02 Enclosure 3 Risk Quantification _ I -

Enclosure 3 Risk Quantification Table of Contents 1.0 Introduction 2.0 Scope of Risk Assessment 3.0 Failure Mode Identification 4.0 PRA Model Changes 5.0 Sub-Model Development 6.0 Scenario Development 7.0 Debris Sources 8.0 Chemical Effects 9.0 Debris Transport 10.0 Strainer Debris Impact Evaluation 11 .0 Debris Penetration Evaluation 12.0 Debris Penetration Effects 13.0 Sub-Model Integration 14.0 Systematic Risk Assessment 15.0 References E3-1

Enclosure 3 Risk Quantification 1.0 Introduction Figure 3-1 shows a high-level summary of the GSl-191 risk quantification including the interface between the probabilistic risk assessment (PRA) evaluation, the phenomenological evaluation, and the risk guidelines from RG 1.174 (Reference 1).

4. PRA quantification of CDF & LERF with no GS!-191 failures

2. Evaluation of GSI*

1. PRA Identification 191 phenomena for of accident scenarios each high likelihood 3. PRA quantification

5. Calculate llCDF &

and high likelihood configuration to of CDF & LERF with equipment lllERF determine CfPs for GSl-191 failures configurations each strainer/core failure basic event 7a. Identification of

8. Output to analytical refinements to submittal analysis documentation 7b. Identification of potential plant modifications (physical or procedural)

Figure 3 Flow Chart Illustrating Risk-Informed GSl-191 Evaluation The following steps are included in this flow chart:

• Step 1: PRA identification of accident scenarios and high likelihood equipment configurations Step 1 is an important part of the overall evaluation because it defines the scope of the GSl-191 problem. This step has two important parts: identification of accidents requiring recirculation through the emergency core cooling system (ECCS) strainers, and identification of high likelihood equipment configurations.

As discussed in Enclosure 1, Section 2.1, the VEGP risk-informed GSl-19'1 evaluation considered all breaks that require strainer recirculation, which includes loss of coolant accidents (LOCAs) and secondary side breaks inside containment (SSBls). For the accidents that need to be evaluated, it is important to identify the high likelihood equipment configurations. For example, because VEGP has two E3-2 ________ _J

Enclosure 3 Risk Quantification 100 percent capacity independent trains of ECCS and containment spray system (CSS) including a containment spray (CS) pump, a residual heat removal (RHR) pump, a safety injection (SI) pump, and a charging pump, there could be a large number of random failure combinations unrelated to GSl-191 issues (i.e., failure to start or failure to run). However, some equipment failure combinations may have sufficiently low probability that they can be addressed with a bounding analysis. The high likelihood equipment configurations are described in Section 6.3.

• Step 2: Evaluation of GSl-191 phenomena for each high likelihood scenario to determine CFPs for each basic event Step 2 is the process of using NARWHAL to calculate each of the conditional failure probabilities (CFPs) required by the PRA model. The details of this analysis are expanded upon in Section 13.0.
• Step 3: PRA quantification of CDF and LERF with GSl-191 failures Step 3 is a straightforward quantification of the core damage frequency (CDF) and large early release frequency (LERF) based on the GSl-191 failures for the as-built/as-operated plant.

As discussed in Enclosure 1, Section 2.2, some minor modifications to the base PRA were necessary to perform this evaluation. Specifically, the GSl-191 PRA model includes basic events for both strainer failures and core failures to capture the GSl-191 CFPs. The modifications are described in more detail in Section 4.0.

• Step 4: PRA quantification of CDF and LERF with no GSl-191 failures Step 4 is essentially identical to Step 3, but all GSl-191 CFPs were set to zero to represent a hypothetical modification to the plant that would prevent any GSl-191 failures (e.g., removal of all sources of fiber debris).
• Step 5: Calculate ~CDF and ~LERF Step 5 is a simple comparison of the values calculated in Steps 3 and 4. The change in CDF (~CDF) and change in LERF (~LERF) values represent the risk associated with GSl-191.
• Step 6: Comparison to RG 1.174 Guidelines Step 6 is a comparison of the CDF, LERF, ~CDF, and ~LERF values to the acceptance guidelines defined in RG 1.174 (Reference 1). The ~CDF limits are defined as 1x1 o-5/yr for the boundary between Regions I and II (low risk), and 1x1 o-6/yr for the boundary between Regions II and Ill (very low risk). The ~LERF E3-3

Enclosure 3 Risk Quantification limits are one order of magnitude lower than the ~CDF limits (1 x1 o-6/yr for the boundary between Regions I and II, and 1x1 o-7/yr for the boundary between Regions II and Ill). Ideally, the evaluation would show that the risk associated with GSl-191 is in Region Ill. However, as described in RG 1.174, Region II is acceptable for a risk-informed submittal (Reference 1).

• Step 7a: Identification of analytical refinements to analysis In general, risk-informed evaluations should not contain significant conservatisms because conservatism skews the results in a way that might mask the true sources of risk. However, in practice, some level of bias that is generally accepted to be conservative is incorporated in risk-informed evaluations to avoid the extensive analysis and testing that would be required to develop more refined inputs and models. Therefore, if the results of the risk-informed evaluation are unacceptable, Step 7a can be used to identify specific conservatisms and implement refinements that may reduce the calculated risk to an acceptable level.

Over the course of developing the VEGP risk-informed GSl-191 evaluation, a variety of refinements and simplifications were incorporated for different aspects of the model to balance realism and conservatism.

• Step 7b: Identification of potential plant modifications (physical or procedural)

Step 7b is very similar to Step 7a, but is predicated on the analysis being as refined as practical (i.e., the inputs and analytical models used in the risk-informed evaluation are considered to be a reasonable representation of the post-accident conditions). In this case, if the risk is determined to be unacceptably high, it is necessary to make one or more plant modifications. The modifications could include physical changes (strainer replacement, insulation replacement, buffer change, degraded coatings remediation, etc.) and/or procedural changes (securing pumps earlier in the event, initiating hot leg recirculation earlier, etc.). If a plant modification is necessary, various options can be evaluated to determine the impact on risk, as well as the cost-benefit of implementing each option. Note that even if Step 6 shows that the risk is acceptably low, it may be beneficial to evaluate and implement some minor modifications that significantly reduce risk. For VEGP, a combination of physical and procedural modifications were determined to be beneficial. As described in Enclosure 2, Section 3.j, the RHR strainer height will be reduced by removing disks, and the procedures for switching over from refueling water storage tank (RWST) injection to sump recirculation are being modified to inject more water into containment for breaks that do not initiate containment sprays. These modifications will help ensure that the strainers are completely submerged for an increased number of postulated E3-4

Enclosure 3 Risk Quantification* LOCA scenarios, which significantly reduces the likelihood of debris-related failures.

• Step 8: Output to submittal documentation Once the risk has been determined to be acceptable, a licensing submittal can be prepared to document the risk quantification, defense-in-depth, safety margin, etc. This submittal is documented for VEGP in Enclosures 1 through 5.

2.0 Scope of Risk Assessment As discussed in Enclosure 1, Sections 2.0 and 3.0, the scope of the risk model includes:

• Reactor coolant system (RCS) pipe breaks resulting in small, medium, and large LOCAs (includes breaks ranging from %" partial breaks to double-ended guillotine breaks (DEGBs) on every Class 1 ISi weld within the first isolation valve) o This includes breaks due to seismically-induced LOCAs.
• SSBls that result in a consequential LOCA upon failure to terminate safety injection or a stuck open power operated relief valve (PORV) (includes DEGBs at least every 5 ft on main steam line and feedwater line piping inside containment)

Water hammer LOCAs, non-piping LOCAs, and breaks past the first isolation valve were qualitatively addressed or screened out as described in Enclosure 1. In addition, as discussed in Enclosure 1, the risk model was evaluated for Mode 1, which is bounding for Modes 2 through 6. 3.0 Failure Mode Identification The specific GSl-191 failure modes that were included in the risk model are:

• Strainer head loss exceeds the net positive suction head (NPSH) margin for the RHR and CS pumps when the strainer is fully submerged.
• Strainer head loss exceeds half of the submerged strainer height when the strainer is partially submerged.
• Strainer head loss exceeds the strainer structural margin.
• Gas voids downstream of the strainers exceed the acceptable void fraction limits of the RHR or CS pumps.
• Debris accumulation on the strainer exceeds tested debris limits. Although this by itself is not a unique failure mode, it represents an unknown head loss condition where any of the failure modes above may occur.
• Debris accumulation in the core exceeds debris limits for core blockage and boric acid precipitation.

Note that containment sprays are not required for containment cooling. Therefore, although CS pump failures were evaluated in the NARWHAL model, these failures were E3-5

Enclosure 3 Risk Quantification not included in the CFP results. As discussed in Enclosure 1, Section 1.0, other failure modes (upstream blockage, vortexing, ex-vessel downstream effects, and the LOCA deposition model (LOCADM) portion of the analysis of in-vessel effects) were addressed in a bounding manner for the range of possible breaks with no issues of concern, and were therefore not explicitly modeled in NARWHAL. Note that no significant direct or indirect effects associated with these excluded failure modes have been identified with respect to the risk model. 4.0 PRA Model Changes To perform the risk-informed GSl-191 evaluation, a few relatively minor changes were required for the base PRA model to incorporate the events for the GSl-191 sump strainer and core blockage failures, along with the associated LOCA initiating events and equipment configurations. The following configurations were represented in the NARWHAL evaluations to determine CFPs for GSl-191 related strainer and core failures, and were included in the modifications to develop the VEGP GSl-191 PRA model:

• No equipment failed (all ECCS trains operating)
• One RHR train (RHPA or RHPB) failed
• One charging (CP) train (CPPA or CPPB) failed
• One SI train (SIPA or SIPS) failed
• One nuclear service cooling water (NSCW) train (NSCWA or NSCWB) failed, causing failure of one ECCS train
• One CS train (CSPA or CSPB) failed
• Two CS trains (CSPA and CSPB) failed Each potential sump strainer and core blockage failure can be represented in the PRA model with a basic event that is combined with the appropriate LOCA initiating event and pump failure logic to represent the equipment configurations listed above. Some simplifications of the modeled configurations were performed based on the assumptions described below:
• The CS pumps were assumed not to actuate during a medium or small LOCA; thus, configurations with one or both CS trains failed were modeled for large LOCA initiating events, and were also modeled for the medium LOCA initiating events for sensitivity purposes only.
• CS is not required to be modeled in the VEGP PRA for containment cooling; therefore, the sump strainers for the CS pump recirculation suction, and associated GSl-191 failures, were also not included in the VEGP GSl-191 PRA model.
• Because the charging and SI pumps piggyback off the RHR pumps, failure of one charging or one SI train has the same impact on debris transport to the RHR sump strainer as the configuration with all ECCS trains operating; therefore, these configurations were grouped together.

E3-6

Enclosure 3 Risk Quantification The VEGP base PRA model currently includes events to represent independent and common cause plugging of ECCS containment sumps A and B. For convenience, these existing sump-plugging events were combined under new logic gates NON-GSl-191-A and NON-GSl-191-B. These existing sump plugging events may ultimately be removed from the VEGP PRA model of record and replaced with the more detailed representation of GSl-191 sump strainer failure logic described below. The logic developed for GSl-191 RHR sump strainer failures was modeled under new fault tree gates GSl-191-SUMP-A, GSl-191:.SLJMP-B, and GSl-191-SUMP-AB. Although the sump strainer CFPs may be zero for some of the scenarios included in the logic, these events were retained for potential sensitivity studies. The new logic gate for GSl-191 failure of RHR Strainer A (GSl-191-SUMP-A) was included under the existing PRA model gates for loss of flow from the ECCS containment sump A (ECCS-SUMP-A and ECCS-SUMP-A-ACR). Similarly, the new logic gate for GSl-191 failure of RHR sump strainer B (GSl-191-SUMP-B) was included under the existing PRA model gates for loss of flow from the ECCS containment sump B (ECCS-SUMP-B and ECCS-SUMP-B-ACR). Finally, the new logic gate for GSl-191 failure of both RHR sump strainers A and B (GSl-191-SUMP-AB) was included under the existing PRA model gates for loss of flow from sump A and from sump B (ECCS-SUMP-A, ECCS-SUMP-A-ACR, ECCS-SUMP-B, and ECCS-SUMP-B-ACR). The logic developed for GSl-191 core blockage was modeled under new fault tree gate GSl-191-CORE. As with the sump strainer CFPs, the core blockage CFPs may be zero for some of the scenarios included in the logic; however, these events were also retained for potential sensitivity studies. The new logic gate for GSl-191 core blockage (GSl-191-CORE) was added under the existing PRA model top gate for core damage (GDF-TOTAL). A GSl-191 core blockage logic gate was also added under the existing PRA model gates for LERF end states 01 through 08 (gates LERF-01 through LERF-08). End states LERF-01 through LERF-06 can result from small LOCAs only; therefore, the small LOCA core blockage gate (GSl-191-CORE-SL) was combined with the containment event tree logic associated with each of those end states (e.g., gate LERF01 X ... LERF06X). End state LERF-07 can result from medium or large LOCAs; therefore, the medium and large LOCA core blockage gates (GSl-191-CORE-ML "OR" GSl-191-CORE-LL) was combined with the containment event tree logic for end state LERF-07 (gate LERF07X). Finally, end state LERF-08 can result from small, medium, or large LOCAs; therefore, the core blockage gate for all LOCAs (GSl-191-CORE) was combined with the containment event tree logic for end state LERF-08 (gate LERF08X). Proceduralized operator actions for switching from RWST injection to sump recirculation, switching from cold leg recirculation to hot leg recirculation, and securing containment sprays were all included in the risk-informed GSl-191 evaluation. However, no operator actions were credited to recover from the effects of debris-related failures on the strainer or the core in the b.CDF and b.LERF calculation.* E3-7

Enclosure 3 Risk Quantification 4.1 Seismic PRA Model To quantify the contribution of seismically-induced LOCAs to GSl-191 risk, the initiating event frequency for such LOCAs was determined by considering the seismic fragilities for several RCS components whose failure could cause a LOCA. While LOCAs are generally related to piping failures for the internal events PRA, the piping fragility analyses for VEGP demonstrated that the RCS loop piping and the Nuclear Class 1 piping inside containment have high seismic capacity. Other systems with LOCA sensitive piping, including the chemical volume and control system (CVCS), RHR, SI, reactor vessel head vent, and reactor vessel level indication system (RVLIS), were also considered for the potential occurrence of a seismically-induced LOCA within the size range of concern. This ensured seismically-induced pressure boundary failure was considered for a range of components to identify the bounding median seismic capacity for potential LOCAs. For development of a seismically-induced LOCA frequency, the seismic capacity of the reactor coolant pump (RCP) was selected. The RCP has the lowest seismic capacity among the nuclear steam supply system (NSSS) components. The selected bounding seismic capacity is based on an indirect seismically-induced LOCA due to the failure of the RCP column assembly support, which provides the least seismic margin of safety. Due to the flexibility and strength of the primary system piping and its supports, and the uncertainty attached to the postulated break size, this indirect seismically-induced failure could range in size from a small LOCA to a large LOCA. Therefore, the failure frequency is divided equally among the three LOCA initiating event sizes modeled in the VEGP PRA. The VEGP internal events PRA model modifications made to perform the risk-informed GSl-191 evaluation were used as a guide to make corresponding modifications to the VEGP seismic PRA model. These modifications included the incorporation of the CFPs calculated by NARWHAL for strainer and core failures due to the effects of debris generated by a large break LOCA. It is assumed that these CFPs are also applicable to a (direct or indirect) seismically-induced LOCA occurring in one location. The indirect seismic LOCAs that are postulated are for a break in the cold leg or crossover leg piping next to any one of the RCPs. The RCP break location is postulated based on a collapse of the RCP support structure following a seismic event. The resulting break could be any size from a small break up to a DEGB of the cold leg or crossover leg. The CFP values were calculated based on a range of break sizes (from Y2 inch to a DEGB) postulated at all Class 1 ISi welds within the first isolation valve. The NARWHAL evaluation showed that none of the small and medium breaks generate sufficient debris quantities to cause GSl-191 failures (including small and medium breaks on the cold legs and crossover legs). Many large breaks also do not generate sufficient debris quantities to cause failures. The large break GSl-191 CFP values were calculated based on the LOCA frequencies for a random pipe break (with a higher frequency for 6-inch breaks and a much lower frequency for 31-inch DEGBs). Although the frequency as a function of break size would likely be different for a seismically-induced LOCA, it is E3-8

Enclosure 3 Risk Quantification reasonably conservative to use the large break CFP values, because a seismic event could result in a small or medium break that would not cause any debris related failures. A seismic event also has the potential to dislodge insulation from piping and components inside containment. The insulation is contained inside a fabric cover; therefore, any dislodged insulation would be expected to fail in relatively-large intact pieces that would be unlikely to transport to the sumps or have a significant effect on the strainers. 5.0 Sub-Model Development The evaluation of failures related to the effects of debris was performed outside the VEGP PRA model using the NARWHAL software. The VEGP NARWHAL model includes the following sub-models:

• Post-accident conditions o Plant configuration o Plant states o Random equipment failures o Water volume and level o Flow rates o Pressure and temperature o pH
• Debris sources o Zone of influence (ZOl)-generated insulation, fire barrier, and qualified coatings debris o Unqualified coatings debris o Latent debris o Miscellaneous debris
• Chemical effects o Release via corrosion/dissolution o Solubility o Precipitate debris quantity
• Debris transport o Slowdown o Washdown o Pool fill o Recirculation o Erosion
• Strainer head loss o Debris groups o Clean strainer head loss o Conventional debris head loss o Chemical debris head loss o Head loss correction o Head loss extrapolation E3-9

Enclosure 3 Risk Quantification

• Strainer air intrusion
• Strainer and pump acceptance criteria o Strainer flashing o Strainer structural margin o Strainer partial submergence limit o Pump void fraction limit o Pump NPSH margin o Debris limits
• Strainer penetration o Fiber o Particulate
• Core acceptance criteria These sub-models are described in more detail in the following sections.

6.0 Scenario Development The post-accident conditions are described in the following sub-sections. These conditions include the plant configuration, plant state changes (due to automatic and manual operator actions), and various thermal-hydraulic parameters that were used in the VEGP NARWHAL model. For all breaks evaluated, a 30-day mission time was used in the NARWHAL model. This is consistent with the mission time used for deterministic GSl-191 evaluations (Reference 2). Note that the VEGP PRA model uses the typical PRA mission time of 24 hours. Any RHR strainer and core failures predicted by NARWHAL (regardless of failure time) were included in the CFP values that were used in the GSl-191 PRA calculation. Although additional operator actions and compensatory measures (such as refilling the RWST) could be taken to mitigate late failures, these actions were not included in either the GSl-191 PRA model or the NARWHAL model. 6.1 Plant Configuration The plant configuration determines the flow paths during the different phases of accident mitigation. At VEGP, there are a total of four separate sump strainer assemblies and two engineered safety features (ESF) trains. An ESF train consists of pumps associated with the ECCS and the CSS. A single train of ECCS consists of a high head centrifugal charging pump, a medium head SI pump, and a low head RHR pump. In addition to these pumps, each train of ECCS contains two SI accumulators. A single train of CSS contains a CS pump. There are two strainers dedicated to each ESF train, one for the ECCS pumps and one for the CS pump. Note that the strainers dedicated to the ECCS pumps are referred to as RHR strainers, and the strainers dedicated to the CSS are referred to as CS strainers. Figure 3-2 shows the connection diagram from NARWHAL that was used to model VEGP. Valves are used in NARWHAL to define whether a flow path is currently active. E3-10

Enclosure 3 Risk Quantification For example, each pump has a connection to the RWST and the strainer. During injection, the valves connecting the RWST to each of the pumps would be open and the valves connecting the pumps to the sump would be closed. The activity state and flow paths associated with each pump are discussed further in Section 6.2. As shown in the figure, the equipment was arranged in a manner that is consistent with the plant-specific emergency operating procedures (EOPs). The ECCS pumps operate in parallel while taking suction from the RWST, and the charging and SI pumps take suction from the RHR discharge during recirculation. Additionally, these pumps can provide flow to the RCS through either the cold legs or the hot legs. The accumulators are aligned to provide rapid cooling through the cold legs. The CS pumps provide flow to the CS headers and can take suction from the RWST during injection and the sump during recirculation. E3-11

Enclosure 3 Risk Quantification Figure 3 VEGP Connection Diagram E3-12

6.2 Plant States A plant state represents each of the different plant modes of post-accident operation and is defined by the activity state of the valves and pumps in the connection diagram. The plant states are consistent with the plant EOPs and are a function of the initiating event. For VEGP, two procedures mandate unique states as a function of break size and break side (i.e., hot leg vs. cold leg side). These procedures address SI accumulator injection and CS activation. As discussed in Section 13.2, the SI accumulators were modeled to only inject for breaks greater than or equal to 2 inches. Also as discussed in Section 13.2, the containment sprays were modeled to activate for hot leg breaks larger than 15 inches and operate for a duration of 24 hours. Note that different states are required for cold leg (CL) recirculation and hot leg (HL) recirculation. During HL recirculation, the RHR and SI pumps are aligned to discharge into the HLs, while the charging pumps continue to discharge into the CLs. Because the charging pump alignment after switchover to HL recirculation does not have any effect on the overall results, the NARWHAL model was simplified to align all ECCS pumps to discharge into the HLs during HL recirculation. Accident mitigation at VEGP was modeled using the states shown in Table 3-1 and Table 3-2. Note that Table 3-1 describes the initiators that dictate which breaks enter a given state, and Table 3-2 describes the operating equipment during each state. In Table 3-1, the term "Full Recirc" is used to describe the state in which all active pumps are taking suction through the sump strainers (or in piggy-back mode). Table 3 Plant State Initiators Min Break Max Break Break State Name Initiating Variable Size Size Side Injection (small break 0 1.999 Both Time= 0 minutes no CS) Injection 1.999 15.001 Both Time= 0 minutes (2<Break<15, no CS) Injection 15.001 31.001 Cold Time= 0 minutes (CL Break> 15, no CS) Injection (CS) 15.001 31.001 Hot Time= 0 minutes ECCS Switchover Level RHR Recirc (CS) 15.001 31.001 Hot in RWST (CS active) Full Recirc ECCS Switchover Level 0 15.001 Both (Break < 15, no CS) in RWST (CS inactive) Full Recirc ECCS Switchover Level 15.001 31.001 Cold (CL Break> 15, no CS) in RWST (CS inactive) CSS Switchover Level in Full Recirc (CS) 15.001 31.001 Hot RWST Hot Leg Switchover 0 15.001 Both Time= 450 minutes (Break< 15, no CS) Hot Leg Switchover 15.001 31.001 Cold Time = 450 minutes (CL Break > 15, no CS)

Enclosure 3 Risk Quantification Min Break Max Break Break State Name Initiating Variable Size Size Side Hot Leg Switchover 15.001 31.001 Hot Time= 450 minutes (CS) CS Termination 0 31.001 Both Time= 1440 minutes Table 3 Plant State Component Activity State Name Component Activity Description RHR Pumps Active Suction from RWST SI Pumps Active Suction from RWST Injection (small break Chan::iing Pumps Active Suction from RWST no CS) CS Pumps Inactive Not Operating SI Accumulators Inactive Not Operating Reactor Vessel Flow CL N/A RHR Pumps Active Suction from RWST Injection SI Pumps Active Suction from RWST (2<Break<15, no CS) Charging Pumps Active Suction from RWST Injection CS Pumps Inactive Not Operating (CL Break>15, no CS) SI Accumulators Active Injection into CL Reactor Vessel Flow CL N/A RHR Pumps Active Suction from RWST SI Pumps Active Suction from RWST Charging Pumps Active Suction from RWST Injection (CS) CS Pumps Active Suction from RWST SI Accumulators Active Injection into CL Reactor Vessel Flow CL N/A RHR Pumps Active Suction from Sump SI Pumps Active Suction from RWST Charging Pumps Active Suction from RWST RHR Recirc (CS) CS Pumps Active Suction from RWST SI Accumulators Inactive N/A Reactor Vessel Flow CL N/A RHR Pumps Active Suction from Sump Full Recirc SI Pumps Active Suction from RHR (Break<15, no CS) Charging Pumps Active Suction from RHR Full Recirc CS Pumps Inactive Not Operating (CL Break>15, no CS) SI Accumulators Inactive N/A Reactor Vessel Flow CL N/A Full Recirc (CS) RHR Pumps Active Suction from Sump E3-14

Enclosure 3 Risk Quantification State Name Component Activity Description SI Pumps Active Suction from RHR Charging Pumps Active Suction from RHR CS Pumps Active Suction from Sump SI Accumulators Inactive N/A Reactor Vessel Flow CL N/A RHR Pumps Active Suction from Sump Hot Leg Switchover SI Pumps Active Suction from RHR (Break<15, no CS) Charging Pumps Active Suction from RHR Hot Leg Switchover CS Pumps Inactive Not Operating (CL Break>15, no CS) SI Accumulators Inactive N/A Reactor Vessel Flow HL N/A RHR Pumps Active Suction from Sump SI Pumps Active Suction from RHR Hot Leg Switchover Charging Pumps Active Suction from RHR (CS) CS Pumps Active Suction from Sump SI Accumulators Inactive N/A Reactor Vessel Flow HL N/A RHR Pumps Active Suction from Sump SI Pumps Active Suction from RHR Charging Pumps Active Suction from RHR CS Termination CS Pumps Inactive Not Operating SI Accumulators Inactive N/A Reactor Vessel Flow HL N/A 6.3 Random Equipment Failures Random equipment failures are defined as failure to start or failure to run due to issues unrelated to GSl-191. Based on symmetry and the inputs used in the VEGP NARWHAL model, pump failures in one train are analytically identical to the same pump failures in the other train. In addition, although there is a slight difference during the RWST injection phase, the VEGP NARWHAL CFP calculation showed that the GSl-191 CFP results (i.e., the breaks that fail) were identical for the cases with no equipment failures, a single charging pump failure, and a single SI pump failure, because the charging and SI pumps piggyback off of the RHR pumps during recirculation. The switchover of the CS pumps from RWST injection to sump recirculation requires a manual operator action. Due to the human failure probability associated with this action, the probability of losing both CS pumps at the start of recirculation is higher than the probability of a single CS pump randomly failing to start or failing to run. E3-15

Enclosure 3 Risk Quantification Table 3-3 shows the functional failure probabilities for the different equipment configurations from the VEGP GSl-191 PRA model. Note that these overall probabilities are based on the logic described above and have been normalized to 100 percent. Table 3 Functional Failure Probabilities Equipment Configuration Functional Failure Probability No Equipment Failures 91.50% 2 CS Pump Failures 5.31% 1 RHR Pump Failure 1.46% 1 CS Pump Failure 1.26% 1 RHR Pump+ 1 CS Pump Failures 0.39% 1 RHR Pump+ 2 CS Pump Failures 0.07% Total 100% Due to the low functional failure probability, the case with one RHR pump and two CS pump failures was not evaluated explicitly, and the CFP was conservatively set to 1.0 for the risk quantification. 6.4 LOCA Frequencies Table 3-4 shows the mean LOCA frequencies taken from the VEGP GSl-191 PRA model. As discussed in Enclosure 1, Section 3.0, the medium and large LOCA frequencies are based on the geometric aggregation from NUREG-1829 (Reference 3). Table 3 Mean LOCA Frequencies from GSl-191 PRA Model Break Exceedance Size Frequency (yr- 1) 0.375 4.73E-04 2 1.39E-04 6 1.85E-06 31 1.50E-08 Note that the frequencies were used to calculate conditional failure probability as a function of PRA initiating event break size ranges. There are three break size ranges evaluated in the VEGP PRA. The small LOCA category comprises random breaks in the RCS in the range of 3/8-inch to 2-inch equivalent diameter. The medium LOCA initiating event is defined as a break in the RCS that is greater than or equal to 2 inches and less than 6 inches equivalent diameter. The large LOCA initiating event is defined as a break in the RCS that is greater than or equal to 6 inches up to a DEGB of the largest pipe (31 inches) in the RCS. Because the equivalent break size for a 31-inch DEGB is 43.8 inches (31 *-'12 =43.84), the frequency for each PRA size category is: E3-16

Enclosure 3 Risk Quantification

• FsLocA =Fo.375" - F2" =4.73E 1.39E-04 =3.34E-04 yr-1
• FMLOCA =F2" - Fa" =1.39E 1.85E-06 =1.37E-04 yr-1
• FLLOCA =Fa" - F43.B" 1 =1.85E 0 =1.85E-06 yr- 1 6.5 Water Volume and Level The height of the pool is a function of the total quantity of water in the sump. The pool level is calculated as a linear function of pool volume in gallons.

Hpool -- 1.20071 (ft)

• 10 -5 gal
• Vpool - 0.058 ft Equation 1 Nomenclature:

Hpool =the height of the sump pool in ft Vpoo1 = the volume of the recirculation pool in gallons At VEGP, three sources of water contribute to the recirculation pool inventory. These sources are the RWST, the SI accumulators, and the RCS.

• The total quantity of water delivered from the RWST is the difference between the initial and final levels. There are two final levels that are important in the analysis, the low-low level alarm (ECCS recirculation level) and the empty alarm (CS recirculation level). Note that this is different for breaks that do not activate containment sprays. The empty level alarm is the ECCS recirculation level for these breaks.
• Four SI accumulators provide immediate cooling to the core for breaks large enough to rapidly depressurize the RCS.
• A portion of the water initially in the RCS will be released as steam or spill to the pool through the break opening at the beginning of a LOCA.

Several hold-up volumes reduce the sump pool height:

• The amount of water held up as steam (vapor) in the containment atmosphere is a function of time.
• During recirculation, the total hold-up in the RCS is a function of break size and elevation. *
• The containment spray falling through the containment building represents a transitory hold-up volume. Note that this hold-up was only applied while containment spray pumps were operating. After containment spray was terminated, this quantity was returned to the sump pool.

1 Because the exceedance frequency for a 43.8-inch break is not available, the frequency was set to zero, which conservatively maximizes the frequency for large LOCAs. E3-17

Enclosure 3 Risk Quantification

• The break flow falling through containment also represents a transitory hold-up volume that was applied to all breaks.
• The volume of the CS discharge piping is another hold-up that was applied for breaks that initiate containment sprays.
• Other miscellaneous hold-up volumes include the containment sump pits, the elevator pit, and the containment floor drains.
• The reactor cavity and in-core tunnel is one of the largest hold-up volumes. Due to the restricted flow paths into the reactor cavity, this hold-up volume was applied in a time-dependent manner based on the break location and whether containment sprays are activated.

The long-term water level is high enough to fully submerge the RHR and CS strainers for all breaks. For some reactor cavity breaks, there is a short period where the RHR strainers are not fully submerged just after the RHR pumps are switched to recirculation. However, for these breaks, the water level rises enough to submerge the strainers before the CS pumps finish drawing down the RWST. 6.6 Flow Rates The pump flow rates that were used in the VEGP NARWHAL model are design flow rates for the SI pumps, charging pumps, and CS pumps. A flow rate approximately 20 percent higher than the design value was used for the RHR pumps. The break flow rate is the sum of the flow through the RHR, SI, and charging pumps during RWST injection, and is the sum of the RHR pump flow rates during recirculation (when the charging and SI pumps are piggybacking off of the RHR pumps). No credit was taken for reduced RHR flow rates for smaller break sizes. Note that for secondary side breaks in a feed and bleed scenario, only the charging and containment spray pumps were assumed to be active, which results in a significantly lower flow rate. As discussed in Section 6.5, the reactor cavity and in-core tunnel have restricted flow paths. Because of the restricted flow paths and the fact that the volume is fairly large, a function was implemented to model the flow rate into the reactor cavity as a function of time. Note also that the flow rate into the reactor cavity is a function of break location. For the breaks in the reactor cavity, the entire cavity is assumed to fill before the start of recirculation. In addition, the total reactor cavity hold-up volume is larger due to the flow rate inside of the cavity and the height of the flow paths that connect the cavity to the sump. For breaks outside the reactor cavity, the cavity would fill relatively slowly and would not be completely filled until after the start of recirculation. E3-18

Enclosure 3 Risk Quantification The core boil-off flow rate used in the NARWHAL model was calculated based on the core power and the ANSl/ANS-5.1-1979 decay heat curve (Reference 4). An additional 20 percent margin was added to the boil-off flow rate. 6.7 Pressure and Temperature The sump temperature, containment temperature, and containment pressure profiles were used in NARWHAL to determine time-dependent thermal-hydraulic properties. Design basis temperature profiles were used for both sump temperature and containment temperature. These temperature profiles were based on a DEGB in the primary loop piping with minimum safeguards, and therefore represent the maximum temperature profiles. Although the actual temperature profiles would be significantly lower for smaller break sizes, the same temperature profiles were conservatively used for all break sizes. Note that there are competing factors associated with sump temperature, which could result in a lower temperature resulting in more failures. However, based on sensitivity analysis (see Section 14.2.2), it was determined that maximizing the temperature is more conservative for the VEGP model. Containment accident pressure was not credited for NPSH margin calculations in the VEGP NARWHAL model. Because the technical specification minimum containment pressure is -0.3 psig at VEGP, the containment pressure profile was specified to be saturation pressure at pool temperatures above 210.96 degrees F, and 14.396 psia at pool temperatures below 210.96 degrees F. For the purpose of degasification and flashing calculations, up to 3.5 psi of accident pressure was credited. Both flashing and degasification are most problematic when the sump temperature is near or above 212 degrees F. The pool temperature is greater than 212 degrees F for approximately the first 120 minutes after a LOCA. Additional details are provided in Enclosure 2 Sections 3.f.14 and 3.g.14. 6.8 Sump and Spray pH The maximum sump recirculation pH was conservatively rounded up to 7.8 for the VEGP NARWHAL model. Use of the maximum pH provides bounding chemical release quantities from submerged materials. The minimum sump pH was rounded down to 7.0 for the VEGP NARWHAL model. The minimum sump pH was used to calculate the aluminum solubility limit in NARWHAL. Note that using different pH values to calculate release and solubility results in an over-prediction of the actual precipitate quantity. The spray pH during RWST injection was specified to be 5.72, which is the maximum pH of the RWST. Using the maximum RWST pH maximizes chemical release from E3-19 I

Enclosure 3 Risk Quantification unsubmerged sources during the injection phase. After switchover to recirculation, the spray pH is equivalent to the sump pH. 7.0 Debris Sources As described in Enclosure 2, Section 3.a.3, the types of debris in the NARWHAL model include Nukon fiberglass insulation, lnteram fire barrier, and qualified epoxy and IOZ coatings debris generated inside the ZOI. The model also includes unqualified epoxy, IOZ, and alkyd coatings, latent dirt/dust and fiber, and miscellaneous debris, which are present in containment or generated by the post-accident environmental conditions. The debris sources generated inside the ZOI range from essentially zero debris for the smallest break sizes up to 2,229 ft 3 of Nukon, 60 lbm of lnteram, 220 lbm of qualified epoxy, and 65 lbm of qualified IOZ debris for the bounding breaks. The other debris sources are independent of the break location and size and were therefore applied to all breaks. The debris quantities used in the NARWHAL model (including some operating margin) were 2,729 lbm for unqualified epoxy, 56 lbm for unqualified IOZ, 59 lbm for unqualified alkyd, 30 lbm for latent fiber, 170 lbm for latent dirt/dust, and 50 ft2 for miscellaneous debris. A four-category size distribution (fines, small pieces, large pieces, and intact blankets) was used for the Nukon debris based on the guidance in the appendices of NEI 04-07 Volume 2 (Reference 2). A more conservative two-category size distribution (fines and small pieces) was used for the lnteram debris. All of the qualified and unqualified coatings debris was conservatively treated as fines. The latent debris was also treated as fines. The miscellaneous debris was simply treated as a reduction in the total strainer surface area. Note that 25 percent overlap of the miscellaneous debris was credited in the NARWHAL model, which is consistent with the guidance in NEI 04-07 Volume 2 (Reference 2). Debris from all sources was essentially treated as being generated at the beginning of the event. The miscellaneous debris surface area reduction was applied prior to other debris transporting to the strainer. Unqualified coatings were modeled as failing after the pool fill phase (no transport to inactive cavities), but were available to transport at the start of recirculation. This is described further in Section 9.0. 8.0 Chemical Effects As described in Enclosure 2 Section 3.o.2, the formation of chemical products was analyzed as a function of the temperature, pH, and pool volume inputs, as well as debris quantities and exposed aluminum and concrete surface areas. There are two parts to the chemical product generation model: the elemental chemical release from materials in containment, and chemical precipitate formation. Note that these E3-20

Enclosure 3 Risk Quantification processes are based on break-dependent conditions and were therefore analyzed separately for each postulated break. 8.1 Elemental Chemical Release The Nukon and lnteram debris both contribute to chemical release, which was quantified in NARWHAL using the WCAP-16530 release equations (Reference 6) and the break-specific debris quantities. Note that lnteram debris only releases silicon. Therefore, the lnteram has no effect on the chemical product generation because the only aluminum precipitate that is tracked in the VEGP NARWHAL model is sodium aluminum silicate (SAS) (see Section 8.2), and NARWHAL conservatively assumes an infinite source of silicon when SAS is the only aluminum precipitate tracked (Reference 7). The amount of elemental chemical release from a given debris source is limited by the quantity of that element within the source. E-Glass (which includes Nukon) has 1.95 percent aluminum and 2.16 percent calcium by mass. The exposed surfaces include aluminum metal and concrete, which would either be submerged in the containment pool or exposed to containment sprays. The same surface areas were analyzed for each break. The chemical release from exposed concrete was evaluated using the WCAP-16530 release equations (Reference 6), and the chemical release from aluminum was evaluated using the University of New Mexico (UNM) release equations (Reference 8). 8.2 Chemical Product Formation The chemical precipitates analyzed in the VEGP NARWHAL model were SAS and calcium phosphate. The calcium phosphate was modeled in NARWHAL as precipitating immediately when calcium is released in solution (Reference 7). The SAS precipitates when the concentration of aluminum in the pool exceeds the aluminum solubility limit as calculated with the Argonne National Laboratory (ANL) solubility equation (Reference 7). Note that if precipitation of SAS was not predicted before 24 hours, then precipitation was forced at that time. Also note that aluminum was assumed not to remain dissolved in the pool after precipitation occurred (i.e., the aluminum solubility limit was only credited for precipitate timing). Forcing precipitation at 24 hours, as well as not taking credit for aluminum remaining dissolved in the pool, are conservative factors in the chemical product formation model. The effects of the chemical precipitates on strainer head loss are described in Section 10.0, and the effects on core blockage are described in Section 12.0. E3-21

Enclosure 3 Risk Quantification 9.0 Debris Transport As described in Enclosure 2, Section 3.e, debris transport includes the transport of debris during the blowdown, washdown, pool fill, and recirculation phases, as well as debris erosion. Transport can vary significantly as a function of flow rate, water level, etc. However, in the VEGP NARWHAL model, many of these parameters were not explicitly modeled and were conservatively represented using bounding conditions (e.g., the bounding flow rates for large break conditions were used to calculate recirculation transport fractions that were applied to breaks of all sizes). The specific factors affecting transport that were included in the VEGP NARWHAL model include debris type and size, break location (i.e., breaks in the steam generator compartment on the Loop 1/3 side, breaks on the Loop 2/4 side, breaks in the reactor cavity, breaks in the pressurizer compartment, or breaks in the annulus), whether sprays are initiated or not, and whether one or two trains are operating. The blowdown transport fractions are a function of the break location, as well as the size of debris. The only debris transported during the blowdown phase would be debris generated inside the ZOI. Fine debris was transported with the blowdown flow, with no credit for retention on structures. The transport of small and large pieces of Nukon and small pieces of lnteram was dependent on the break location as well as the location of grating that debris would have to be blown through to reach upper containment or the containment floor. The washdown transport fractions were based on containment spray initiation as well as the size of debris. If containment sprays were initiated, 100 percent of fine debris was modeled as being washed down to the containment floor. Some credit was taken for small pieces of Nukon and lnteram being retained in upper containment, and most large pieces of Nukon that were transported to upper containment were modeled as being retained in upper containment. If containment sprays were not initiated, the transport would be significantly reduced. However, 10 percent of fine debris was still modeled as washing down from upper containment due to condensation drainage. For pool fill transport, a relatively small fraction of debris was modeled as transporting to inactive cavities and the ECCS strainers as the sump cavities were filled. These transport fractions were applied to all debris that was in the containment pool at the end of the blowdown phase. This includes debris generated inside the ZOI as well as latent debris, but not unqualified coatings. Recirculation transport fractions were developed based on computational fluid dynamics (CFD) modeling. Several simulations were run to determine the transport fractions for the various types and sizes of debris corresponding to different break locations, number of trains operating, and whether containment sprays were initiated. Small and large pieces of Nukon and lnteram debris that are retained in upper containment would be subject to erosion due to containment sprays. A one percent E3-22

Enclosure 3 Risk Quantification erosion fraction was used for this debris for breaks where containment sprays were initiated. Similarly, small and large pieces of Nukon and lnteram debris in the containment pool would be subject to erosion and a 10 percent erosion fraction was used. Because the debris size is important with respect to penetration, NARWHAL applies the pool erosion for both the debris that settles in the pool as well as the debris that transports to the strainers (Reference 7). This conservatively maximizes the quantity of fine debris. Each of the transport processes described above were used to determine the total quantity of debris that reaches the strainer. However, the timing was not assumed to be instantaneous for all of these processes. Slowdown was treated as an instantaneous process at the beginning of the event. Washdown was treated as occurring after the inactive and sump cavities were filled during the pool fill phase, but before the start of recirculation. Debris that was transported to the ECCS strainers during the pool fill phase was modeled on the strainers at the start of recirculation (note that a fraction of the fine fiber debris penetrates prior to the start of recirculation as described in Section 11.0). For VEGP, failure time was not credited for unqualified coatings, and all of the unqualified coatings debris was treated as being in the pool at the start of recirculation. Although erosion is a time-dependent process, all of the erosion fines were treated as being generated at the start of recirculation. The actual transport to the strainers during the recirculation phase was modeled as a time-dependent process, where debris arrives on the strainers as a function of the pool turnover time (i.e., as a function of the pool volume and strainer flow rates). The debris accumulation on each strainer was proportional to the flow split to each strainer. This flow split was evaluated at each time step. Therefore, the relative accumulation of debris changes over time when various pumps were switched from RWST injection to sump recirculation, containment sprays were secured, etc. 10.0 Strainer Debris Impact Evaluation As discussed in Enclosure 2 Section 3.f.4, the head loss associated with debris accumulation on the strainer was determined using the results of prototypical strainer module testing. There are a total of four separate sump strainer assemblies for each unit at VEGP. There are two for the CSS (i.e., CS strainers) and two for the ECCS (i.e., RHR strainers). Each RHR and CS strainer assembly consists of four parallel vertical stacks connected to a plenum installed over each sump pit. The modified height of each RHR strainer (not including the curb) will be approximately 3.77 ft, and the effective surface area will be 677.6 ft2. The height of each CS strainer (not including the curb) is approximately 3.3 ft, and the effective surface area is 590 ft2. E3-23

Enclosure 3 Risk Quantification 10.1 Strainer Head Loss The conventional and chemical head loss values were corrected based on the sump thermal-hydraulic conditions compared to the test conditions. As specified in the March 2008 Nuclear Regulatory Commission (NRC) guidance (Reference 9), flow sweep data was used to develop the flow and temperature scaling. This scaling was performed at each time step using the time-dependent approach velocity and temperature in the VEGP NARWHAL model. Additionally, the test results were extrapolated to 30 days in accordance with the March 2008 NRC guidance (Reference 9) to account for chemical head loss that had not leveled off by the end of the test. For convenience, the 30-day head loss extrapolation value (which represents a gradual increase over time) was instantaneously applied at 7.5 hours in the VEGP NARWHAL model. Note that the head loss extrapolation value was also corrected based on the time-dependent approach velocity and temperature. The bounding clean strainer head loss of 0.375 ft at 4,500 gpm was used in the VEGP NARWHAL model for all cases. NARWHAL uses a rule-based approach to calculate head loss based on the results of head loss testing with a prototypical strainer module. For the VEGP NARWHAL model, if the fiber debris load on the strainer was less than the tested quantity from the thin bed test, then the maximum thin bed conventional head loss (0.625 ft) was returned. If the fiber quantity is greater than what was tested in the thin bed test, then the maximum full load conventional head loss (5.46 ft) was returned. The head loss effects of calcium phosphate and SAS were each analyzed separately from the conventional debris head loss. Chemical head loss was only applied if the fiber debris quantity on the strainer was greater than 0.45-inch equivalent (see , Section 3.f.10). If this condition was met, and any calcium phosphate accumulated on the strainer, the head loss corresponding to the full quantity of calcium phosphate debris (1.11 ft) was added. Similarly, given the accumulation of sufficient fiber and any SAS on the strainer, the head loss corresponding to the full SAS debris quantity (5.24 ft) was added. In addition, because the extrapolation constant was associated with the chemical head loss, this constant was only applied if the fiber bed was thick enough for the chemical head loss to be added (i.e., greater than 0.45 inches). Because the head loss results are only applicable for debris quantities up to what was tested, debris limits were specified in the VEGP NARWHAL model corresponding to the tested quantities. Separate debris limits were specified for fiber, particulate, lnteram, calcium phosphate, and SAS debris. If any one of these debris limits were exceeded, the strainer was assumed to fail. E3-24

Enclosure 3 Risk Quantification 10.2 Degasification and Flashing NARWHAL calculates degasification and flashing based on the total strainer head loss and other important parameters (containment pressure, sump temperature, water level, etc.) (Reference 7). For the degasification calculations; 2.5 psi of accident pressure was credited. The midpoint of the strainer was used to calculate the average degasification across the entire strainer. NARWHAL performs the flashing calculation at the same reference elevation used to calculate degasification (i.e., the midpoint of the strainer in the VEGP model). However, because the pressure would be lower at the top of the strainer, using the midpoint of the strainer for the flashing calculations is equivalent to crediting additional accident pressure equal to the hydrostatic head from the top of the strainer to the midpoint (approximately 1 psi). Therefore, the VEGP NARWHAL model credits 2.5 psi accident pressure for degasification and approximately 3.5 psi accident pressure for flashing. Note that no accident pressure was credited for NPSH in the VEGP NARWHAL model. The strainer was automatically assumed to fail if any flashing occurs. The degasification calculation results in a gas void fraction that was compared against the pump limits (a void fraction limit of 0.02 was used for all pumps). If the void fraction exceeds a pump limit, the pump was assumed to fail. 10.3 Structural and NPSH Margin The NPSH margin is the NPSH available (excluding strainer head losses) minus the NPSH required. The NPSH available was calculated in the VEGP NARWHAL model based on the time-dependent containment pressure, sump temperature, water level, and major and minor losses in the pump suction piping. As discussed in Section 6.7, a bounding pressure and temperature profile was used, which conservatively minimizes NPSH available. The NPSH required was calculated as a function of the time-dependent flow rate based on the pump curves. The NPSH required was also adjusted as a function of the void fraction as described in RG 1.82 (Reference 10). The strainer head loss was compared against the strainer structural margin (24.0 ft) and the pump NPSH margin at each time step to determine whether a failure occurred. 11.0 Debris Penetration Evaluation As described in Enclosure 2, Section 3.n.1, fiber penetration correlations were developed based on testing with a prototypical strainer module. These equations were used to calculate the fine fiber quantity that passes through the strainer from both prompt penetration and longer-term shedding. A penetration fraction of 0.48 was also applied to the fine fiber that transports to the . strainers during pool fill. As shown in Enclosure 2, Section 3.n.1, this penetration fraction bounds the prompt penetration corresponding to clean strainer conditions. E3-25

Enclosure 3 Risk Quantification 12.0 Debris Penetration Effects As discussed in Enclosure 1, Section 1.0 and Enclosure 2, Section 3.m, ex-vessel downstream effects were addressed in a bounding manner and therefore were not included in the VEGP NARWHAL model. As described in Enclosure 2, Section 3.n.1, core blockage and boron precipitation were evaluated using assumed fiber debris limits and acceptance criteria. VEGP uses Westinghouse fuel and the reactor vessel has an upflow barrel/baffle design with pressure relief holes in the core plates that are not currently credited in the long-term core cooling analyses. Any debris that does not accumulate in the reactor vessel was modeled as automatically returning to the sump pool where it would be available to transport and potentially pass through the strainers again. 12.1 Cold Leg Breaks The fiber debris that penetrates the RHR strainers transports to the reactor vessel through the ECCS flow. During cold leg recirculation, a portion of the ECCS flow that enters the reactor vessel through the cold legs would travel through the core inlet as make-up for boil-off, while the rest of the cold-leg flow would spill from the break. The fraction of fiber debris that was caught on the core inlet was the ratio of the boil-off flow rate to the ECCS flow rate into the cold leg. Note that margin was added to the boil-off flow rate (see Section 6.6). Once switchover to hot leg recirculation had occurred, debris no longer accumulated at the core inlet. Instead, any penetrated debris that entered the reactor vessel was captured incore. However, by the time hot leg recirculation was initiated, most of the fiber fines had bee.n captured on the strainers, and very little additional fiber was transported to the core. Any fiber that was captured in the reactor vessel was assumed to remain there for the duration of the event. For cold leg breaks, an in-vessel failure was recorded if the core inlet fiber load was greater than the specified limit (see Enclosure 2, Section 3.n.1 ). 12.2 Hot Leg Breaks For hot leg breaks, a significant quantity of fiber could accumulate on both the core inlet as well as incore. Flow would either enter the core inlet or the alternate flow path based on the head loss due to debris-related resistance across the core inlet at the bottom core plate. To determine the flow split for the ECCS flow that entered the reactor vessel through the cold leg, several preliminary calculations were performed. Initially, a Ksplit variable was calculated based on an assumed function of ECCS flow. Next, the current Ktactor variable was calculated based on an assumed function of the fiber quantity on the core inlet and the presence of chemical precipitates. If chemicals E3-26

Enclosure 3 Risk Quantification had already precipitated, then the current Ktactor was set to a very high value. Similarly, no matter the time, if the quantity of fiber on the core inlet was greater than an assumed threshold then the current Ktactor was also set to a very high value. Note that in the context of core fiber accumulation, the chemical precipitation timing refers to the time at which aluminum precipitation occurs (see Section 8.2). For all other conditions, the Ktactor was calculated based on an assumed piecewise function of the current fiber debris load. Using the !<split and Ktactor values, an msplit variable was calculated. If the Ktactor was less than l<split, then msplit was set to 0. If the Ktactor was very high, then the msplit was set to 1. If neither of these conditions were met, then the msplit variable was calculated based on an assumed function of the Ktactor and Ksplit values. Note that an assumed maximum curve fit was used, which results in the minimum msplit value. With these variables calculated, the fraction of debris that was caught on the core inlet and within the core for a hot leg break was calculated using the following equations. Qcold leg Core Inlet (HL Break) = ( 1 - ffisplit ) *------- Qhot leg + Qcold leg Qcold leg Qhot leg lncore (HL Break) = msplit * +------- Qhot leg + Qcold leg Qhot leg + Qcold leg Following switchover to hot leg recirculation, all of the fiber carried with the ECCS flow was modeled as accumulating in the core. For hot leg breaks, an in-vessel failure was recorded if any of the following failure criteria were met:

1. The calculated Ktactor exceeded the specified limit before the specified tb1ock time.

2. The incore fiber load was greater than the specified limit.

3. The reactor vessel fiber load (sum of the incore and core inlet fiber quantities) was greater than the specified limit.

See Enclosure 2, Section 3.n.1 for additional details on the hot leg break failure criteria. 13.0 Sub-Model Integration The following flow diagrams show an overview of the analytical models that were used to determine how water was transported, how the conventional debris was generated and transported, how chemical precipitates formed and transported, how the strainer failure criteria were analyzed, and how the core failure criteria were analyzed. All of these models were addressed holistically in a time-dependent manner. E3-27

Enclosure 3 Risk Quantification

5. Water held-up in the

                                                                         ,-4   RCS or other geometric hold-up volumes

2. ECCS pumps provide .__

                             .....-     water to cool the core

1. LOCA occurs, 7. RWST is drained, injection from the RWSTbegins

                       --                                               _____. 4. Water spills directly to the sump recirculation from the sump begins

3. CSS pumps are used

                             '---I                               >---

to cool containment

6. Water is held-up in

                                                                          .__.      transitory and geometric hold-up volumes Figure 3 Water Transport and Accumulation

3. Debris quantities outside ZOI

{unqualified coatings, latent, miscellaneous)

4. Blowdown 5. \Vashdown

2. ZOI debris quantities

1. Select unique transport to transport from from BADGER break location, size, ~ }----? containment ~

H containment (insulation and and orientation qualified coatings) compartments and sump pool compartments to sump pool

6. Pool fill transport 1

7. Recirculation from sump pool to L'

transport from sump strainers and inactive pool to strainers cavities l

9. Debris accumulation S. Debris penetration on core ~- through strainers Figure 3 Debris Generation and Transport E3-28

Enclosure 3 Risk Quantification

1. Corrosion/

dissolution of metals, ~

3. Precipitate solubility concrete, and debris limit byes II

2. Corrosion/

dissolution of metals, ' 4. Formation of concrete, and debris chemical precipitates in sump pool

5. Recirculation transport from sump <E------

pool to strainers

                                                                                \[/

6. Debris penetration through strainers

                                                                                                   - 7. Debris accumulation on core Figure 3 Chemical Product Formation and Transport E3-29

Enclosure 3 Risk Quantification O.Doe

14. Does quantity No quantity No

9. Strainer debris exceed 13. Core debris exceed accumulation tested accumulation blockage debris limits?

limits? Yes 12. Pass Yes 16. Pass core Strainer criteria criteria

11. Fail Strainer criteria 15. Fail core criteria Figure 3 Comparison of Strainer Head Loss and Core Debris Loads to Failure Criteria Any failures that occurred were binned as strainer failures or core failures, and these results were used to calculate the GSl-191 CFP values as described in more detail below. The NARWHAL software, which was used to integrate all of the sub-models, is described in Section 13.1, and Section 13.2 provides a summary of the assumptions used in the VEGP NARWHAL evaluation.

In order to calculate the CFP values, the following steps were taken:

1. GSl-191 failures were computed for each break as described above. For VEGP, the failures for input into the PRA were computed in the following 12 categories (Strainer A and B correspond to the RHR strainers):

a. Small breaks

i. Core failures E3-30

Enclosure 3 Risk Quantification ii. Strainer A and Strainer B failures (without core failures) iii. Strainer A failures (without core or Strainer B failures) iv. Strainer B failures (without core or Strainer A failures)

b. Medium breaks

i. Core failures ii. Strainer A and B failures (without core failures) iii. Strainer A failures (without core or Strainer B failures) iv. Strainer B failures (without core or Strainer A failures)

c. Large breaks

i. Core failures ii. Strainer A and B failures (without core failures) iii. Strainer A failures (without core or Strainer B failures) iv. Strainer B failures (without core or Strainer A failures)

2. Overall plant-wide LOCA frequencies were allocated to individual welds and break sizes using a top-down LOCA frequency methodology.

a. Plant-wide LOCA frequencies were based on the small, medium, and large break frequencies in the VEGP GSl-191 PRA model with log-linear interpolation for intermediate break sizes.

b. The frequency for a given break size was allocated to individual welds (that can experience a break of that size).

3. The PRA model category for large breaks was broken up into size ranges to more accurately calculate the overall CFP. Smaller breaks within a given size range were assumed to have the same probability as larger breaks within the size range.

4. The CFP for a PRA category (e.g., large breaks) was calculated based on the combined CFP and LOCA frequency weight for each size range.

Figure 3-7 shows an example of how the size ranges were used to calculate the strainer CFP for the case with no random equipment failures. In this example, the overall frequencies result in a probability weight of 82.3 percent in Size Range 1, 15.1 percent in Size Range 2, and 2.6 percent in Size Range 3. The corresponding strainer CFP values are 0.0 percent for Size Range 1, 3.0 percent for Size Range 2, and 27.8 percent for Size Range 3. Therefore, the overall strainer CFP for large* breaks is 1.2 percent as shown below: CFPLarge = 0.823

• 0.000 + 0.151*0.030+0.026
• 0.278 = 0.012 E3-31

_ _ _ _ _J

Enclosure 3 Risk Quantification Size Range Definition 1.E-02 0.9 1.E-03 0.8 Large Large La rge Size Rangel Size Range 2 Size Range 3 1.E-04 0.7

  ~                      P{SR,)=82 .3%                               P(SR2)=15.1%                                    P(SR1 )=2.6%
  ~                      CFP(SR 1)=0.06                          CFP(SR )=3.0                                   CFP(SR )=27 8          0.6 u   1.E-05 c

J CT 0.5 ~

   ~

u. 1.E-06 CFP(Large)=l.2% 0.4 50 1.E-07 0.3 02 l.E-08 0.1 1.E-09 0 0 .5 2 4 6 8 10 12 14 15 16 17 18 19 20 21 22 23 24 25 26 27 27.5 28 28.S 29 29.5 30 30.5 31 Break Size (Inches)

                       -   RHR StrainPr Ostbri'i limit failurpc: tAfl Pump   Av~il.1blel    -   VfGP PRA Mf'<ln lOC'A fteflUPncv Figure 3 Illustration of CFP Calculation for Large Breaks Based on Three Size Ranges 13.1 NARWHAL Softwa re NARWHAL is an object-oriented program that models the connections between important plant components (pumps , strainers, tanks , etc.) based on user-defined inputs . The software performs mass balance calculations that determine the time-dependent quantity of water, debris , and chemical solutes associated with each physical object. Using these time-dependent quantities along with other user-specified conditions , each aspect of GSl-191 can be evaluated in an integrated manner. The software can be used to simulate a single break or many thousands of breaks to evaluate the risk associated with GSl-191 .

At any given time during the simulation , the state of the plant can be defined by a fixed set of parameters (i.e., the on/off states of components , the quantity of debris stored by components , etc.). This collection of parameters is called the state vector. NARWHAL updates the state vector by determining the amount of change in each variable given a change in time. The amount of change in each variable is determined using a series of algorithms called "marching algorithms", which are graph traversal algorithms that maintain consistent flow through the plant system . For example , if a 10,000 gpm pump is fed by a pump that is limited to 1,000 gpm , the algorithms will determine that the high capacity pump can only pump 1,000 gpm . These algorithms are essentially a way for NARWHAL components to communicate information to one another. E3-32

Enclosure 3 Risk Quantification A single NARWHAL simulation relies on a series of marching algorithms (Figure 3-8). The algorithms were designed to handle generic configurations of components, meaning that the user can design arbitrary networks as long as the configuration is valid (i.e., there is a source, active pumps, and a sink). The first algorithm, the activity march, exists solely to determine the health of the network and its components. This algorithm is responsible for detecting valid and invalid paths of flow through the system. For instance, it is necessary to detect the failure of a pump if the strainer that feeds the pump fails. Conversely, this algorithm determines that a strainer will not receive flow if the pumps feeding from it shut down or fail. The second algorithm, the source march, determines that flow through the network and its components is consistent. This is important when, for instance, pump flow rates are a function of other pumps (e.g., piggy-backing). It is also important when considering flow restrictions in the system, such as break size dependent flow rates. The third algorithm, the water march, determines the flow rates and storage balances on all components in the network. This algorithm implements the mass balance equations inherent to the NARWHAL base component. After this algorithm has run, water is moved from the network sources, through all active components, to the network sinks. The fourth and final algorithm, the debris march, uses information generated by the previous algorithms to determine the mass balance of debris and chemicals in the network. This includes determining the release of chemicals, the movement of debris, the capture of debris, and the formation of precipitates. E3-33

Enclosure 3 Risk Quantification Simulation Initialization Reporting Activity March Yes Simulation No Source Termination? March Component Water March Tests Debris March Figure 3 NARWHAL's Basic Procedure In each time step , after the marching algorithms have run and a new state vector has been calculated , a series of tests are run against a number of failure criteria (i .e. , strainer structural margin , component debris limits , strainer submergence, etc.). If a failure occurs , it is noted in the results , but the simulation is allowed to continue running as if the failure had not occurred (e.g ., a pump that fails due to loss of NPSH is allowed to continue running normally for the remainder of the simulation). At the end of a simulation , NARWHAL reports the outcome of the simulation in one of three ways . If a single break simulation is being run , NARWHAL outputs a list of time-dependent vectors representing a number of variables including core debris quantities, strainer debris quantities, component failure states, and component flow rates . If a bulk simulation is being run , NARWHAL will not report time vectors. Instead , it reports summary variables such as failure times , total debris on each strainer at the end of the simulation , maximum head loss for each strainer, and total fiber on the core at switchover to hot leg recirculation. In addition , descriptive break information is reported including the break location and size . In this mode , each time a simulation for a given break is completed , a new record containing the summary information is automatically entered into the results file. If a probability or sensitivity simulation is being run , NARWHAL outputs the summary CFP values for each bulk simulation . Additionally, descriptive information about the variables being modified is entered into the results file . In addition to calculating which breaks pass and fail the GSl-191 acceptance criteria , NARWHAL can also be used to post-process the results to calculate the CFP values based on the approach described in Section 13.0. E3-34

Enclosure 3 Risk Quantification 13.2 NARWHAL Model Assumptions The following assumptions were used for the VEGP NARWHAL model evaluation:

1. It was assumed that breaks less than 2 inches do not result in rapid, full depressurization of the RCS. Therefore, injection by the SI accumulators is not required for these breaks. This assumption is consistent with the minimum water level calculation.

2. The containment sprays were assumed to only be activated for hot leg breaks greater than 15 inches. Note that this includes all partial breaks and DEGBs greater than 15 inches on the hot legs; however, no failures on the cold or intermediate legs are assumed to actuate containment sprays. Although there is some uncertainty in which breaks initiate containment sprays, the relatively high containment pressure required to actuate sprays (21.5 psig) significantly reduces the likelihood that most breaks will exceed the set point and actuate containment sprays. The assumption that only very large hot leg breaks will initiate containment sprays is consistent with the results of best-estimate thermal-hydraulic modeling for a range of potential break sizes on the hot and cold leg piping. This modeling showed that a hot leg DEGB initiated containment sprays, while all other evaluated breaks (including a cold leg DEGB and partial 15-inch breaks on both the hot and cold legs) did not. Assuming that hot leg breaks greater than 15 inches activate containment sprays reasonably represents what was learned in the best-estimate thermal hydraulic modeling.

3. It was assumed that containment sprays would be secured at 24 hours. Although there is some uncertainty in the spray duration, this is a reasonably conservative assumption because sprays are required to operate for at least 2 hours once they are initiated, and running sprays longer than 2 hours would significantly increase the quantity of aluminum released from unsubmerged sources.

4. The RHR flow rate used in the NARWHAL CFP calculation was assumed to be 3,700 gpm. The design flow rate for the RHR pumps is 3,000 gpm. However, it is not expected that the actual plant flow rate would be this low. The use of a higher flow rate is generally conservative in terms of recirculation timing, flashing calculations, and head loss correction. The use of 3,700 gpm for the RHR flow rate is consistent with what was used in the deterministic NPSH calculation and the single train value used in the ECCS system head curve.

5. The minimum sump recirculation pH was assumed to be 7.0. This conservatively bounds the calculated minimum value of 7.12 and is consistent with the minimum acceptable pH documented in the buffer verification calculation. Note that the minimum sump recirculation pH was used to calculate the aluminum solubility limit in the NARWHAL model.

6. The spray pH during injection was assumed to be 5.72 as determined from the conservative maximum pH of the RWST. This maximizes chemical release from unsubmerged sources during injection.

E3-35

Enclosure 3 Risk Quantification

7. It was assumed that the breaks downstream of the first isolation valve do not need to be addressed in this quantification because they are not risk significant. This is a reasonable assumption because there would have to be a coincidental failure of the valve along with the pipe break, which is a low probability event. Additionally, there are no known quantities of localized problematic insulation types or any other factors that are unique to the isolable weld locations that would significantly increase the probability of debris-related failures.

8. The amount of latent debris documented in the debris generation calculation is 60 lbm. This value was conservatively increased to a total of 200 lbm in containment for operating margin.

9. The amount of miscellaneous debris documented in the debris generation calculation is 4 ft2. This value was conservatively increased to a total quantity of 50 ft2 in containment for operating margin.

1O. It was assumed that there is a total quantity of 926.6 ft2 of unsubmerged aluminum metal and 348.4 ft2 of submerged aluminum metal in containment. This includes some margin for future additions of aluminum, as described in the chemical product generation calculation.

11. The total amount of exposed concrete that would be submerged in the pool was assumed to be 10,000 ft2. This is a conservatively large surface area used to maximize the potential for chemical release and is consistent with the quantity used in the chemical product generation calculation. Note that the chemical product generation calculation did not evaluate the chemical effects associated with unsubmerged concrete due to the conservative quantity used for submerged concrete.

12. VEGP requires modifications to the existing RHR strainers to maintain long-term submergence during recirculation for all breaks. Two disks per disk stack will be removed from each of the RHR strainer assemblies to achieve submergence.

13. The containment pressure was assumed to be saturation pressure when the pool temperature is greater than 210.96 degrees F, and 14.396 psia after.the pool temperature has dropped below 210.96 degrees F. The pressure of 14.396 psia was calculated based on the minimum containment pressure of -0.3 psig per VEGP Technical Specification 3.6.4 and an atmospheric pressure of 14.696 psia. The temperature of 210.96 degrees F is the corresponding saturation temperature at 14.396 psia determined from linear interpolation. This is a conservative assumption because it only credits the accident pressure necessary to keep the pool as a liquid.

Note that a small amount of containment accident pressure was credited for degasification and flashing calculations (see Assumption 16).

14. It was assumed that head loss due to chemical precipitates is only applied once the theoretical fiber bed is greater than 0.45 inches. This is supported by the 2009 thin bed head loss test.

15. It was assumed that all random equipment failures evaluated for the different NARWHAL CFP evaluations occurred at the beginning of recirculation. This is a conservative assumption because it results in a quicker switchover to recirculation when compared to failure at the beginning of the event. Additionally, for CS pump E3-36

Enclosure 3 Risk Quantification and/or RHR pump failure cases, it results in more debris accumulation on the remaining active strainers.

16. For the purpose of degasification and flashing calculations, it was assumed that up to 3.5 psi of accident pressure would be available. As described in Assumption 13, no credit was taken for accident pressure in the NPSH available calculations. The basis for this assumption is described in more detail in Enclosure 2 Sections 3.f.14 and 3.g.14. Note that the actual accident pressure credited for degasification is 2.5 psi, and the accident pressure credited for flashing is 2.5 psi plus the pressure head from the top of the strainer to the midpoint of the strainer (approximately 1 psi).

17. It was assumed that the assumed maximum msplit curve fit for core blockage calculations should be used in the NARWHAL CFP evaluation. This is a conservative assumption because it maximizes the quantity of fiber that accumulates on the core inlet. Although the curve fit results in a higher msplit value initially, it results in a lower msplit value for greater resistances across the core inlet. This allows more overall fiber to accumulate at the core inlet for breaks that challenge the core inlet fiber limit.

18. The accumulators were assumed to not inject for any secondary side break. This is a reasonable assumption because secondary side breaks do not result in rapid depressurization of the RCS, which would trigger accumulator injection.

19. Both trains of containment spray were assumed to operate for all secondary side breaks. This provides a reasonable starting point for evaluating the secondary side breaks. To address the uncertainty in this assumption, several different equipment configurations were evaluated, including single CS pump failure and failure of both CS pumps.

20. For a secondary side break, the total flow to the ECCS was assumed to be provided by both charging pumps at a flow rate of 230 gpm/pump. This is a reasonable assumption because the PORV water-relieving capacity is 230 gpm per PORV.

Note that recirculation for secondary side breaks would occur due to overflow of the pressurizer relief tank through the two PORVs. The use of the maximum flow rate through the PORVs is conservative because it maximizes the flow split to the RHR strainers during recirculation. 21.All secondary side breaks were assumed to accumulate fiber in-vessel in the same way as a hot leg break. This is a reasonable assumption because the RCS would be bled through the pressurizer relief tank during a feed and bleed scenario. The pressurizer relief tank is connected downstream of the hot leg.

22. The general inputs used to calculate the CFPs for LOCAs were assumed to be applicable for secondary side breaks. These inputs include the following:

       - Thermal hydraulic inputs (water sources, flow rates, temperature profiles, pH, etc.)
       - Debris inputs (unqualified coatings quantities, debris transport fractions, latent debris, etc.)
       - Chemical precipitate debris
       - Strainer geometry
       - Strainer head loss
       - Strainer failure options E3-37

Enclosure 3 Risk Quantification

       - Strainer penetration equations
       - Core blockage equations This provides a reasonable result set of conditions to evaluate the risk impact of secondary side breaks.

23. It was assumed that a LOCA that occurs during full power operation (i.e., Mode 1) is equivalent or bounding compared to the other operating modes. This is a reasonable assumption because the RCS pressure and temperature (key inputs affecting the ZOI size) would either be approximately the same or significantly lower for Modes 2 through 6. In addition, the flow rate required to cool the core (a key input affecting core blockage) would be reduced significantly for low power or shutdown modes.

14.0 Systematic Risk Assessment As described in Section 4.0, the VEGP GSl-191 PRA model includes the necessary modifications to calculate the risk impact associated with the effects of debris on the strainers and in the core. No new human failure events due to the effects of debris were identified, and no human actions to mitigate the effects of debris were credited in the risk calculation. The only common inputs that were used in both the NARWHAL model and the GSl-191 PRA model are the LOCA frequencies and the equipment configurations. As discussed in Section 6.4, the GSl-191 CFPs were calculated using the same LOCA frequencies that were used in the GSl-191 PRA model. In addition, as discussed in Section 6.3, the high likelihood equipment configurations were explicitly evaluated in NARWHAL to calculate separate sets of CFPs for each high likelihood configuration. These CFPs were entered for the appropriate equipment configurations in the GSl-191 PRA model. Therefore, the VEGP NARWHAL model and GSl-191 PRA model are consistent. 14.1 VEGP NARWHAL.CFP Evaluation The VEGP NARWHAL CFP calculation showed that there were no small or medium break LOCAs that fail for any equipment configurations. Table 3-5 shows the NARWHAL CFP results for large break LOCAs for each equipment configuration. E3-38

Enclosure 3 Risk Quantification Table 3 NARWHAL CFP Results for Large Break LOCAs Equipment Strainer A Strainer A Strainer B Core Configuration and B only only No Equipment Failure 0 0.0118 0 0 RHR Pump B Failure 0 NIA 0.0679 NIA Charging Pump B Failure 0 0.0118 0 0 SI Pump B Failure 0 0.0118 0 0 Train B Failure 0 NIA 0.0736 NIA CS Pump B Failure 0 0.0139 0 0 Both CS Pumps Failure 0 0.0177 0 0 Table 3-6 and Table 3-7 show the CFP results for DEGBs on the feedwater line piping and main steam line piping, respectively. None of the feedwater line breaks produced a sufficient quantity of debris to fail, and main steam line breaks only failed under the equipment configuration where both CS pumps fail due to issues unrelated to debris. Table 3 Feedwater Line DEGB CFP Results Equipment Strainer A Strainer A Strainer B Core Configuration and B only only No Equipment Failure 0 0 0 0 Charging Pump B Failure 0 NIA 0 NIA CS Pump B Failure 0 0 0 0 Both CS Pumps Failure 0 0 0 0 Table 3 Main Steam Line DEGB CFP Results Equipment Strainer A Strainer A Strainer B Core Configuration and B only only No Equipment Failure 0 0 0 0 Charging Pump B Failure 0 NIA 0 NIA CS Pump B Failure 0 0 0 0 Both CS Pumps Failure 0 0.475 0 0 The total baseline risk (from internal events, internal fire, internal flood, and seismic events) for the VEGP Unit 1 PRA model is 4.39x10-5 per reactor-year (ry- 1) for GDF and

1. 73x10- 5 ry- 1 for LERF. The total baseline risk for the VEGP Unit 2 PRA model is 5.05x10- 5 ry- 1 for GDF and 1.90x1 o-6 ry- 1 for LERF. Using the CFP results described above, the change in risk calculated with the VEGP GSl-191 PRA model is shown in Table 3-8.

E3-39

Enclosure 3 Risk Quantification Table 3 VEGP Total Risk Impact due to GSl-191 Failures ACDF ALE RF Case (ry-1) crv-1> Risk increase from GSl-191 failures for high-likelihood 2.32x1Q*B 3.10x10- 11 LOCA configurations Bounding risk increase from GSl-191 failures for 1.41x10-9 4.09x10- 12 unlikely LOCA configurations Risk increase from GSl-191 failures for seismically-1.50x10-9 1.50x10-10 induced LOCAs Risk increase from GSl-191 failures for SSBls 1.39x10-9 8.25x10-11 Total risk increase associated with GSl-191 2.75x10-s 2.68x10-10 These CDF, LERF, flCDF, and flLERF values fall well within the RG 1.174 Region Ill guidelines. Therefore, the effects of debris have very low risk at VEGP. Table 3-9 provides a summary of information on each Class 1 ISi weld inside the first isolation valve. The results shown in this table (specifically the maximum transported fiber and whether a failure was observed at the weld) are based on the single train failure equipment configuration. Table 3 Weld Information List Max Weld Break Weld Transported Degradation Failure Compartment ID Side Identifier Fiber to RHR Mechanism  ? A Strainer (ft3 ) 11201-004-29 hot SG_1-4 639.75 D&C yes 6-RB 11201-001-29 hot SG_1-4 637.70 D&C yes 5-RB 11201-001-29 hot SG_1-4 620.24 D&C yes 3-RB 11201-004-29 hot SG_1-4 605.83 D&C yes 4-RB 11201-003-29 hot SG_2-3 482.44 D&C yes 5-RB 11201-002-29 hot SG_2-3 476.86 D&C yes 5-RB 11201-002-29 hot SG_2-3 473.54 D&C yes 3-RB 11201-003-29 hot SG_2-3 462.00 D&C yes 3-RB 11201-008-31 cold SG_1-4 196.90 D&C yes 1-RB 11201-005-31 cold SG_1-4 194.24 D&C yes 1-RB 11201-008-31 cold SG_1-4 186.47 D&C yes 2-RB E3-40

Enclosure 3 Risk Quantification Max Weld Break Weld Transported Degradation Failure Compartment ID Side Identifier Fiber to RHR Mechanism  ? A Strainer (ft3 ) 11201-005-31 cold SG_1-4 185.94 D&C yes 2-RB 11201-006-31 cold SG_2-3 168.92 D&C yes 1-RB 11201-005-31 cold SG_1-4 168.81 D&C yes 3-RB 11201-007-31 cold SG_2-3 167.20 D&C yes 1-RB 11201-008-31 cold SG_1-4 162.99 D&C yes 3-RB 11201-006-31 cold SG_2-3 158.19 D&C yes 2-RB 11201-007-31 cold SG_2-3 156.75 D&C yes 2-RB 11201-006-31 cold SG_2-3 150.85 D&C yes 3-RB 11201-001-29 hot Reactor Cavity 150.42 D&C yes 1-RB 11201-004-29 hot Reactor Cavity 148.06 D&C yes 1-RB 11201-V6-29 hot Reactor Cavity 148.06 D&C yes 001-W37-RB 11201-V6-29 hot Reactor Cavity 147.50 D&C yes 001-W36-RB 11201-003-29 hot Reactor Cavity 142.41 D&C yes 1-RB 11201-V6-29 hot Reactor Cavity 141.36 D&C yes 001-W33-RB 11201-002-29 hot Reactor Cavity 141.13 D&C yes 1-RB 11201-V6-29 hot Reactor Cavity 139.03 D&C yes 001-W40-RB 11201-005-31 cold SG_1-4 137.98 D&C yes 4-RB 11201-008-31 cold SG_1-4 137.56 D&C yes 4-RB 11201-007-31 cold SG_2-3 136.77 D&C yes 3-RB 11201-006-31 cold SG_2-3 131.69 D&C yes 4-RB 11201-008-31 cold SG_1-4 124.63 D&C yes 8-RB 11201-005-31 cold SG_1-4 124.57 D&C yes 8-RB E3-41

Enclosure 3 Risk Quantification Max Weld Break Weld Transported Degradation Failure Compartment ID Side Identifier Fiber to RHR Mechanism  ? A Strainer (ft 3 ) 11201-005-31 cold SG_1-4 123.83 D&C yes 7-RB 11201-008-31 cold SG_1-4 122.59 D&C yes 7-RB 11201-012-27.5 cold SG_1-4 122.33 D&C yes 1-RB 11201-009-27.5 cold SG_1-4 120.98 D&C yes 1-RB 11201-007-31 cold SG_2-3 114.03 D&C yes 4-RB 11201-007-30.5 cold SG_2-3 103.72 D&C yes 7-RB 11201-006-31 cold SG_2-3 103.70 D&C yes 7-RB 11201-006-31 cold SG_2-3 100.47 D&C yes 8-RB 11201-007-31 cold SG_2-3 100.21 D&C yes 8-RB 11201-011-27.5 cold SG 2-3 98.14 D&C yes 1-RB 11201-010-27.5 cold SG 2-3 97.10 D&C yes 1-RB 11201-053-12.814 hot SG 1-4 54.67 TF, D&C yes 2-RB 11201-053-12.814 hot SG_1-4 44.09 TF,D&C yes 3-RB 11201-009-27.5 cold Reactor Cavity 42.36 D&C yes 8-RB 11201-V6- PWSCC, 27.5 cold Reactor Cavity 41.09 no 001-W35-RB D&C 11201-009-27.5 cold Reactor Cavity 41.09 D&C no 9-RB 11201-012-27.5 cold Reactor Cavity 40.77 D&C no 8-RB 11201-010-27.5 cold Reactor Cavity 39.41 D&C no 6-RB 11201-V6- PWSCC, 27.5 cold Reactor Cavity 39.27 no 001-W38-RB D&C 11201-012-27.5 cold Reactor Cavity 39.27 D&C no 9-RB 11201-010-27.5 cold Reactor Cavity 38.45 D&C no 7-RB 11201-V6- PWSCC, 27.5 cold Reactor Cavity 38.43 no 001-W34-RB D&C E3-42

Enclosure 3 Risk Quantification Max Weld Break Weld Transported Degradation Failure Compartment ID Side Identifier Fiber to RHR Mechanism  ? A Strainer (ft 3 ) 11201-011-27.5 cold Reactor Cavity 37.77 D&C no 7-RB 11201-011-27.5 cold Reactor Cavity 36.70 D&C no 8-RB 11201-V6- PWSCC, 27.5 cold Reactor Cavity 36.70 no 001-W39-RB D&C 11201-053-12.814 hot SG_1-4 36.37 TF,D&C yes 1-RB 11201-004-12.814 hot SG 1-4 35.50 TF,D&C no 2-RB 11201-053-12.814 hot SG_1-4 26.42 TF,D&C no 4-RB 11204-124-8.75 cold SG_1-4 26.20 D&C no 16-RB 11204-124-8.75 cold SG 1-4 25.64 D&C no 17-RB 11204-127-8.75 cold SG_1-4 25.48 D&C no 20-RB 11201-036-10.5 hot SG_1-4 25.01 D&C no 5-RB 11201-036-10.5 hot SG_1-4 24.65 D&C no 6-RB 11204-127-8 cold SG_1-4 24.62 D&C no 21-RB 11201-049-10.5 hot SG_1-4 24.35 D&C no 1-RB 11201-004-10.5 hot SG_1-4 24.32 D&C no 3-RB 11201-036-10.5 hot SG_1-4 24.31 D&C no 4-RB 11204-124-8.75 cold SG_1-4 24.27 D&C no 18-RB 11204-:126-8.75 cold SG_2-3 23.80 D&C no 16-RB 11201-001-10.5 hot SG_1-4 23.44 D&C no 2-RB 11204-126-8.75 cold SG_2-3 23.42 D&C no 17-RB 11204-127-8.75 cold SG_1-4 23.40 D&C no 22-RB 11201-036-10.5 hot SG_1-4 22.82 D&C no 1-RB 11201-009-8.75 cold SG_1-4 22.70 D&C no 6-RB E3-43

Enclosure 3 Risk Quantification Max Weld Break Weld Transported Degradation Failure Compartment ID Side Identifier Fiber to RHR Mechanism  ? A Strainer (ft3 ) 11204-126-8.75 cold SG_2-3 22.27 D&C no 18-RB 11204-125-8.75 cold SG_2-3 21.99 D&C no 16-RB 11201-012-8.75 cold SG 1-4 21.99 D&C no 6-RB 11201-049-10.5 hot SG_1-4 21.53 D&C no 2-RB 11204-125-8.75 cold SG_2-3 21.47 D&C no 17-RB 11201-049-10.5 hot SG_1-4 20.97 D&C no 6-RB 11201-011-8.75 cold SG_2-3 20.39 D&C no 5-RB 11204-125-8.75 cold SG_2-3 20.34 D&C no 18-RB 11201-036-10.5 hot SG_1-4 20.25 D&C no 3-RB 11201-036-10.5 hot SG_1-4 20.18 D&C no 2-RB 11204-021-10.5 hot SG 1-4 20.08 D&C no 27-RB 11201-049-10.5 hot SG_1-4 20.05 D&C no 3-RB 11201-053-11.188 hot SG_1-4 20.00 TF,D&C no 5-RB 11204-021-10.5 hot SG_1-4 19.83 D&C no 28-RB 11201-049-10.5 hot SG_1-4 19.60 D&C no 4-RB 11201-049-10.5 hot SG_1-4 19.36 D&C no 5-RB 11201-010-8.75 cold SG_2-3 18.68 D&C no 4-RB 11201-V6- Pressurizer 11.188 hot 16.54 D&C no 002-W22-RB Compartment 11201-058- Pressurizer 5.189 cold 14.94 D&C no 6-RB Compartment 11201-V6- Pressurizer 5.189 cold 14.93 D&C no 002-W17-RB Compartment 11201-059- Pressurizer 5.189 cold 14.87 D&C no 7-RB Compartment 11201-V6- Pressurizer 5.189 cold 14.78 D&C no 002-W18-RB Compartment E3-44

Enclosure 3 Risk Quantification Max Weld Break Weld Transported Degradation Failure Compartment ID Side Identifier Fiber to RHR Mechanism  ? A Strainer (ft3 ) 11201-059- Pressurizer 5.189 cold 14.75 D&C no 6-RB Compartment 11201-056- Pressurizer 5.189 cold 14.73 D&C no 5-RB Compartment 11201-V6- Pressurizer 5.189 cold 14.73 D&C no 002-W19-RB Compartment 11201-059- Pressurizer 5.189 cold 14.69 D&C no 2-RB Comoartment 11201-V6- Pressurizer 5.189 cold 14.69 D&C no 002-W20-RB Compartment 11201-059- Pressurizer 5.189 cold 14.66 D&C no 5-RB Comoartment 11201-059- Pressurizer 5.189 cold 14.58 D&C no 4-RB Comoartment 11201-057- Pressurizer 5.189 cold 14.56 D&C no 2-RB Comoartment 11201-059- Pressurizer 5.189 cold 14.53 D&C no 8-RB Comoartment 11201-057- Pressurizer 5.189 cold 14.53 D&C no 7-RB Comoartment 11201-058- Pressurizer 5.189 cold 14.48 D&C no 2-RB Compartment 11201-059- Pressurizer 5.189 cold 14.48 D&C no 3-RB Comoartment 11201-056- Pressurizer 5.189 cold 14.41 D&C no 2-RB Compartment 11201-058- Pressurizer 5.189 cold 14.39 D&C no 5-RB Comoartment 11201-058- Pressurizer 5.189 cold 14.33 D&C no 7-RB Comoartment 11201-058- Pressurizer 5.189 cold 14.30 D&C no 3-RB Comoartment 11201-057- Pressurizer 5.189 cold 14.30 D&C no 6-RB Comoartment 11201-057- Pressurizer 5.189 cold 14.28 D&C no 8-RB Compartment 11201-059- Pressurizer 5.189 cold 14.27 D&C no 9-RB Comoartment 11201-057- Pressurizer 5.189 cold 14.21 D&C no 3-RB Comoartment 11201-056- Pressurizer 5.189 cold 14.14 D&C no 4-RB Comoartment 11201-058- Pressurizer 5.189 cold 14.14 D&C no 4-RB Comoartment E3-45

Enclosure 3 Risk Quantification Max Weld Break Weld Transported Degradation Failure Compartment ID Side Identifier Fiber to RHR Mechanism  ? A Strainer (ft3 ) 11201-059- Pressurizer 5.189 cold 14.00 D&C no 10-RB Compartment 11201-056- Pressurizer 5.189 cold 13.99 D&C no 6-RB Compartment 11201-057- Pressurizer 5.189 cold 13.91 D&C no 5-RB Compartment 11201-057- Pressurizer 5.189 cold 13.87 D&C no 4-RB Compartment 11201-012-3.438 cold SG 1-4 13.68 D&C no 4-RB 11201-056- Pressurizer 5.189 cold 13.67 D&C no 3-RB Compartment 11201-030-3.438 cold SG_1-4 13.57 D&C no 1-RB 11201-009-3.438 cold SG_1-4 13.56 D&C no 4-RB 11201-029-3.438 cold SG 1-4 13.52 D&C no 1-RB 11201-V6- Pressurizer 3.438 cold 13.48 D&C no 002-W21-RB Compartment 11201-059- Pressurizer 5.189 cold 12.97 D&C no 11-RB Compartment 11201-059- Pressurizer 5.189 cold 12.84 D&C no 12-RB Compartment 11201-059- Pressurizer 5.189 cold 12.83 D&C no 13-RB Compartment 11201-060- Pressurizer 5.189 cold 12.82 D&C no 1-RB Compartment 11201-060- Pressurizer 5.189 cold 12.82 D&C no 2-RB Compartment 11201-009-2.626 cold SG 1-4 12.80 VF,TF, D&C no 5-RB 11201-011-2.626 cold SG_2-3 12.78 VF,D&C no 4-RB 11208-009-2.626 cold SG_1-4 12.77 VF,TF, D&C no 6-RB 11201-048-2.626 cold SG_2-3 12.73 VF,D&C no 1-RB 11201-012-2.626 cold SG 1-4 12.67 VF,TF, D&C no 5-RB 11201-008-2.626 cold SG_1-4 12.63 D&C no 5-RB 11201-005-2.626 cold SG_1-4 12.63 D&C no 5-RB E3-46

Enclosure 3 Risk Quantification Max Weld Break Weld Transported Degradation Failure Compartment ID Side Identifier Fiber to RHR Mechanism  ? A Strainer (ft 3 ) 11201-006-2.626 cold SG_2-3 12.62 D&C no 5-RB 11201-007-2.626 cold SG_2-3 12.61 D&C no 5-RB 11204-023-5.189 hot SG_1-4 12.54 D&C no 20-RB 11204-023-5.189 hot SG_1-4 12.48 D&C no 21-RB 11201-011-2.626 cold Reactor Cavity 12.48 TF,D&C no 6-RB 11201-012-2.626 cold Reactor Cavity 12.47 TF, D&C no 7-RB 11201-029-3.438 cold SG_1-4 12.45 D&C no 2-RB 11204-246-2.626 cold Reactor Cavity 12.44 TF,D&C no 36-RB 11201-005-2.626 cold SG_1-4 12.44 D&C no 9-RB 11201-006-2.626 cold SG_2-3 12.43 D&C no 9-RB 11201-008-2.626 cold SG_1-4 12.42 D&C no 9-RB 11201-007-2.626 cold SG_2-3 12.41 D&C no 9-RB 11204-245-2.626 cold Reactor Cavity 12.37 TF, D&C no 33-RB 11201-030- Pressurizer 5.189 cold 12.37 D&C no 29-RB Compartment 11201-030- Pressurizer 5.189 cold 12.35 TF,D&C no 30-RB Compartment 11201-030- Pressurizer 5.189 cold 12.32 TF,D&C no 34-RB Compartment 11201-030- Pressurizer 5.189 cold 12.32 TF, D&C no 33-RB Compartment 11201-029-3.438 cold SG_1-4 12.30 D&C no 3-RB 11208-007-2.626 cold SG_1-4 12.29 VF,TF, D&C no 6-RB 11201-029-3 3.438 cold SG 1-4 12.26 D&C no A-RB 11201-030- Pressurizer 5.189 cold 12.24 D&C no 28-RB Compartment 11201-029-3.438 cold SG_1-4 12.23 D&C no 8-RB E3-47

Enclosure 3 Risk Quantification Max Weld Break Weld Transported Degradation Failure Compartment ID Side Identifier Fiber to RHR Mechanism  ? A Strainer (ft3 ) 11201-029-3.438 cold SG_1-4 12.21 D&C no 5-RB 11201-029-3.438 cold SG_1-4 12.17 D&C no 6-RB 11201-029-3.438 cold SG_1-4 12.16 D&C no 9-RB 11201-029-3.438 cold SG_1-4 12.15 D&C no 7-RB 11201-048-2 cold SG_2-3 12.14 VF, D&C no 2-RB 11201-030- Pressurizer 5.189 cold 12.13 TF,D&C no 35-RB Compartment 11201-030- Pressurizer 5.189 cold 12.13 TF,D&C no 32-RB Compartment 11201-030- Pressurizer 5.189 cold 12.11 TF,D&C no 31 A-RB Compartment 11201-011-2.125 cold SG_2-3 12.11 D&C no 3-RB 11201-009-2.125 cold SG_1-4 12.11 D&C no 3-RB 11201-012-2.125 cold SG_1-4 12.11 D&C no 3-RB 11201-010-2.125 cold SG_2-3 12.11 D&C no 3-RB 11201-029-3.438 cold SG_1-4 12.08 D&C no 4-RB 11201-030- Pressurizer 3.438 cold 12.04 D&C no 20-RB Compartment 11201-030- Pressurizer 3.438 cold 12.03 D&C no 19-RB Compartment 11201-048-2.626 cold SG_2-3 12.01 VF, D&C no 3-RB 11201-009-2.626 cold SG_1-4 12.00 TF, D&C no 7-RB 11201-010-2.626 cold SG_2-3 11.97 TF,D&C no 5-RB 11204-243-2.626 cold SG 1-4 11.97 TF,D&C no 34-RB 11201-030- Pressurizer 3.438 cold 11.93 D&C no 21-RB Compartment 11204-244-2.626 cold SG_2-3 11.92 TF, D&C no 28-RB 11204-023-5.189 hot SG_1-4 11.91 D&C no 19-RB E3-48

Enclosure 3 Risk Quantification Max Weld Break Weld Transported Degradation Failure Compartment ID Side Identifier Fiber to RHR Mechanism  ? A Strainer (ft 3 ) 11208-007-2.626 cold SG 1-4 11.89 VF,TF, D&C no 5-RB 11208-007-2.626 cold SG 1-4 11.86 VF,TF, D&C no 4-RB 11201-030- Pressurizer 3.438 cold 11.85 D&C no 18-RB Compartment 11201-030- Pressurizer 3.438 cold 11.83 D&C no 17-RB Compartment 11201-030-3.438 cold SG_1-4 11.83 D&C no 12-RB 11201-029- Pressurizer 3.438 cold 11.82 D&C no 25-RB Compartment 11201-030-2 cold SG_1-4 11.81 D&C no 14-RB 11201-030-3.438 cold SG_1-4 11.81 D&C no 13-RB 11201-030-3.438 cold SG_1-4 11.81 D&C no 11-RB 11208-009-2.626 cold SG 1-4 11.79 VF,TF, D&C no 5-RB 11201-030- Pressurizer 3.438 cold 11.78 D&C no 22-RB Compartment 11201-030-3.438 cold SG_1-4 11.78 D&C no 5-RB 11208-009-2.626 cold SG_1-4 11.77 VF,TF, D&C no 4-RB 11201-009-1.689 cold SG_1-4 11.77 D&C no 2-RB 11201-030-3.438 cold SG_1-4 11.77 D&C no 8-RB 11201-011-1.689 cold SG_2-3 11.77 D&C no 2-RB 11201-012-1.689 cold SG_1-4 11.77 D&C no 2-RB 11201-010-1.689 cold SG_2-3 11.77 D&C no 2-RB 11201-029- Pressurizer 3.438 cold 11.76 D&C no 24-RB Compartment 11201-030-3.438 cold SG_1-4 11.76 D&C no 9-RB 11201-030-3.438 cold SG_1-4 11.76 D&C no 4-RB 11201-007-1.689 cold SG_2-3 11.76 D&C no 6-RB E3-49

Enclosure 3 Risk Quantification Max Weld Break Weld Transported Degradation Failure Compartment ID Side Identifier Fiber to RHR Mechanism  ? A Strainer (ft3 ) 11201-005-1.689 cold SG_1-4 11.76 D&C no 6-RB 11208-012-1.689 cold Annulus 11.76 VF,D&C no 3-RB 11201-006-1.689 cold SG_2-3 11.76 D&C no 6-RB 11201-008-1.689 cold SG_1-4 11.76 D&C no 6-RB 11201-030-2 cold SG_1-4 11.75 D&C no 7-RB 11208-012-5 1.689 cold Annulus 11.75 VF,D&C no 8-RB 11201-031-1.689 cold SG_1-4 11.75 D&C no 1-RB 11201-029- Pressurizer 3.438 cold 11.74 D&C no 22-RB Compartment 11201-029-3.438 cold SG_1-4 11.74 D&C no 18-RB 11201-030-3.438 cold SG_1-4 11.74 D&C no 6-RB 11201-011-1.689 cold SG_2-3 11.74 D&C no 9-RB 11201-010-1.689 cold SG_2-3 11.74 D&C no 8-RB 11201-009-1.689 cold SG_1-4 11.74 D&C no 10-RB 11201-012-1.689 cold SG_1-4 11.74 D&C no 10-RB 11201-029-3.438 cold SG_1-4 11.73 D&C no 19-RB 11201-030-3.438 cold SG_1-4 11.73 D&C no 3-RB 11201-042-1.689 cold SG_2-3 11.73 D&C no 1-RB 11201-051-1.689 cold SG_1-4 11.72 D&C no 1-RB 11201-046- - 1.689 cold SG_2-3 11.72 D&C no 1-RB 11208-012-5 1.689 cold Annulus 11.72 VF,D&C no A-RB 11201-042-1.689 cold SG_2-3 11.72 TF,D&C no 2-RB 11201-051-1.689 cold SG_1-4 11.72 TF,D&C no 2-RB E3-50

Enclosure 3 Risk Quantification Max Weld Break Weld Transported Degradation Failure Compartment ID Side Identifier Fiber to RHR Mechanism  ? A Strainer (ft3 ) 11201-046-1.689 cold SG_2-3 11.72 TF,D&C no 2-RB 11208-012-1.689 cold Annulus 11.72 VF, D&C no 4-RB 11201-031-1.689 cold SG_1-4 11.72 TF,D&C no 2-RB 11208-012-1.689 cold Annulus 11.72 VF,D&C no 5-RB 11201-030-2 cold SG_1-4 11.71 D&C no 2-RB 11208-007-2.626 cold SG_1-4 11.71 VF,TF, D&C no 3-RB 11201-029- Pressurizer 3.438 cold 11.70 D&C no 26-RB Compartment 11201-029-3.438 cold SG_1-4 11.69 D&C no 17-RB 11208-009-2.626 cold SG_1-4 11.68 VF,TF, D&C no 3-RB 11201-046-1.689 cold SG_2-3 11.68 TF, D&C no 3-RB 11201-030- Pressurizer 5.189 cold 11.68 TF, D&C no 38-RB Compartment 11201-051-1.689 cold SG_1-4 11.68 TF, D&C no 3-RB 11201-042-1.689 cold SG_2-3 11.68 TF, D&C no 3-RB 11201-029-3.438 cold SG_1-4 11.68 D&C no 12-RB 11201-031-1.689 cold SG_1-4 ., ...-;-"'~ 11.68 TF, D&C no 3-RB 11201-029-3.438 cold SG_1-4 11.68 D&C no 13-RB 11201-060- Pressurizer 2.626 cold 11.67 D&C no 3-RB Compartment 11201-029-3.438 cold SG_1-4 11.67 D&C no 14-RB 11201-029- Pressurizer 3.438 cold 11.66 D&C no 23-RB Compartment 11201-029-3.438 cold SG_1-4 11.66 D&C no 11-RB 11201-029-3.438 cold SG_1-4 11.65 D&C no 16-RB 11201-048-2.626 cold SG_2-3 11.64 VF,TF, D&C no 5-RB E3-51

Enclosure 3 Risk Quantification Max Weld Break Weld Transported Degradation Failure Compartment ID Side Identifier Fiber to RHR Mechanism  ? A Strainer (ft3 ) 11201-048-2.626 cold SG_2-3 11.64 VF,TF, D&C no 4-RB 11201-060- Pressurizer 2.626 cold 11.64 D&C no 5-RB Compartment 11201-029-3.438 cold SG_1-4 11.64 D&C no 15-RB 11201-030- Pressurizer 3.438 cold 11.64 D&C no 24-RB Compartment 11201-030- Pressurizer 3.438 cold 11.64 D&C no 23-RB Compartment 11201-030- Pressurizer 5.189 cold 11.64 TF, D&C no 37-RB Compartment 11201-059- Pressurizer 2.626 cold 11.64 D&C no 14-RB Compartment 11201-060- Pressurizer 2.626 cold 11.63 D&C no 4-RB Compartment 11201-030- Pressurizer 5.189 cold 11.63 TF, D&C no 36-RB Compartment 11201-003-5.189 hot SG_2-3 11.61 D&C no 2-RB 11201-048-2.626 cold SG_2-3 11.61 VF,TF, D&C no 6-RB 11201-059- Pressurizer 2.626 cold 11.60 D&C no 15-RB Compartment 11201-060- Pressurizer 2.626 cold 11.60 D&C no 6-RB Compartment 11201-030- Pressurizer 3.438 cold 11.60 D&C no 25-RB Compartment 11201-030- Pressurizer 3.438 cold 11.59 D&C no 26-RB Compartment 11201-059- Pressurizer 2.626 cold 11.59 D&C no 16-RB Compartment 11201-030- Pressurizer 3.438 cold 11.59 D&C no 27-RB Compartment 11201-051-1.689 cold SG_1-4 11.56 D&C no 4-RB 11201-048-2.626 cold SG_2-3 11.54 VF,TF, D&C no 7-RB 11201-059- Pressurizer 2.626 cold 11.54 D&C no 17-RB Compartment 11204-246-1.338 cold Reactor Cavity 11.54 TF, D&C no 35-RB 11201-030- Pressurizer 3.438 cold 11.53 TF,D&C no 39-RB Compartment E3-52

Enclosure 3 Risk Quantification Max Weld Break Weld Transported Degradation Failure Compartment ID Side Identifier Fiber to RHR Mechanism  ? A Strainer (ft3 ) 11201-030- Pressurizer 3.438 cold 11.52 D&C no 16-RB Compartment 11201-030- Pressurizer 3.438 cold 11.51 D&C no 15-RB Compartment 11201-048-2.626 cold SG 2-3 11.51 VF,TF, D&C no 10-RB 11201-051-1.689 cold SG_1-4 11.50 D&C no 6-RB 11201-048-2.626 cold SG_2-3 11.50 VF,TF, D&C no 8-RB 11201-048-2.626 cold SG_2-3 11.50 VF,TF, D&C no 9-RB 11201-051-1.689 cold SG_1-4 11.49 D&C no 5-RB 11201-029- Pressurizer 3.438 cold 11.49 D&C no 21-RB Compartment 11204-245-1.338 cold Reactor Cavity 11.49 TF,D&C no 32-RB 11201-029- Pressurizer 3.438 cold 11.49 D&C no 20-RB Compartment 11201-030- Pressurizer 3.438 cold 11.47 TF,D&C no 40-RB Compartment 11201-042-1.689 cold SG 2-3 11.46 D&C no 4-RB 11201-030- Pressurizer 3.438 cold 11.45 TF,D&C no 41-RB Compartment 11201-030- Pressurizer 3.438 cold 11.45 TF,D&C no 44-RB Compartment 11201-030- Pressurizer 3.438 cold 11.45 TF,D&C no 42-RB Compartment 11201-030- Pressurizer 3.438 cold 11.45 TF,D&C no 43-RB Compartment 11201-060- Pressurizer 2.626 cold 11.45 D&C no 7-RB Compartment 11201-030- Pressurizer 1.689 cold 11.44 VF,TF, D&C no 31-RB Compartment 11208-012- Pressurizer 1.689 cold 11.43 VF,D&C no 6-RB Compartment 11201-059- Pressurizer 2.626 cold 11.43 D&C no 18-RB Compartment 11201-042-1.689 cold SG_2-3 11.43 D&C no 5-RB 11201-031-1.689 cold SG_1-4 11.43 D&C no 5-RB E3-53

Enclosure 3 Risk Quantification Max Weld Break Weld Transported Degradation Failure Compartment ID Side Identifier Fiber to RHR Mechanism  ? A Strainer (ft 3 ) 11201-031-1.689 cold SG_1-4 11.42 D&C no 6-RB 11204-246-1.338 cold Reactor Cavity 11.42 TF,D&C no 34-RB 11204-243-1.338 cold SG_1-4 11.42 TF,D&C no 33-RB 11204-245-1.338 cold Reactor Cavity 11.42 TF,D&C no 31-RB 11201-042-1.689 cold SG_2-3 11.42 D&C no 6-RB 11201-046-1.689 cold SG_2-3 11.41 D&C no 5-RB 11201-046-1.689 cold SG_2-3 11.41 D&C no 4-RB 11201-031-1.689 cold SG_1-4 11.41 D&C no 4-RB 11201-030-1.16 cold SG_1-4 11.40 D&C no 10-RB 11201-059- Pressurizer 2.626 cold 11.40 D&C no 19-RB Compartment 11201-029-1.16 cold SG_1-4 11.40 D&C no 10-RB 11204-244-1.338 cold SG 2-3 11.40 TF, D&C no 27-RB 11204-246-1.338 cold Reactor Cavity 11.40 TF, D&C no 33-RB 11201-030-1.16 cold SG_1-4 11.40 D&C no 46-RB 11201-029-1.16 cold SG_1-4 11.40 D&C no 27-RB 11201-060- Pressurizer 2.626 cold 11.40 D&C no 8-RB Compartment 11204-245-1.338 cold Reactor Cavity 11.39 TF,D&C no 30-RB 11201-060- Pressurizer 2.626 cold 11.38 D&C no 10-RB Compartment 11204-243-1.338 cold SG_1-4 11.37 TF, D&C no 32-RB 11201-059- Pressurizer 2.626 cold 11.37 D&C no 20-RB Compartment 11201-060- Pressurizer 2.626 cold 11.37 D&C no 9-RB Compartment 11204-243-1.338 cold SG_1-4 11.36 D&C no 30-RB E3-54

Enclosure 3 Risk Quantification Max Weld Break Weld Transported Degradation Failure Compartment ID Side Identifier Fiber to RHR Mechanism  ? A Strainer (ft3 ) 11204-243-1.338 cold 31-RB SG_1-4 11.36 TF,D&C no 11204-245-1.338 cold 29-RB SG_2-3 11.36 D&C no 11208-045-1.338 cold 14 A-RB SG_2-3 11.35 VF, D&C no 11208-045-1.338 cold 14 B-RB SG 2-3 11.35 VF,D&C no 11208-045-1.338 cold 14-RB SG_2-3 11.35 VF,D&C no 11208-043-5 1.338 cold A-RB SG_2-3 11.35 VF,D&C no 11208-043-5 1.338 cold B-RB SG 2-3 11.35 VF,D&C no 11208-024-5 1.338 cold A-RB SG_1-4 11.35 VF,D&C no 11208-024-5 1.338 cold B-RB SG 1-4 11.35 VF,D&C no 11208-047-1.338 cold 14 A-RB SG 1-4 11.35 VF,D&C no 11208-047-1.338 cold 14 B-RB SG_1-4 11.35 VF,D&C no 11204-244-1.338 cold 26-RB SG_2-3 11.34 TF,D&C no 11208-047-1.338 cold 14-RB SG_1-4 11.34 VF,D&C no 11208-024-1.338 cold 5-RB SG 1-4 11.34 VF,D&C no 11208-043-1.338 cold 5-RB SG_2-3 11.34 VF,D&C no 2.626 cold 11201-059- Pressurizer 11.34 D&C no 21-RB Compartment 11208-045-1.338 cold 13 A-RB SG_2-3 11.33 VF,D&C no 11204-244-1.338 cold 25-RB SG 2-3 11.33 TF,D&C no 11208-045-1.338 cold 9-RB SG_2-3 11.33 VF,D&C no 11208-045-1.338 cold 8-RB SG_2-3 11.33 VF,D&C no 11208-045-9 1.338 cold A-RB SG_2-3 11.33 VF, D&C no 11208-045-1.338 cold 16-RB SG_2-3 11.33 VF, D&C no E3-55

Enclosure 3 Risk Quantification Max Weld Break Weld Transported Degradation Failure Compartment ID Side Identifier Fiber to RHR Mechanism  ? A Strainer (ft 3 ) 11208-045-1.338 cold SG_2-3 11.33 VF,D&C no 12-RB 11208-045-1.338 cold SG_2-3 11.33 VF,D&C no 15-RB 11208-024-1.338 cold SG_1-4 11.33 VF, D&C no 3-RB 11208-045-9 1.338 cold SG 2-3 11.32 VF, D&C no B-RB 11208-045-1.338 cold SG_2-3 11.32 VF,D&C no 13-RB 11208-045-1.338 cold SG_2-3 11.32 VF,D&C no 10-RB 11208-045-1.338 cold SG_2-3 11.32 VF,D&C no 11-RB 11208-024-4 1.338 cold SG_1-4 11.32 VF, D&C no B-RB 11204-246-1.338 cold SG_1-4 11.32 TF, D&C no 32-RB 11208-047-9 1.338 cold SG 1-4 11.32 VF, D&C no C-RB 11208-047-9 1.338 cold SG 1-4 11.32 VF, D&C no B-RB 11208-047-1.338 cold SG 1-4 11.32 VF, D&C no 16-RB 11208-024-4 1.338 cold SG_1-4 11.31 VF, D&C no A-RB 11208-024-1.338 cold SG 1-4 11.31 VF,D&C no 7-RB 11208-043-1.338 cold SG_2-3 11.31 VF,D&C no 6-RB 11208-047-9 1.338 cold SG_1-4 11.31 VF,D&C no A-RB 11208-043-3 1.338 cold SG_2-3 11.31 VF,D&C no A-RB 11208-043-3 1.338 cold SG_2-3 11.31 VF,D&C no B-RB 11208-043-1.338 cold SG_2-3 11.31 VF, D&C no 3-RB 11208-043-3 1.338 cold SG_2-3 11.31 VF, D&C no C-RB 11208-043-3 1.338 cold SG_2-3 11.31 VF,D&C no D-RB 11208-047-5 1.338 cold SG_1-4 11.30 VF, D&C no B-RB E3-56

Enclosure 3 Risk Quantification Max Weld Break Weld Transported Degradation Failure Compartment ID Side Identifier Fiber to RHR Mechanism  ? A Strainer (ft3 ) 11208-047-1.338 cold SG_1-4 11.30 VF, D&C no 15-RB 11208-043-4 1.338 cold SG_2-3 11.30 VF, D&C no A-RB 11208-043-1.338 cold SG_2-3 11.30 VF, D&C no 7-RB 11208-024-3 0.5 cold SG_1-4 11.30 VF, D&C no A-RB 11208-024-3 0.5 cold SG_1-4 11.30 VF, D&C no B-RB 11208-024-0.5 cold SG 1-4 11.30 VF, D&C no 4-RB 11208-024-0.5 cold SG 1-4 11.30 VF, D&C no 6-RB 11208-045-0.5 cold SG 2-3 11.30 VF,D&C no 3-RB 11208-045-0.5 cold SG 2-3 11.30 VF, D&C no 4-RB 11208-045-5 0.5 cold SG_2-3 11.30 VF,D&C no A-RB 11208-045-5 0.5 cold SG_2-3 11.30 VF, D&C no B-RB 11208-045-0.5 cold SG_2-3 11.30 VF, D&C no 5-RB 11208-045-0.5 cold SG_2-3 11.30 VF, D&C no 6-RB 11208-045-0.5 cold SG_2-3 11.30 VF, D&C no 7-RB 11208-047-0.5 cold SG_1-4 11.30 VF, D&C no 3-RB 11208-047-5 1.338 cold SG_1-4 11.30 VF, D&C no A-RB 11208-043-4 1.338 cold SG_2-3 11.30 VF, D&C no B-RB 11204-244-0.5 cold SG 2-3 11.30 TF,D&C no 24-RB 11208-024-4 0.5 cold SG_1-4 11.30 VF, D&C no AA-RB 11208-043-0.5 cold SG_2-3 11.30 VF, D&C no 4-RB 11204-025-5.189 hot SG_2-3 11.28 D&C no 24-RB 11201-002-5.189 hot SG_2-3 11.26 D&C no 2-RB E3-57

Enclosure 3 Risk Quantification Max Weld Break Weld Transported Degradation Failure Compartment ID Side Identifier Fiber to RHR Mechanism  ? A Strainer (ft 3 ) 11204-024-5.189 hot SG_2-3 11.23 D&C no 19-RB 11204-024-5.189 hot SG_2-3 11.20 D&C no 14-RB 11204-021-5.189 hot SG_1-4 11.12 D&C no 18-RB 11204-025-5.189 hot SG_2-3 11.09 D&C no 23-RB 11204-025-5.189 hot SG_2-3 11.06 D&C no 20-RB 11204-021-5.189 hot SG_1-4 10.97 D&C no 19-RB 11204-021-5.189 hot SG_1-4 10.89 D&C no 17-RB 11204-024-5.189 hot SG_2-3 10.84 D&C no 18-RB 11204-025-5.189 hot SG_2-3 10.83 D&C no 22-RB 11204-025-5.189 hot SG_2-3 10.82 D&C no 21-RB 11204-024-5.189 hot SG_2-3 10.67 D&C no 17-RB 11204-021-5.189 hot SG_1-4 10.41 D&C no 26-RB 11204-021-5.189 hot SG_1-4 10.41 D&C no 22-RB 11204-024-5.189 hot SG_2-3 10.41 D&C no 16-RB 11204-024-5.189 hot SG_2-3 10.33 D&C no 15-RB 11204-023-5.189 hot SG_1-4 10.27 D&C no 17-RB 11204-021-5.189 hot SG 1-4 10.18 D&C no 25-RB 11204-021-5.189 hot SG_1-4 10.15 D&C no 20-RB 11204-021-5.189 hot SG_1-4 10.15 D&C no 23-RB 11204-023-5.189 hot SG 1-4 10.03 D&C no 18-RB 11204-021-5.189 hot SG 1-4 10.03 D&C no 24-RB 11204-023-5.189 hot SG_1-4 9.93 D&C no 16-RB E3-58

Enclosure 3 Risk Quantification Max Weld Break Weld Transported Degradation Failure Compartment ID Side Identifier Fiber to RHR Mechanism  ? A Strainer (ft3 ) 11204-021-5.189 hot SG_1-4 9.84 D&C no 21-RB 11201-003-2.125 hot SG_2-3 9.15 D&C no 4-RB 11201-004-2.125 hot SG_1-4 9.15 D&C no 5-:RB 11201-002-2.125 hot SG_2-3 9.14 D&C no 4-RB 11201-001-2.125 hot SG_1-4 9.14 D&C no 4-RB 11204-025-4 hot SG_2-3 8.48 D&C no 25-RB 14.2 NARWHAL Uncertainty and Sensitivity The purpose of this section is to describe the sensitivity analysis and uncertainty quantification associated with the GSl-191 phenomenological models evaluated using NARWHAL. 14.2.1 Simplified Risk Estimation Methodology For the purposes of sensitivity analysis and uncertainty quantification, a simplified method was used to estimate LiCDF in NARWHAL. The simplification includes a reduction in the number of equipment configurations explicitly evaluated for each sensitivity, and also directly calculates LiCDF in NARWHAL using the LOCA frequencies and the equipment functional failure probabilities (along with the NARWHAL calculated CFP results). The GSl-191 risk can be reasonably estimated for the purpose of sensitivity analysis without explicitly modeling all six of the configurations listed in Section 6.3. The scenario with a combined failure of one RHR pump and two CS pumps has a very low probability resulting in a negligible impact on the risk quantification. Although it is not necessarily bounded by any of the other cases, it was assumed to have the same conditional failure probability as the scenario with a combined failure of one RHR pump and one CS pump. The scenario with one RHR pump failure is bounded by the scenario with a combined failure of one RHR pump and one CS pump. In addition, the scenario with one CS pump failure is bounded by the scenario with two CS pump failures. Therefore, the functional failure probabilities listed in Section 6.3 were combined for the purpose of sensitivity analysis and uncertainty quantification as shown in Table 3-10. E3-59

Enclosure 3 Risk Quantification Table 3 Combined Functional Failure Probabilities Functional Failure Equipment Configuration Probability No Equipment Failures 91.50% 2 CS Pump Failures 6.57% 1 RHR Pump+ 1 CS Pump Failures 1.92% Total 100% For the VEGP NARWHAL model, the CFPs were reported for each PRA size category and success criterion. To estimate the LlCDF, the CFPs for each PRA success criterion were summed within each PRA size category. Therefore, a single CFP value was calculated for each PRA size category. The following equation was then used to estimate LlCDF. i=N j=X

 .1CDF =   II i=O j=O IEFi

• CFPij
• FF11 Nomenclature:

                = Each PRA size category j       = Each equipment configuration IEF     = Initiating event frequency for each PRA size category CFP     =Conditional failure probability for each PRA size category and each equipment configuration FFP     =Functional failure probability for each equipment configuration Note that initiating event frequency of each PRA size category was defined by the LOCA frequencies from Section 6.4.

If there are no medium or small breaks that fail, the equation can be simplified as shown below: j=X

 .1CDF = IEF    *I

• j=O CF11 FF11 Nomenclature:

j = Each equipment configuration IEF = Initiating event frequency for large LOCAs CFP =Large LOCA conditional failure probability for each equipment configuration FFP = Functional failure probability for each equipment configuration E3-60

Enclosure 3 Risk Quantification Given a LOCA frequency for large breaks of 1.85x1 o-6 yr 1 (see Section 6.4), the functional failure probabilities shown in Table 3-10, and the conditional failure probabilities shown in Table 3-5, LiCDF can be estimated as shown below: LlCDF = 1.85 x 10- 6 * (0.9150

• 0.0118 + 0.0657
• 0.0177 + 0.0192
• 0.0736) = 2.47 x 10-s This value is nearly identical to the LiCDF value that was calculated using the GSl-191 PRA model (2.46x1 o-8 yr 1 excluding the SSBI contribution in Table 3-8). Therefore, this method provides an efficient and accurate LiCDF estimate.

Because the base CDF and LERF values at VEGP are well within the RG 1.174 acceptance guidelines (Reference 1), and LiLERF is more than two orders of magnitude lower than LiCDF (see Table 3-8), the risk sensitivity was evaluated by comparing LiCDF to the LiCDF acceptance guideline. As shown in Enclosure 1, Section 2.2, the risk associated with GSl-191 is considered to be small for a mean LiCDF below 1x10-5 yr 1 and very small for a mean LiCDF below 1 x1 o-6 yr 1 . 14.2.2 Sensitivity Analysis Parametric sensitivity analysis was performed to identify which inputs have the greatest impact on the risk quantification results. The parametric sensitivity analysis includes the process of identifying input variables to evaluate, selecting minimum, nominal, and maximum values for each variable, quantifying risk in terms of LiCDF as a common output that can be compared for each sensitivity, and using the LiCDF results to rank the sensitivity of each input variable. The VEGP NARWHAL model includes numerous inputs that could have been included in the sensitivity analysis. However, some of these input parameters are directly correlated to other parameters (and therefore should not be independently analyzed), some parameters were pre-screened as having an insignificant effect on the results, and some parameters do not require an independent analysis because they would have the same type of effect as other similar parameters that are evaluated. A consistent methodology was used to determine the minimum, nominal, and maximum values for each of the parametric sensitivity inputs. Consistency is important because using a very large range for one parameter and a very small range for another parameter may mask the true sensitivity of the second parameter and indicate that the first parameter has a much greater effect on the results. However, selecting consistent minimum and maximum values is challenging due to practical considerations. For example, debris transport fractions can vary between 0 percent and 100 percent, the initial RWST level may vary between the technical specification minimum limit and the high level alarm, and debris head loss may vary from 0 ft at the low end to an unknown value at the high end. In addition, some parameters are not fixed values and may be determined as a function of time (e.g., pool temperature) or as a correlation based on other calculated parameters (e.g., penetration fraction). The following methodology E3-61

Enclosure 3 Risk Quantification provides an approach for evaluating the various input parameters in a consistent manner.

• The nominal value was defined as the input value used in the NARWHAL base case. As discussed in Section 0, the base case NARWHAL model that was used for sensitivity analysis and uncertainty quantification was equivalent to the model from the NARWHAL CFP calculation with the exception of a smaller break database.
• The minimum and maximum values for each sensitivity input depend on the nominal value and the available information. If the nominal value was conservatively skewed toward the minimum direction, the minimum value used for the parametric sensitivity was 10 percent lower than the nominal value. Similarly, if the nominal value was conservatively skewed toward the maximum direction, the maximum value used for the parametric sensitivity was 10 percent higher than the nominal value.
• For all other cases, the minimum and maximum values were determined by the available information. Design limits were used preferentially if they were available. If a range of values was determined analytically, the minimum or maximum from the range was used if design limits were not available.
• If no information was available for the range of a given input, then the minimum or maximum value was assumed to be +/- 25 percent of the nominal value.

The results of the parametric sensitivity analysis were used to rank each input parameter. This was done using a tornado diagram, which illustrates how sensitive the chosen output metric (LiCDF) is to changing an input variable's value from nominal to maximum (or minimum). The tornado diagram was created by first running NARWHAL with all inputs set at nominal conditions, and recording the output metric. One variable was then changed to its maximum value (with all others held constant), the software was re-run, the output metric was recorded, and the results were compared to the nominal case. This process was repeated with each variable being independently modified to the maximum and minimum values. The output responses were then sorted by magnitude and shown from highest output response (most risk-sensitive parameter) to lowest output response (least risk-sensitive parameter). The minimum and maximum values used in the sensitivity analysis are shown in Table 3-11. The LiCDF results are shown in Table 3-12, and the difference in LiCDF (compared to the NARWHAL base case value of 2.46E-08 yr 1 ) was plotted in the tornado diagram shown in Figure 3-9. Note that based on the methodology described above, some of the parametric sensitivity cases consider input values that are outside the plant operating conditions (e.g., the RWST volume used for the NARWHAL base case corresponds to the technical specification minimum level, so the minimum volume used for the sensitivity is less than the technical specification minimum). Therefore, these sensitivity results are only intended to provide insights into the relative importance of each input parameter. E3-62 __J

Enclosure 3 Risk Quantification Table 3 Maximum and Minimum Parametric Sensitivity Inputs Input Parameter Units Minimum Maximum Input Input Simulation Time minutes 32,400 54,000 Initial RWST Level lbm 5,062,577 6,025,079 RHR Pump Flow Rate gpm 2,775 4,500 CS Pump Flow Rate gpm 1,950 3,374 75% *Design 110%

• Pressure and Temperature Profiles Basis Design Basis Sump pH 7.1 8.1 75%
• 110%
• ZOI Debris Quantity Base Case Base Case Latent Debris Quantity Particulate lbm 51 187 Fiber ft 3 3.75 13.75 2

Miscellaneous Debris Quantity ft 2 55 Submerged Aluminum Surface Area ft2 278.7 383.2 Unsubmerged Aluminum Surface Area ft2 741.3 1,019.3 Debris Head Loss Conventional for Fiber~ 3.1 ft3 ft of HzO 0.47 0.78 Conventional for Fiber> 3.1 ft3 ft of HzO 3.50 6.83 Calcium Phosphate ft of HzO 0.83 2.25 SAS ft of HzO 3.24 6.55 Strainer Debris Limits Fiber ft3 9.927 13.79 Particulate lbm 327.348 454.65 Fire Barrier lbm 26.24 36.45 Calcium Phosphate lbm 4.77 6.63 SAS lbm 8.046 11.18 Containment Accident Pressure psi 3.15 4.375 75%

• 125%
• Strainer Penetration Fractions Correlation Correlation Results Results Containment Spray Duration minutes 120 43,200 Reactor Vessel Hot Leg Break Fine Fiber Limit g/FA 50 125 Reactor Vessel Cold Leg Break Fine Fiber Limit g/FA 11.25 18.75 Geometric LOCA Frequency Values 5th Percentile 95th Percentile E3-63

Enclosure 3 Risk Quantification Table 3 Results of Parametric Sensitivity Analysis

                                                                   .dCDF at Input Parameter                     .dCDF at       Maximum Minimum Input        Input Simulation Time                                     2.46E-08        2.47E-08 Initial RWST Level                                  1.18E-08        2.47E-08 RHR Pump Flow Rate                                  2.21 E-08       5.75E-08 CS Pump Flow Rate                                   2.73E-08        2.25E-08 Pressure and Temperature Profiles                   1.09E-08        2.46E-08 Sump pH                                             1.08E-08        2.56E-08 ZOI Debris Quantity                                 8.54E-09        3.33E-08 Latent Debris Quantity                              2.30E-08        2.48E-08 Miscellaneous Debris Quantity                       2.22E-08        2.48E-08 Submerged Aluminum Surface Area                     2.47E-08        2.47E-08 Unsubmerged Aluminum Surface Area                   2.47E-08        2.47E-08 Debris Head Loss                                    2.47E-08        2.60E-08 Strainer Debris Limits                              3.47E-07        5.32E-08 Containment Accident Pressure                       2.47E-08        2.47E-08 Strainer Penetration Fractions                      2.50E-08        2.98E-08 Containment Spray Duration                          2.85E-08        2.47E-08 Reactor Vessel Hot Leg Break Fine Fiber Limit       1.09E-07        2.47E-08 Reactor Vessel Cold Leg Break Fine Fiber Limit      2.47E-08        2.47E-08 Geometric LOCA Frequency Values                     4.16E-11        6.25E-08 E3-64

Enclosure 3 Risk Quantification Change in NARWHAL Base Case ACDF l! Oll!I OO!ill f.00 _ _ _ _ _ _ _ _ _ _ U1E-M Ructor V*s~I Hot ltl B1ok FiM fiotr Limit (8C: 100 a/F~. M<> 5'l I/FA [Co<e lnltt UrMj. \4ox: *l5 l Georittnc LOO: Frtquerv:y v.iues (l!C* PAA Modtl, Mon: NU~fG-lB19 5th. Mn: NUREG-JB29 95th) 3.791-08

                                                                                         *l.<6E-08 * - - - - - - - - - ** - - - - - - - - - -

R..,R Pwnp flow lhite (BC: 3.700 IPm, M*'l:-25 Milt *,IOOIJltn (R*"""utp 3.151-08

                                                                                                         *2551-09 * - - - - - - - .- - - - - - - - - -

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• MuWN Figure 3 Tornado Diagram Showing Risk Sensitivity Ranking The parameters that have the most significant effect on LiCDF are the in-vessel fiber limit for hot leg breaks, the LOCA frequency values , and pump flow rates . The inputs that have no effect on the LiCDF (within the range analyzed) are the containment accident pressure used for degasification and flashing , the in-vessel fiber limit for cold leg breaks , and the submerged and unsubmerged aluminum surface areas.

Reducing the total fine fiber limit in the reactor vessel for hot leg breaks affects the number of core failures . Note that for the NARWHAL base case, no core fiber limit failures were observed. However, when reducing the total fiber limit to 50 g/FA, some failures were observed . Adjusting the LOCA frequencies to the 5th or 95th percentiles affects the results in two ways-first, it is an input for calculating the CFP values as described in Section 13.0, and second , it is a direct input for calculating LiCDF as described in Section 0. The effect on the CFP values is relatively minor and is driven by the change in frequency as E3-65

Enclosure 3 Risk Quantification a function of break size. The direct effect on b.CDF is driven by the magnitude of the LOCA frequency, which has a significant effect because the 5th and 95th percentiles are in some case orders of magnitude different from the mean LOCA frequencies. The flow rates of the RHR and CS pumps also affect the b.CDF. Generally, a higher RHR flow rate and a lower CS flow rate results in more fiber debris deposited on the RHR strainers. A lower RHR flow rate and higher CS flow rate results in more fiber being deposited on the CS strainers. As previously mentioned, the debris limit failure criterion is the most significant contributor to failure. Therefore, altering inputs that affect debris deposition on the strainer results in a change in the risk results. The strainer penetration fraction resulted in a higher b.CDF for the minimum input and the maximum input when compared to the base case. The maximum penetration fractions increased b.CDF because it resulted in an increased quantity of fine fiber deposition in the reactor vessel. This led to core failures, which were not seen in the NARWHAL base case. The minimum penetration fractions increased b.CDF because it resulted in an increased quantity of fine fiber deposited on the strainers. At each time step, less of the fine fiber that arrives penetrated the strainer, and less fine fiber that was previously on the strainer was shed. Although this does not have a large effect, it led to a few more fiber debris limit failures than were seen in the NARWHAL base case. Additionally, the containment temperature and pressure profiles parameter resulted in a lower b.CDF for both the minimum and maximum values. The minimum temperature and pressure profiles result in a lower b.CDF because the quantity of chemical precipitates that form is reduced, which leads to less debris limit failures. The maximum temperature and pressure profiles result in a negligibly lower b.CDF than when compared to the base case. This is because the higher temperature results in a larger volume of water in the containment pool, which affects the rate at which debris accumulates on the strainer. For a handful of breaks, this variation in the rate of debris accumulation resulted in success instead of failure due to the debris limit failure criterion. 14.2.3 Uncertainty Quantification As described in Enclosure 1, Section 5.0, uncertainty quantification includes parametric uncertainty, model uncertainty, and completeness uncertainty. The parametric and model uncertainties were quantified by running NARWHAL sensitivity cases. Note that the parametric uncertainty evaluation has a different purpose than the parametric sensitivity analysis described in Section 14.2.2. The purpose of the parametric sensitivity analysis was to determine the effect of one-at-a-time changes in various input parameters to understand the independent effect of each parameter on the results. In many cases, the parameter changes went outside the bounds of the realistic plant-specific conditions. However, the purpose of the parametric uncertainty quantification was to quantify the overall uncertainties associated with the input E3-66

Enclosure 3 Risk Quantification parameters. Therefore, the effect of simultaneous variations in multiple input parameters was considered, but none of the inputs were shifted beyond the bounds of realistic plant-specific conditions. The parametric uncertainties were quantified using a series of sensitivities with a bounding set of input parameters with respect to a) strainer failures and b) core failures. In cases where the bounding direction for a given input parameter (e.g., pool volume/level) could not be determined, both the minimum and maximum values were run. Table 3-13 shows the worst case conditions for strainer failures, and Table 3-14 shows the worst case conditions for core failures. Table 3-13 -Worst Case Conditions for Strainer Failure Bounding NARWHAL Base Sensitivity Case Parameter Direction Case Input Input Fiber Insulation Debris Consensus Same as NARWHAL Maximum Quantity (Maximum) Base Case Qualified Coatings Consensus Same as NARWHAL Maximum Debris Quantity (Maximum) Base Case Fire Barrier Debris Consensus Same as NARWHAL Maximum Quantity (Maximum) Base Case Unqualified Coatings Consensus Same as NARWHAL Maximum Debris Quantity (Maximum) Base Case Consensus Same as NARWHAL Latent Debris Quantity Maximum (Maximum) Base Case Miscellaneous Debris Consensus Same as NARWHAL Maximum Quantity (Maximum) Base Case Debris Transport Consensus Same as NARWHAL Maximum Fractions (Maximum) Base Case Minimum or Minimum and Pool Volume/Level Minimum Maximum Maximum Consensus Same as NARWHAL Containment Pressure Minimum (Minimum) Base Case Same as NARWHAL Base Case Minimum or Design Basis (Maximum is Pool Temperature Maximum (Maximum) Conservative Based on Parametric Sensitivity Results) Design (based on comparison of the ECCS Flow Rate Maximum Maximum pump curve and system resistance) CS Flow Rate (assuming Minimum Design Minimum sprays initiate) E3-67 J

Enclosure 3 Risk Quantification Bounding NARWHAL Base Sensitivity Case Parameter Direction Case Input Input Function of Water Function of Water ECCS/CS Switchover Minimum Volume and Flow Volume and Flow Time Rates Rates Hot Leg Switchover Procedural Step Same as NARWHAL N/A Time (Minimum) Base Case Minimum or Minimum and Secure CS Time Midpoint Maximum Maximum Consensus Same as NARWHAL Boil-off Flow Rate N/A (Maximum) Base* Case Maximum for Minimum or Same as NARWHAL pH release, Minimum for Maximum Base Case solubilitv Consensus Same as NARWHAL Head Loss Maximum (Maximum) Base Case Same as NARWHAL Structural Margin Minimum Design (Minimum) Base Case Consensus Same as NARWHAL NPSH Margin Minimum (Minimum) Base Case Pump Void Fraction Consensus Same as NARWHAL Minimum Limit (Minimum) Base Case Consensus Minimum (no Penetration Minimum (Maximum) penetration) Consensus Same as NARWHAL Core Fiber Limit N/A (Minimum) Base Case Maximum (95th LOCA Frequency Maximum Nominal (mean) percentile) Table 3-14 -Worst Case Conditions for Core Failure Bounding NARWHAL Base Sensitivity Case Parameter Direction Case Input Input Fiber Insulation Debris Consensus Same as NARWHAL Maximum Quantity (Maximum) Base Case Qualified Coatings Consensus Same as NARWHAL N/A Debris Quantity (Maximum) Base Case Fire Barrier Debris Consensus Same as NARWHAL N/A Quantity (Maximum) Base Case Unqualified Coatings Consensus Same as NARWHAL N/A Debris Quantity (Maximum) Base Case Consensus Same as NARWHAL Latent Debris Quantity Maximum (Maximum) Base Case Miscellaneous Debris Consensus Same as NARWHAL N/A Quantity (Maximum) Base Case E3-68

Enclosure 3 Risk Quantification Bounding NARWHAL Base Sensitivity Case Parameter Direction Case Input Input Debris Transport Consensus Same as NARWHAL Maximum Fractions (Maximum) Base Case Minimum or Minimum and Pool Volume/Level Minimum Maximum Maximum Consensus Same as NARWHAL Containment Pressure N/A (Minimum) Base Case Same as NARWHAL Base Case Minimum or Design Basis (Maximum is Pool Temperature Maximum (Maximum) Conservative Based on Parametric Sensitivity Results) Design (based on Minimum or comparison of the Minimum and ECCS Flow Rate Maximum pump curve and Maximum s*vstem resistance) CS Flow Rate (assuming Minimum Design Minimum sprays initiate) Function of Water Function of Water ECCS/CS Switchover Minimum Volume and Flow Volume and Flow Time Rates Rates Hot Leg Switchover Procedural Step Maximum Maximum Time (Minimum) Secure CS Time Minimum Midpoint Minimum Consensus Same as NARWHAL Boil-off Flow Rate Maximum (Maximum) Base Case Maximum for Minimum or Same as NARWHAL pH release, Minimum Maximum Base Case for solubility Consensus Same as NARWHAL Head Loss N/A (Maximum) Base Case Same as NARWHAL Structural Margin N/A Design (Minimum) Base Case Consensus Same as NARWHAL NPSH Margin N/A (Minimum) Base Case Pump Void Fraction Consensus Same as NARWHAL N/A Limit (Minimum) Base Case Consensus Same as NARWHAL Penetration Maximum (Maximum) Base Case Consensus Same as NARWHAL Core Fiber Limit Minimum (Minimum) Base Case E3-69

Enclosure 3 Risk Quantification Bounding NARWHAL Base Sensitivity Case Parameter Direction Case Input Input Maximum (951h LOCA Frequency Maximum Nominal (mean) percentile) Because minimum and maximum inputs were considered for both the pool volume and CS duration inputs, a 2x2 matrix of simulations was required for the bounding strainer failure cases. Similarly, because minimum and maximum inputs were considered for both the pool volume and RHR flow rate inputs, a separate 2x2 matrix of simulations was required for the bounding core failure cases. The difference in the bounding strainer and core failure cases compared to the NARWHAL base case are summarized below:

1. Strainer Failure Cases (2x2 Matrix)

a. Water Volume

i. Minimum (NARWHAL Base Case Inputs) ii. Maximum (Approximately 500,000 lbm Additional Water)

b. Maximum RHR Flow Rate (4,500 gpm)

c. Minimum CS Flow Rate (1,950 gpm)

d. CS Duration

i. Minimum (120 minutes) ii. Maximum (43,200 minutes)

e. Minimum Penetration (0 percent)

f. Maximum LOCA Frequency (95 1h Percentile)

2. Core Failure Cases (2x2 Matrix)

a. Water Volume

i. Minimum (NARWHAL Base Case Inputs) ii. Maximum (Approximately 500,000 lbm Additional Water)

b. RHR Flow Rate

i. Minimum (2,775 gpm) ii. Maximum (4,500 gpm)

c. Minimum CS Flow Rate (1,950 gpm)

d. Maximum hot leg switchover (HLSO) Time (563 minutes)

e. Minimum CS Duration (120 minutes)

f. Maximum LOCA Frequency (95 1h Percentile)

Table 3-15 shows the LiCDF for each of the parametric uncertainty cases. Figure 3-10 illustrates the change in LiCDF for each of the parametric uncertainty cases in comparison to the NARWHAL base case. E3-70

Enclosure 3 Risk Quantification

                    - - Resu Itsof Paramet"r1c uneertamty Ta bl e 3 15                                . t Q uan ff 1 1caf ion Change in Sensitivity                                                            ACDF from Description                 ACDF Case                                                                NARWHAL Base Case Strainer Case  1   Min water volume and min CS duration     1.22E-07        9.77E-08 Strainer Case  2   Min water volume and max CS duration     1.22E-07        9.77E-08 Strainer Case  3   Max water volume and min CS duration     1.19E-07        9.43E-08 Strainer Case  4   Max water volume and max CS duration     1.19E-07        9.42E-08 Min water volume and min RHR flow Core Case 1                                                 7.11 E-08       4.65E-08 rate Min water volume and max RHR flow Core Case 2                                                 1.16E-07        9.13E-08 rate Max water volume and min RHR flow Core Case 3                                                 7.23E-08        4.76E-08 rate Max water volume and max RHR flow Core Case 4                                                 1.12E-07        8.77E-08 rate I

E3-71

Enclosure 3 Risk Quantification Parametric Uncertainty sensitivity Results

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SensltivltyCase Figure 3-1 O - Comparison of Parametric Uncertainty Sensitivity Cases to the NARWHAL Base Case The process described above evaluates the parametric uncertainty in a very conservative manner by analyzing the worst-case combinations of input values. Although the scenario is hypothetically possible , the probability of all of the worst-case conditions occurring simultaneously is extremely unlikely. The overall results of this evaluation show that the parametric uncertainty is low. To meet the guidance in NUREG-1855 (Reference 11 ), model uncertainty must be addressed for any models or approaches for which no consensus exists . As discussed in Enclosure 1, Section 5.0, most of the GSl-191 models used for the VEGP evaluation are consensus models that have been widely used by the industry and accepted by the NRC . However, the following models used for VEGP are not consensus models and therefore were included in the model uncertainty quantification evaluation :

• Break model
• LOCA frequencies E3-72

Enclosure 3 Risk Quantification

• LOCA frequency allocation to individual welds
• CS actuation
• Aluminum metal release equation
• Fiber bed thickness required for chemical head loss
• LBLOCA size range discretization To address the uncertainty in these models, alternative models were evaluated as shown in Table 3-16.

T a bl e 3 16 - Alterna f 1ve M odes . t 1ry M o d e I U nee rtamty I Use d t 0 Q uan ff Model NARWHAL Base Case Sensitivity Case(s) Break model Continuum break model DEGB-only model VEGP PRA frequencies NUREG-1829 arithmetic LOCA frequencies (derived from NUREG-mean frequencies 1829 geometric mean) Hybrid allocation with multiple options (based on weld LOCA frequency allocation Top-down allocation degradation mechanism probability weighting) Multiple options including no Hot leg breaks larger than CS actuation breaks and all breaks larger 15inches than 2 inches WCAP-16530 release Aluminum metal release UNM release equation equation Fiber thickness required for 0.45 inches 0 inches chemical head loss Multiple options with a biased LBLOCA size range (6-15, 15-25, and 25-43.84 allocation of frequencies to discretization inches) smaller break sizes and larger break sizes E3-73

Enclosure 3 Risk Quantification Table 3-17 shows the ~CDF for each of the model uncertainty cases. Figure 3-11 illustrates the change in ~CDF for each of the model uncertainty cases in comparison to the NARWHAL base case.

                         - - Resu Itsof Mo deI Unee rtamcy T a bl e 3 17                           . t Q uan ff 1 1caf ion Change in Model with No                                                          .dCDF from Sensitivity Case           .dCDF Consensus                                                            NARWHAL Base Case Continuum Break Model       DEGB-Only Model                   8.10E-08        5.63E-08 Top-Down LOCA Top-Down LOCA Frequency Frequency Allocation Allocation with NUREG-1829        5.28E-07        5.04E-07 with Values from the Arithmetic Mean Values VEGP PRA Hybrid Allocation Methodology:

Skewed to High Rupture 4.90E-11 -2.46E-08 Probability Welds Methodology to Allocate Hybrid Allocation Methodology: LOCA Frequency to Skewed to High and Medium 3.55E-09 -2.11 E-08 Welds Rupture Probability Welds Hybrid Allocation Methodology: Spread Equally Across all Welds 2.47E-08 O.OOE+OO (top-down) All Breaks >15" 2.42E-08 -4.46E-10 Breaks Activating All Breaks >6" 2.39E-08 -7.61E-10 Containment Sprays All Breaks >2" 2.39E-08 -7.61E-10 No Breaks 2.70E-08 2.39E-09 UNM Aluminum Metal WCAP-16530 Equation 2.57E-08 9.98E-10 Release Equation 0.45-inch Fiber 0-inch Fiber Thickness Required Thickness Required for 6.74E-08 4.27E-08 for Chemical Head Loss Chemical Head Loss Bias 1 (6-10, 10-15, and 15-LBLOCA Size Range 5.13E-08 2.66E-08 43.84 inches) Discretization (6-15, 15-Bias 2 (6-20, 20-27, and 27-25, and 25-43.84 inches) 2.33E-08 -1.36E-09 43.84 inches) E3-74

Enclosure 3 Risk Quantification Model Uncertainty Sensitivity Results 100{ 06 I 00£. 07

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Sltnwt'd to ~d to Sptf!<id Al Sttih All 8ru*~ >6~ Al)81f!.D.s >Z~ ,.., 8'"Nh WUJl.* 16S30 Otn<t~ &t.b 1 Bi.u l At1th~llt H.ghl\""luflf H'istund cqu.iltyAnon >15" Mf!~n Prob;ab ty MtdH.lm ,a!I Wtkb W('lck llluptllflf P'fot>> ty Wtlth 9.re.;I( Model lOCA LOCA frequency Allouhon toWlkts Aluminum fiber LBUXA. Sil!' lt.ange Met.ii ~l'W.Ut thkkneu l>'~Jf!liZilttOO Rtquwtdfor Chf!miul lfff'Ct~ Hf!.ld Sen.sltlvltyC.11e "'" Figure 3 Comparison of Model Uncertainty Sensitivity Cases to the NARWHAL Base Case The results of this evaluation show that the model uncertainty is low (i.e ., the resulting b.CDF for each of the model uncertainty cases is within Region 3 as defined in RG 1.174). Because all of the cases that were evaluated for model uncertainty and parametric uncertainty resulted in a b.CDF less than 1x1 o-6 , it can be concluded with high confidence that the risk associated with GSl-191 is very low as defined by the acceptance guidelines in RG 1.174 (Reference 1). 14.3 PRA Model Uncertainty and Sensitivity The purpose of this section is to address the impact of PRA modeling epistemic uncertainty on the GSl-191 risk assessment. The baseline internal events and seismic PRA models document assumptions and sources of uncertainty, and these have been reviewed during the model peer reviews. Therefore, the approach taken was to review these PRA models and documentation to identify those items that may be directly relevant to the GSl-191 risk assessment, perform sensitivity analyses where appropriate , and discuss the results with dispositions for the uncertainties. E3-75

Enclosure 3 Risk Quantification 14.3.1 Internal Events PRA Model Uncertainty The epistemic uncertainty analysis approach described below applies to the Internal Events PRA. The baseline Internal Events PRA model uncertainty report was developed based on the guidance in NUREG-1855 (Reference 11 ). As described in NUREG-1855, sources of uncertainty include "parametric" uncertainties, "modeling" uncertainties, and "completeness" (or scope and level of detail) uncertainties. (The epistemic uncertainties unique to the seismic PRA are addressed in a later section.) Parametric uncertainty was addressed as part of the VEGP baseline PRA model aleatory uncertainty analysis. Assumptions are made during the PRA development as a way to address a particular modeling uncertainty because there is not a single definitive approach. The assumptions are defined consistent with the definition provided in NUREG-1855 (Reference 11 ). Plant-specific assumptions made for each of the VEGP Internal Events PRA technical elements are noted in the individual PRA notebooks. These assumptions were collected from each notebook and evaluated to determine if they are related to source of modeling uncertainty, and if so that uncertainty was characterized. In addition, EPRI TR-1016737 (Reference 12) compiled a listing of generic sources of modeling uncertainty for each PRA technical element, which were also considered. Completeness uncertainty addresses scope and level of detail of the PRA model. Uncertainties associated with scope and level of detail are documented in the PRA, but are only considered for their impact on a specific application. From the characterization of potential sources of uncertainty in the baseline Internal Events PRA model and of supplementary issues from EPRI TR-1 016737 (Reference 12), the following items may impact the internal events PRA results. Sensitivity analyses are included to further evaluate these items as a source of uncertainty. Table 3-18 provides a summary of the evaluation assessing the impact of the identified sources of model (epistemic) uncertainty on the GSl-191 risk assessment. For each of the sources of uncertainty, the potential impact on the GSl-191 risk assessment is addressed, either qualitatively or by an appropriate sensitivity case, to determine the impact on the GSl-191 application. E3-76

Table 3 Assessment of VEGP Internal Events PRA Epistemic Uncertainty Impacts Source of Epistemic Related Assumptions Sensitivity Case Disposition Uncertainty High RCS pressure Scenarios with significant Sensitivity cases were The GSl-191 risk impacts the potential for RCP seal leakage, a stuck performed for the assessment demonstrates induced steam generator open pressurizer valve, or a baseline Internal Events that only large LOCAs could tube rupture (SGTR). pressurizer PORV open for PRA by reclassifying the result in debris related feed-and-bleed cooling are identified scenarios as failures. Therefore, the Medium and large LOCA conservatively considered low RCS pressure to possible overestimation of and reactor vessel rupture high RCS pressure determine impact on induced SGTR for high are treated as low RCS scenarios. LERF. pressure scenarios has no pressure scenarios. All impact on the GSl-191 risk other core damage assessment. sequences are considered high pressure sequences where the induced SGTR failure mode is possible. Therefore, the baseline PRA model may overestimate the contribution of induced SGTR to LERF. Certain initiating events The generic industry None The GSl-191 phenomena can be affected by frequency for the LOSP are of concern for initiating seasonal variations (e.g., event developed in events that could generate loss of offsite power NUREG/CR-6890 is debris from insulation (LOSP), loss of service applicable to the VEGP site. materials and coatings inside water (SW), etc.) and The NSCW cooling towers containment, which could baseline PRA does not are not required during cold then be transported to the address seasonal weather months. containment sump and fail variations. the ECCS sump suction strainers during the recirculation phase needed

Enclosure 3 Risk Quantification Source of Epistemic Related Assumptions Sensitivity Case Disposition Uncertainty to maintain core cooling. For VEGP, the initiating events that meet these criteria are LOCAs and SSBI. Therefore, seasonal variations of certain other initiating events have no impact on the GSl-191 risk assessment. The method of calculation Detailed evaluations of The overall modeling Since the VEGP PRA model of human error probabilities HEPs are performed for the uncertainty associated is based on industry (HEPs) for the Human risk significant, pre- and with the general basis consensus modeling Reliability Analysis (HRA) post-initiator human failure for HEPs is addressed approaches for its HEP may introduce uncertainty events (HFEs) using by the standard baseline calculations, and there are based on the particular industry consensus PRA HEP sensitivity no additional HFEs added for methodology applied. methods. The Technique for cases for the internal the GSl-191 risk Human Error Rate events PRA. assessment, this is not Prediction (THERP) method considered a significant is applied for pre-initiator source of epistemic HFEs. The Cause-Based uncertainty and therefore Decision Tree Method has no impact on the GSI-(CBDTM) is used for 191 risk assessment. cognitive errors and THERP for execution errors for post-initiator HFEs. The VEGP PRA medium None A sensitivity was The LOCA frequency values LOCA frequency is based performed for the in NUREG/CR-6928 are in upon data from Internal Ev~nts to turn based on the LOCA NUREG/CR-6928, which is determine the impact of frequency data from an order of magnitude the increased medium NUREG-1829. NUREG-1829 higher than the previous LOCA frequency from data are used to develop E3-78

Enclosure 3 Risk Quantification Source of Epistemic Related Assumptions Sensitivity Case Disposition Uncertainty data used from NUREG/CR-5750. A LOCA frequencies for the NUREG/CR-5750. more than 10% increase GSl-191 risk assessment. in CDF and nearly 9% The GSl-191 risk impact, increase in LERF occurs however, is not sensitive to due to the updated data. the initiating event frequency for medium LOCAs. No medium LOCAs result in sump strainer or core failures due to the effects of debris. Steam generator (SG) tube If SG tube condition A sensitivity analysis The GSl-191 risk condition affects the degrades, the induced was performed with assessment demonstrates probabilities of induced SGTR probability during average vs. pristine SG that only large LOCA could SGTR. The current VEGP secondary side break or tube conditions. CDF result in sump strainer 3-18SG tube condition is anticipated transient without increased by slightly failure. Therefore, the pristine. scram for pressure- or more than 1%, while possible under-estimation of thermal-induced SGTR in LERF nearly tripled. induced SGTR for high the LERF analysis would pressure scenarios has no increase. impact on the GSl-191 risk assessment. The presence of water in The base internal events A sensitivity study was The risk increase for large the reactor cavity at the VEGP Level 2 PRA performed for a wet early release due to GSl-191 time of vessel breach assumes a dry reactor cavity reactor cavity. CDF is nearly three orders of would affect the probability condition. increased by less than magnitude below the RG of containment failures 2%, and LERF 1.174 Region Ill risk (early release due to steam increased by more than acceptance criteria. A 12% explosion and late release 12%. increase in the large early due to base mat melt release frequency would still through). be well below the Region Ill threshold. E3-79

Enclosure 3 Risk Quantification 14.3.2 Seismic PRA Model Uncertainty The ASME/ANS PRA Standard (Reference 13) and RG 1.200 (Reference 14) include a number of requirements related to identification and evaluation of the impact of assumptions and sources of uncertainty on the PRA results. NUREG-1855 (Reference

11) and EPRI TR-1016737 (Reference 12) provide guidance on assessment of uncertainty for applications of a PRA. Sources of uncertainty within the VEGP seismic PRA model are addressed as follows:

• Parametric uncertainty was addressed as part of the VEGP seismic PRA model quantification.
• Modeling uncertainties and associated assumptions specific to the VEGP seismic PRA technical elements are noted in the seismic PRA documentation and were subject to peer review.
• Completeness uncertainty addresses scope and level of detail. Uncertainties associated with scope and level of detail are documented in the seismic PRA. No specific completeness issues were identified in the VEGP seismic PRA peer review.

A summary of potentially important sources of uncertainty in the VEGP seismic PRA is provided in Table 3-19. E3-80

Enclosure 3 Risk Quantification Table 3-19 -Assessment of VEGP Seismic PRA Uncertainty mpacts PRA Summary of Treatment of Sources Potential Impact on Element of Uncertainty Seismic PRA Results Seismic The VEGP seismic PRA peer review team With regard to aleatory and Hazard noted that both the aleatory and epistemic epistemic uncertainties in the uncertainties were addressed by site response analysis, there characterizing the seismic sources. is an abundance of site-specific data from VEGP Units 3 and 4 that reduces epistemic uncertainty to an The review team commented that the site insignificant level. response analysis did not fully evaluate and model aleatory and epistemic The characterization of the uncertainties in the site response analysis. seismic hazard reasonably reflects sources of uncertainty. Seismic The seismic PRA peer review team had no Several sensitivity studies Fragilities comments on sources of uncertainty evaluate the impact of pertaining to fragilities. changes to fragilities on the seismic PRA results as one means of assessing the impact of fragilities uncertainties on the seismic PRA results. No changes to the model were recommended based on these results. Seismic The seismic PRA peer review team The seismic PRA PRA commented that the VEGP seismic PRA quantification report includes Model team relied on the UN CERT code for the a discussion of sources of propagation of the parametric uncertainties model uncertainty, and in the seismic PRA with little explanation or potentially important sources documentation of the meaning of the have been addressed in the uncertainties results. sensitivity analysis. No changes to the model were recommended based on these results. 15.0 References

1. Regulatory Guide 1.174, Revision 2, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," May 2011

2. NEI 04-07 Volume 2, Revision 0, "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'," December 2004 E3-81

Enclosure 3 Risk Quantification

3. NUREG-1829, "Estimating Loss-of-Coolant Accident (LOCA) Frequencies Through the Elicitation Process," April 2008

4. ANSl/ANS-5.1-1979, "American National Standard for Decay Heat Power in Light Water Reactors," August 1979

5. WCAP-17788-P, Revision 0, "Comprehensive Analysis and Test Program for GSl-191 Closure (PA-SEE-1090)," July 2015

6. WCAP-16530-NP-A, "Evaluation of Post-Accident Chemical Effects in Containment Sump Fluids to Support GSl-191," March 2008

7. NARWHAL-SUM-02, Revision 1, "NARWHAL Version 2.1 Software User's Manual,"

September 9, 2016

8. Howe, Kerry J., ET. Al, "Corrosion and solubility in a TSP-buffered chemical environment following a loss of coolant accident: Part 1 - Aluminum," Nuclear Engineering and Design, Volume 292, October 2015: 296-305

9. ML080230038, "NRC Staff Review Guidance Regarding Generic Letter 2004-02 Closure in the Area of Strainer Head Loss and Vortexing," March 2008

10. Regulatory Guide 1.82, Revision 4, "Water Sources for Long-Term Recirculation Cooling Following a Loss-of-Coolant Accident," March 2012

11. NUREG-1855, Revision 1, "Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decisionmaking," March 2017

12. EPRI Report 1016737, "Treatment of Parameter and Model Uncertainty for Probabilistic Risk Assessments," December 2008 13.ASME/ANS RA-Sa-2009, "Addenda to ASME/ANS RA-S-2008 Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications," February 2009

14. Regulatory Guide 1.200, Revision 2, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities,"

March 2009 E3-82

Vogtle Electric Generating Plant - Units 1 & 2 Supplemental Response to NRC Generic Letter 2004-02 Enclosure 4 Defense-in-Depth and Safety Margin

Enclosure 4 Defense-in-Depth and Safety Margin Table of Contents 1.0 Introduction 2.0 Defense-in-Depth 2.1 Evaluation for RG 1.174 Defense-in-Depth Philosophy 2.2 Detecting and Mitigating Adverse Conditions 2.3 Barriers for Release of Radioactivity 2.4 Emergency Plan Actions 3.0 Safety Margin 4.0 References E4-1

Enclosure 4 Defense-in-Depth and Safety Margin 1.0 Introduction For the purpose of this VEGP risk-informed GSl-191 submittal, defense-in-depth (DID) is defined as the response to the question of what happens if the analysis is wrong about a successful end state and it actually turns out to be a failure. DID includes mitigative design features and actions that address protection of the public from radiation due to sequences that go to failure (e.g., containment integrity, emergency plans, operator actions not credited in the GSl-191 evaluation, use of FLEX, etc.). Similarly, safety margin is defined as the response to the question of what aspects of the analysis increase confidence that a declared success is a success. Therefore, safety margin is a combination of built-in conservatisms that increase confidence that scenarios that go to success remain in success (and why some scenarios that are assumed to fail might actually succeed). The DID evaluation shows that there is adequate system capability to provide assurance that public health and safety are protected in the event that a LOCA results in strainer blockage or loss of long-term core cooling due to effects of LOCA-generated debris. It identifies operator actions that can be taken to mitigate the event and describes the robustness of the design for the VEGP containment buildings. The safety margin evaluation identifies many conservatisms throughout the evaluation, which provides high confidence that successful end states are truly successful, and that many end states that are assumed to fail in reality would also be successful. The conclusion of the evaluation is that there is substantial DID and safety margin. 2.0 Defense-in-Depth The evaluation of DID first addresses whether the impact of the proposed licensing basis (LB) change (individually and cumulatively) is consistent with the DID philosophy, as outlined in Regulatory Guide (RG) 1.174 (Reference 1). This section also presents the measures available to VEGP for preventing, detecting, and mitigating conditions that could challenge long-term core cooling due to strainer blockage and inadequate cooling flow to the reactor core. Finally, the evaluation shows if and how the proposed changes affect the barriers for release of radioactivity and emergency plan actions. 2.1 Evaluation for RG 1.174 DID Philosophy VEGP is proposing a licensing basis change to use a risk-informed approach to address the concerns of GSl-191 with respect to maintaining long-term core cooling following a LOCA. An evaluation was performed to determine whether the change meets the DID principles defined in RG 1.174 (Reference 1). As stated in the RG, consistency with the DID philosophy is achieved if the following occurs: i I L______ ~ E4-2

Enclosure 4 Defense-in-Depth and Safety Margin

• A reasonable balance is preserved among prevention of core damage, prevention of containment failure, and consequence mitigation. VEGP has performed various physical and procedural changes, for example, installation of new strainers with increased surface areas and a reduced opening size, increased RWST inventory to sump pool, removal of problematic insulation materials, procedural changes to delay isolation of RHR pumps from RWST, and program controls to ensure the debris load limits are not exceeded. Additional changes are being planned, for example, modifying the height of the RHR strainers and sump recirculation initiation sequence. These changes reduced the risk associated with the effects of LOCA-generated debris. The new risk-informed elements of the analysis showed a very small increase in risk of containment or reactor failures related to GSl-191, as demonstrated by the very small ~CDF and ~LERF per the RG 1.174 criteria (Reference 1). Therefore, the existing balance among prevention of core damage, prevention of containment failure, and consequence mitigation is preserved.
• Over-reliance on programmatic activities as compensatory measures associated with the change in the licensing basis is avoided. The proposed licensing basis change does not adversely impact any of the programmatic activities, such as the in-service inspection (ISi) program, plant personnel training, RCS leakage detection program, or containment cleanliness inspection activities. Therefore, the licensing change will not cause any over-reliance on these activities.
• System redundancy, independence, and diversity are preserved commensurate with the expected frequency, consequences of challenges to the system, and uncertainties (e.g., no risk outliers). As discussed above, the modifications made as part of the proposed licensing basis change do not change the redundancy, independence, and diversity of the ECCS or containment spray system. These systems have been fully analyzed relative to their contribution to nuclear safety through plant-specific PRA. The risk contribution related to GSl-191 due to the proposed licensing basis change has also been evaluated for the full spectrum of LOCA events. As described in Enclosure 3, Section 14.4, the uncertainties in the risk-informed approach were examined. Although the use of alternate models or variations in inputs can in some cases result in higher calculated LiCDF values, all uncertainty quantification cases showed risk results in Region Ill of RG 1.174 (Reference 1), which provides high confidence that there are no risk outliers.
• Defenses against potential common-cause failures are preserved, and the potential for the introduction of new common-cause failure mechanisms is assessed. The potential for new common-cause failure mechanisms has been assessed for the GSl-191 issues. The primary failure mechanism includes clogging of the sump strainers and/or reactor core, which is not a new failure mechanism. The defenses against these clogging mechanisms are not affected by the physical and procedural changes. Additionally, the new risk-informed E4-3

Enclosure 4 Defense-in-Depth and Safety Margin approach does not introduce any new common-cause failures or reduce the current plant defenses against common-cause failures.

• Independence of barriers is not degraded. The three barriers to a radioactive release are the fuel cladding, RCS pressure boundary, and reactor containment building. For the evaluation of a LOCA, the RCS barrier is postulated to be breached. The proposed licensing basis change for the use of a risk-informed approach to evaluate the effects of LOCA-generated debris does not affect the design and analysis requirements for the fuel. Therefore, the fuel barrier independence is not degraded.

The post-LOCA recirculation function is provided by the ECCS located inside the auxiliary building. During the recirculation phase, the RHR pumps take suction from the containment recirculation sumps and supply flow back to the reactor directly and/or through the CCPs and SI pumps. The pumps, system piping and other components on the recirculation flow path serve as the barrier to release. The auxiliary building has a dedicated ventilation system to control airborne radioactivity during emergency conditions and the building is capable of handling recirculating water leakage. The proposed licensing basis change does not alter the design and operating requirements for ECCS or auxiliary building. Analyses have been performed to show that, assuming a single failure that results in the loss of one air cooling train and one CS train, the containment fan coolers and the CS system can remove sufficient thermal energy from the containment atmosphere following a LOCA or MSLB to maintain the peak containment pressure below design values. The licensing basis change does not alter the design or operating requirements of these systems. It is therefore reasonable to conclude that the independence of the barriers is maintained and not degraded by the licensing basis change.

• Defenses against human errors are preserved. The use of the risk-informed methodology in the GSl-191 analysis does not impose any additional operator actions or increase the complexity of existing operator actions. Thus, the defenses that are already in place with respect to human errors are not impacted by the proposed licensing basis change.
• The intent of the plant's design criteria is maintained. The proposed licensing basis change does not alter any of the ECCS acceptance criteria specified in 10 CFR 50.46. Additionally, the proposed change does not affect the design or design requirements of the plant equipment associated with GSl-191. As discussed above, the risk-informed analysis shows that the risk increase due to GSl-191 related failures is very small and meets the RG 1.174 acceptance criteria (Reference 1). Therefore, the intent of the plant's design criteria is maintained.

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Enclosure 4 Defense-in-Depth arid Safety Margin 2.2 Detecting and Mitigating Adverse Conditions For the purposes of GSl-191 resolution, the primary regulatory objective is specified in 10 CFR 50.46(b)(5) as maintaining long-term core cooling. Adequate DID is maintained by ensuring the capability exists for operators to detect and mitigate adverse conditions due to potential impacts of debris blockage, such as inadequate flow through the strainers and/or through the reactor core. This section evaluates the VEGP DID measures for detecting and mitigating adverse conditions in order to support the VEGP application for a risk-informed approach to resolve GSl-191. Inadequate strainer flow refers to the condition where significant pump cavitation occurs due to inadequate RHR and/or CS pump NPSH margin associated with the high head losses across the sump strainers and debris bed. For VEGP, testing was performed to measure the debris bed head losses using a prototypical strainer configuration and post-LOCA conditions. The effect of debris head loss was conservatively accounted for in the risk-informed analysis. Inadequate reactor core flow refers to the condition where the normal core cooling flow path has become impeded (blocked) and is not allowing sufficient cooling water to reach the core. This condition could result from the formation of a debris bed at the reactor core inlet or at the fuel grid inside the core due to debris that passes through the sump strainers. The effect of debris accumulation in the reactor core was conservatively accounted for in the risk-informed analysis. 2.2.1 Prevention of Strainer Blockage The primary means to delay or prevent strainer blockage is to monitor and reduce the flow through the sump strainers as necessary, and control debris sources inside containment. Specific measures are laid out as follows.

• Various VEGP emergency operating procedures (EOPs) provide the operators with guidance on monitoring sump strainer blockage (e.g., Procedures 19010-C "E-1 Loss of Reactor or Secondary Coolant", 19013-C "ES-1.3 Transfer to Cold Leg Recirculation", and 19111-C "ECA-1.1 Loss of Emergency Coolant Recirculation"). If sump blockage is detected, Procedure 19113-C "ECA-1.3 Recirculation Sump Blockage" provides actions that operators should take to respond to the condition.
• VEGP EOPs incorporated the Bulletin 2003-01 training and procedural guidance to expedite plant cooldown in response to a small break LOCA.
• For small to medium break LOCAs, depletion of the RWST,can be delayed by following Procedure 19012-C "ES-1.2 Post LOCA Cooldown and Depressurization". This procedure provides actions to cool down and depressurize the RCS to reduce the break flow, thereby lowering the injection flow necessary to maintain RCS subcooling and inventory. It is possible to bring the plant to cold shutdown conditions before the RWST is drained to the sump E4-5

Enclosure 4 Defense-in-Depth and Safety Margin recirculation switchover level. Therefore, sump recirculation may not be required and, in that case, sump blockage would not be an issue. 1

• The Technical Specification minimum required RWST volume is 686,000 gallons, and the low-level alarm setting is 696,448 gallons. The RWST level is normally maintained above the low-level alarm setting, and the nominal volume of the tank is 715,000 gallons.

Several measures are in place to control the debris sources inside the VEGP containment buildings.

• Training is provided to personnel accessing containment to raise their awareness of the more stringent containment cleanliness requirements, the potential for sump blockage, and actions being taken to address sump blockage concerns.
• For the Technical Requirements Manual Surveillance Requirement (TRS) 13.5.1.1, VEGP has implemented Procedure 14900-C, "Containment Exit Inspection" in conjunction with 00303-C, "Containment Entry". Per these procedures, prior to entering Mode 4 (Hot Shutdown) from Mode 5, several walk-downs are required to be performed by station management and operations personnel to ensure the containment buildings are free of loose debris. For subsequent entries, inspections of the travel path and work locations are required to ensure the areas are free of loose debris.
• For the Technical Specification Surveillance Requirement (SR) 3.5.2. 7, VEGP has implemented procedures 14903-1/2 both titled "Containment Emergency Sump Inspection", to verify by visual inspection that the suction inlets are not restricted by debris and that the sump strainers are correctly configured according to plant design and show no structural distress or abnormal corrosion.

These procedures also ensure that the protective covers for the TSP baskets and sump strainer are removed. This inspection is required on an 18-month frequency in accordance with the Surveillance Frequency Control Program.

• VEGP Procedure 00309-C is used to control unattended temporary materials in containment. The program includes periodic surveillance and assessment of containment material conditions during Modes 1-4. It imposes strict controls on the types and quantities of materials that may be taken into containment.
• Inspections of the coatings in containment are part of a protective coatings program complying with Regulatory Guide 1.54 (Reference 2) and ANSI N 101.4-1972 (Reference 3), to ensure that coatings do not adversely affect safety-related systems, structures, or components.

2.2.2 Detection of Strainer Blockage During sump recirculation following a LOCA, accumulation of fiber, particulate, and chemical debris on the strainer could cause high flow head losses which may challenge the operation of the RHR and CS pumps. This, in turn, could result in a condition where insufficient cooling is provided for reactor core cooling and/or containment pressure control. When such a condition exists, it is important for the plant operators to be able E4-6

Enclosure 4 Defense-in-Depth and Safety Margin to detect this condition in a timely manner. VEGP maintains a post-accident monitoring instrumentation program, which ensures the capability to monitor plant variables and system status during and following an accident. This program includes those instruments that indicate system status and furnish information regarding the release of radioactive materials, in accordance with Regulatory Guide 1.97 Revision 2 (Reference 4). VEGP has the following methods for detection of sump strainer blockage conditions.

• VEGP has indications in the control room for SI, RHR, and CS pump flows and SI and CS pump discharge pressures. Instrumentation is available to provide the operators with indications of pump cavitation, such as erratic flow or low discharge pressure.
• VEGP has core exit thermocouple (CET) and reactor vessel level indications in the control room to allow monitoring for any potential reduction in core cooling flow due to sump blockage. This indication is also displayed on the computer systems as part of the critical safety system status tree indicators, monitored by the reactor operators and shift technical advisor. The status tree indicators provide changes based on status tree logic to enhance operator recognition of a distress condition developing.

2.2.3 Mitigation of Strainer Blockage Multiple methods are available to mitigate an inadequate recirculation flow condition caused by the accumulation of debris on the sump strainer.

• The VEGP EOPs contain steps to reduce flow through the system up to and including stopping all pumps taking suction from a clogged sump strainer. It has been observed, during strainer head loss testing, that stopping all flow through a debris-laden strainer could dislodge portions of the debris bed from the strainer because the force that holds the debris bed in place was the flow head loss through the debris. This is also an important measure to avoid permanent pump damage that could be caused by the loss of suction condition.
• VEGP Procedure 19111-C "ECA-1.1 Loss of Emergency Coolant Recirculation" minimizes the pumps required depending on plant conditions and directs shutting down all pumps as applicable. Procedure 19113-C "ECA-1.3 Recirculation Sump Blockage" also contains steps to shut down SI pumps and CCPs that piggyback off a potentially cavitating RHR pump during recirculation.
• VEGP Procedure 19113-C "ECA-1.3 Recirculation Sump Blockage" contains steps to initiate makeup to the RWST from, for example, the spent fuel pool.

This would allow realignment of SI and CS pumps to the direct injection flow path from the RWST and provide necessary cooling for an extended period. The operators would establish the minimum flow required for core decay heat removal depending on sub-cooling conditions.

• In response to the Nuclear Regulatory Commission (NRC) Order EA-12-049 (Reference 5), "Mitigation Strategies for Beyond-Design-Basis External Event E4-7

Enclosure 4 Defense-in-Depth and Safety Margin (BDBEE)", VEGP developed diverse and flexible coping strategies (FLEX) to maintain fuel cooling (spent fuel pool and core) and containment integrity. Various modifications have been implemented such that non-emergency equipment can be credited during a BDBEE. For example, the Auxiliary Feedwater System can be used to deliver cooling water from the condensate storage tank (CST) to the steam generators for reactor core cooling. Makeup capabilities were added to refill the CST and Reactor Make-up Water Storage Tank (RMWST), which would serve as suction sources for core cooling. 2.2.4 Prevention of Inadequate Reactor Core Flow The set of actions identified in Section 2.2.1 for reducing or controlling flow through the emergency sump strainers during the recirculation phase can have a similar positive impact on reducing the potential for fuel blockage. Controlling flow to the reactor vessel to maintain fuel coverage and match decay heat has benefits through reduced head loss and delayed onset of any chemical precipitates. The VEGP plant design has simultaneous hot leg and cold leg injection once the RWST is depleted and the RHR and SI pumps have been realigned during the recirculation phase. Initially all of the ECCS pumps would be aligned for cold leg injection. At 7.5 hours after the initiating event, the switchover to simultaneous hot/cold leg injection would be made. For this configuration, the RHR and SI pumps provide cooling water through the hot leg while the CCP continues injecting coolant through the cold leg. It is expected that, with most of the flow traveling through the hot leg, the motive force that holds the debris at the core inlet would be removed and the flow from the hot legs would travel down the heated core to the inlet, which could dislodge the debris bed at the core inlet. 2.2.5 Detection of Inadequate Reactor Core Flow Multiple methods exist for detection of a core blockage condition as manifested by an inadequate RCS inventory or inadequate RCS and core heat removal conditions. The primary methods for detection include core exit thermocouple (GET) temperature indication and reactor water level, as monitored by the reactor vessel level instrumentation system (RVLIS). An additional method for detection of a core blockage condition includes monitoring of containment radiation levels.

• Core exit temperature behavior is the primary indicator of adequate core cooling.

If cold leg recirculation has been established with flow maintained into the RCS, core exit temperature should be stable or slowly lowering during accident recovery. Increasing core exit temperatures while injection flow is maintained, regardless of reactor vessel water level behavior, could be an indication of insufficient core flow. In this regard, VEGP's functional restoration procedure would attempt to establish injection flow of clean water from the RWST. CETs E4-8

Enclosure 4 Defense-in-Depth and Safety Margin are monitored during EOP usage as well as for status tree functional restoration entries and the safety parameter display system (SPDS).

• Reactor vessel water level is also monitored and a decreasing water level could indicate a lower core region flow blockage. VEGP employs the RVLIS to provide instrumentation for the detection of inadequate core cooling. The RVLIS utilizes two sets of differential pressure cells to measure reactor vessel level continuously. The measurement provides an approximate indication of the relative void content or density of the circulating fluid.
• Increasing radiation levels are indicated by alarms in the control room with specific procedural steps in both alarm response procedures and EOPs for addressing the condition. Radiation monitor indication in the auxiliary building may be indication of a LOCA outside containment or provide initial entry conditions due to increasing radiation levels. Abnormal containment radiation could be an indication of fission product barrier degradation, which is monitored by the control room. Due to the sensitivity of the monitors and the low alarm set points, identification of degrading core conditions is expected well before a significant release of radioactivity to containment occurs.

2.2.6 Mitigation of Inadequate Reactor Core Flow Multiple methods are available to mitigate an inadequate reactor core flow condition, as laid out in Procedures 19221-1/2 "FR-C.1 Response to Inadequate Core Cooling" and 19222-1/2 "FR-C.2 Response to Degraded Core Cooling". Upon identification of an inadequate RCS inventory or an inadequate core heat removal condition, the EOPs direct the operators to take actions to restore cooling flow to the RCS including:

• Reestablish SI flow to the RCS
• Reduce RCS pressure by performing rapid secondary depressurization
• Restart RCPs and open pressurizer PORVs These actions are to be performed sequentially. Success, as indicated by improved core cooling and increasing vessel inventory, is evaluated prior to performing the next action in the sequence. Re-initiation of high pressure SI may be, depending on the cause of inadequate core cooling, the most effective method to recover the core and restore adequate core cooling. If some form of high-pressure injection cannot be established or is ineffective in restoring adequate core cooling, the operator takes actions to reduce the RCS pressure in order for the SI accumulators and low-head pumps to inject. Analyses have shown that a rapid secondary depressurization is the most effective means for achieving this objective. If secondary depressurization is not possible, or primary to secondary heat transfer is significantly degraded, and at least one idle SG is available, the operator can start the RCP(s) associated with the available idle SG(s). The RCPs will provide forced two-phase flow through the core and temporarily improve core cooling until some form of makeup flow to the RCS can be established.

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Enclosure 4 Defense-in-Depth and Safety Margin VEGP has also implemented procedures per the severe accident management guidelines (SAMG) which provide the operator with actions to protect fission product boundaries and return the plant to a controlled stable condition when the emergency operating procedures are no longer effective in controlling the casualty. Entry into the SAMG procedures is directed by the emergency operating procedures when certain conditions are met. Some of the operator actions outlined in the SAMG procedures can help maintain reactor core flow, for example, injection into SGs and RCS, depressurization of RCS, makeup to RWST, realignment to injection from RWST, and flooding the containment. Cooling can also be provided to the reactor core using the flow paths established by the FLEX strategy or by reinitiating injection through a refilled RWST, as discussed in Section 2.2.3. If it is determined that the inadequate core cooling condition is caused by clogged sump strainers, the actions discussed in Section 2.2.3 can also be taken to reestablish cooling flow through the strainers. 2.3 Barriers for Release of Radioactivity The purpose of this section is to demonstrate that there are additional defense in depth measures to protect the current barriers for release of radioactivity. The three barriers are the fuel cladding, the RCS boundary, and the reactor containment building. Each of these barriers is addressed in the subsections below. 2.3.1 Fuel Cladding Following a LOCA, the ECCS provides both the initial phase of accident mitigation and long-term cooling to the fuel cladding barrier. For the initial phase of accident mitigation, the proposed licensing basis change for the use of a risk-informed approach to evaluate the effects of debris does not alter the fuel cladding limits, or previous analysis and testing programs that demonstrate the acceptability of ECCS. The primary goal of the VEGP SAMG procedures is to protect fission product boundaries and mitigate any ongoing fission product releases in the event that conditions warrant entry into the SAMGs. Some of the operator actions outlined in the SAMG procedures can help maintain reactor core flow and integrity of the fuel cladding, for example, injection into SGs and RCS, depressurization of RCS, makeup to RWST, realignment to injection from RWST and flooding the containment. 2.3.2 RCS Pressure Boundary The integrity of the RCS pressure boundary is assumed to be compromised for the GSl-191 sump performance evaluation. However, the proposed licensing basis change does not modify the previous analyses or testing programs that demonstrate the integrity of the RCS. Additional measures are in place to prevent and detect pipe breaks, as discussed below. E4-10

Enclosure 4 Defense-in-Depth and Safety Margin

• The inservice inspection (ISi) program provides rules for the examination and repair of piping and other RCS components, and plays an important role in the prevention of pipe breaks. The integrity of the Class 1 welds, piping, and components are maintained at a high level of reliability through the ASME Section XI inspection program (Reference 6). VEGP ISi procedures also ensure that inspections are performed in accordance with the schedule requirements of the code.
• RCS overpressure protection is provided by the pressurizer safety valves, steam generator safety valves, and the reactor protection system and associated equipment. Combinations of these systems ensure compliance with the overpressure requirements of the ASME Boiler and Pressure Vessel Code, Section Ill, Paragraphs NB-7300 and NC-7300, for pressurized water reactor systems (Reference 7).
• The leak detection program at VEGP is capable of early identification of RCS leakage in accordance with RG 1.45 (Reference 8) to provide time for appropriate operator action to identify and address RCS leakage. The effectiveness of this program is not reduced by the proposed licensing basis change to the risk-informed approach for GSl-191.
• Some of the operator actions outlined in the VEGP SAMG procedures can help maintain integrity of the RCS when directed by the emergency operating procedures. Such actions include injection into SGs and RCS, depressurization of RCS, makeup to RWST, realignment to injection from RWST and flooding containment.

2.3.3 Reactor Containment Integrity The VEGP containment buildings are designed such that for all break sizes, up to and including a double-ended guillotine break of an RCS pipe or secondary system pipe, the containment peak pressure is below the design pressure with adequate margin. This has been demonstrated by previous analyses based on conservative assumptions (e.g., minimum heat removal and maximum containment pressure). The analyses also considered the worst single active failure affecting the operation of the ECCS, CSS, and containment fan coolers during the injection phase, and the worst active or passive single failure during the recirculation phase. For primary system breaks, loss of offsite power is also assumed. The analyses showed that the containment fan coolers, in conjunction with the CS system, can remove sufficient thermal energy and decay heat from the containment atmosphere following a LOCA or MSLB to maintain the containment pressure below design values. Therefore, the containment buildings remain a low leakage barrier against the release of fission products for the duration of the postulated LOCAs. The evaluation of post-LOCA debris effects using a risk-informed approach is not part of the analyses that demonstrate containment integrity. The proposed licensing basis change does not affect the methodology, acceptance criteria, or conclusion of the existing analysis. Therefore, the reactor containment integrity is not affected. E4-11

                                                                                         ,j

Enclosure 4 Defense-in-Depth and Safety Margin Additionally, some of the operator actions outlined in the VEGP SAMG procedures can help maintain integrity of the containment when directed by the emergency operating procedures. Such actions include control of containment pressure and hydrogen concentration. 2.4 Emergency Plan Actions The proposed change to the licensing basis to use the methodology of a risk-informed approach does not involve any changes to the emergency plan. There is no change to the strategies for preventing core damage and containment failure, or for consequence mitigation. The use of the risk-informed approach does not impose any additional operator actions or complexity. Implementation of the proposed change would not result in any changes to the response requirements for emergency response personnel during an accident. 3.0 Safety Margin There are numerous conservatisms used throughout the risk-informed GSl-191 evaluation for VEGP. However, not all of these conservatisms were classified as safety margin. Some conservatisms were included to provide future operating margin (i.e., margin added to the current plant conditions to allow for future changes, and flexibility in conducting maintenance or inspections). The key distinction between safety margin and operating margin is that safety margin cannot be reduced without approval from the NRG (Reference 1), whereas operating margin can be modified if necessary based on plant changes. Table 4-1 describes the safety margins included in the risk-informed GSl-191 evaluation. As noted in this table, there are many conservatisms throughout the evaluation, which provide high confidence that successful end states are truly successful, and that many end states that are assumed to fail in reality would also be successful. Note that in several places, the effect of conservatism on the model is described as over-predicting or under-predicting a specific physical phenomena or failure. These terms are generically used to refer to either a change in the actual value predicted by the model or an increase in margin. For example, flashing is not predicted to occur. However, because of conservatism in specific portions of the model, the potential for flashing is over-predicted (i.e., there is more real margin to prevent flashing than is predicted by the model). E4-12 J

Enclosure 4 Defense-in-Depth and Safety Margin Table 4 Description of Safety Margin Conservatism Credited as Topic Realistic Conditions Impact on Evaluation Safety Margin Scenario All frequency associated with Smaller breaks on main steam Overall likelihood of failure is Frequency secondary side breaks is and feedwater piping are much over-predicted for secondary allocated to DEGBs more likely than DEGBs, and side breaks would generate significantly lower debris quantities Scenario Random pump failures are Random pump failures can Failures at the start of the Frequency assumed to occur at switchover occur at the beginning of the event would delay switchover to recirculation event, at the start of to recirculation, failures later recirculation or any time during in the event would result in the event distribution of debris across more strainers Thermal- Initial containment pressure is at Containment pressure would NPSH margin is under-Hydraulics technical specification (TS) be above TS minimum predicted and degasification minimum of -0.3 psig and flashing are over-predicted Thermal- No credit taken for containment The post-LOCA containment . NPSH margin is under-Hydraulics accident pressure in NPSH pressure would be significantly predicted and degasification calculations and minimal credit higher than the saturation and flashing are over-taken for degasification and pressure predicted flashing calculations E4-13

Enclosure 4 Defense-in-Depth and Safety Margin Conservatism Credited as Topic Realistic Conditions Impact on Evaluation Safety Margin Thermal- Design basis containment and Containment and sump Chemical release, precipitate Hydraulics sump temperature profiles used temperature profiles would be quantities, degasification, and for all break sizes significantly lower for smaller flashing are over-predicted, break sizes and aluminum solubility, corrected strainer head loss, and NPSH margin are under-predicted (all impacts are conservative with the exception of under-predicted aluminum solubility and corrected strainer head loss; however, the sensitivity analysis described in Enclosure 3, Section 14.3 showed that the conservative effects outweigh the non-conservative effects) Debris With the exception of shadowing Full offset of pipe DEGBs Quantity of debris generated Generation by concrete walls, no credit was (especially on the primary loop inside the ZOI is over-taken for structures or restraints piping) would be significantly predicted that would limit the quantity of limited due to physical debris generated within a break restraints; also, insulation and ZOI qualified coatings would not be completely destroyed within a given ZOI due to the shielding effects of equipment and other structures Debris 100% failure of unqualified Some types of unqualified Particulate debris quantity on Generation coatings for all breaks coatings may have a relatively strainers is over-predicted low failure fraction E4-14

Enclosure 4 Defense-in-Depth and Safety Margin Conservatism Credited as Topic Realistic Conditions Impact on Evaluation Safety Margin Debris Unqualified epoxy fails as 100% Epoxy coatings are likely to fail Unqualified coatings debris Generation particulate in a range of sizes (including transport and particulate both particulate and chips) debris quantity on strainers are over-predicted Debris Unqualified coatings fail at the Unqualified coatings would fail Unqualified coatings that fail Generation start of the accident gradually and may not fail until in upper containment after much later in the event sprays are secured would not transport Chemical Maximum pH for chemical Consistent time-dependent pH Precipitate quantity is over-Effects release and minimum pH for profile resulting in lower predicted and precipitates solubility release and/or increased would form later than solubility predicted Chemical No aluminum remains in solution Some breaks would never Aluminum precipitate quantity Effects after the solubility limit has been exceed the solubility limit, and and strainer head loss are reached or 24 hours (whichever breaks that do exceed the over-predicted comes first) solubility limit would still have some aluminum in solution Chemical All insulation debris is assumed In reality, a large fraction of the Aluminum and calcium Effects to be in the sump for the chemical debris would be captured in release from insulation are release calculation upper containment, and the over-predicted, resulting in an release of chemicals would be over-prediction of aluminum significantly reduced for breaks and calcium precipitate where containment sprays are quantities not initiated E4-15

Enclosure 4 Defense-in-Depth and Safety Margin Conservatism Credited as Topic Realistic Conditions Impact on Evaluation Safety Margin Debris Fine debris has a high A condensate washdown of The quantity of fine debris Transport condensate washdown fraction 1% is a realistic estimate washed down to lower (10%) when sprays are not (Reference 10). containment (and initiated subsequently transported to the strainers and core) is over-predicted for breaks that do not initiate containment sprays Debris Fine debris has a high spray Some fine debris would be The quantity of fine debris Transport washdown fraction (100%) when blown to locations shielded washed down to lower sprays are initiated from containment sprays and containment (and would be retained in these subsequently transported to locations for the duration of the the strainers and core) is event over-predicted for breaks that initiate containment sprays Debris Fine debris has a high Some fine debris would settle The quantity of fine debris Transport recirculation transport fraction and be retained in stagnant transported to the strainers (100%) for all breaks regions of the recirculation and core is over-predicted pool (especially for cases where fewer pumps are operating) Debris Small and large pieces of Sustained movement of a The quantity of small and Transport fiberglass transport at the piece of debris all the way to large piece debris transported incipient tumbling velocity for the the strainer would require a to the strainers is over-respective debris sizes (note that somewhat higher fluid velocity, predicted the incipient tumbling velocity is particularly in cases where defined as the minimum fluid large debris quantities velocity at which an individual (including a mixture of sizes) piece would begin to move would result in agglomeration (Reference 11 )) of the debris on the containment floor E4-16

Enclosure 4 Defense-in-Depth and Safety Margin Conservatism Credited as Topic Realistic Conditions Impact on Evaluation Safety Margin Debris Small and large pieces of Based on 30-day erosion test The quantity of fines Transport fiberglass debris have a high results, the erosion fraction for generated and subsequently containment pool erosion fraction small pieces of fiberglass transported to the strainers (10%) would be somewhat less than and core is over-predicted 10% and the erosion fraction for large pieces of fiberglass would be less than small pieces Strainer/Pump A strainer is assumed to fail any In many cases, one type of The breaks that fail the Failures time the accumulated debris debris (e.g., calcium - strainer acceptance criteria quantities exceed the tested phosphate) exceeds the tested are over-predicted debris quantities quantity while other types of debris (e.g., sodium aluminum silicate) are significantly below the tested quantity; also, most breaks have available margin to accommodate higher head losses Strainer/Pump Miscellaneous debris (e.g., tags It is likely that a large portion of The strainer surface area is Failures or labels) all transports to the the miscellaneous debris under-predicted, and strainer strainers prior to any other debris would not transport to the head loss and debris limit and reduces the effective strainer strainers, and any failures are over-predicted area miscellaneous debris that does transport would tend to arrive along with or after other debris E4-17

Enclosure 4 Defense-in-Depth and Safety Margin Conservatism Credited as Topic Realistic Conditions Impact on Evaluation Safety Margin Strainer/Pump Debris head loss was Head loss would increase The strainer head loss for Failures conservatively calculated using a gradually as debris both conventional and rule-based approach (i.e., if the accumulates and most breaks chemical debris is over-accumulation of a given debris would not accumulate enough predicted type exceeds a certain threshold, debris to reach the head a bounding head loss is losses that were applied automatically applied) Strainer/Pump Strainer head loss testing was Because flow preferentially The strainer head loss for Failures conservatively performed using a passes through the lower both conventional and strainer module with fewer disks disks, it is likely that a larger chemical debris is over-and scaled up to the full height quantity of debris could predicted strainers based on the area ratios accumulate on the full height strainers than predicted using a simple area ratio Strainer/Pump Calcium phosphate head loss Based on observations from The strainer head loss is Failures was applied for all breaks that autoclave tests described in over-predicted for calcium generate and transport a Enclosure 2 Section 3.o.2.9 phosphate debris sufficient quantity of fiber debris and other tests representative of VEGP conditions, calcium phosphate precipitation is either unlikely to occur or the actual precipitates would have a negligible effect on head loss Strainer/Pump The chemical head loss was The head loss associated with The strainer head loss is Failures extrapolated to 30 days and the the full chemical debris load is over-predicted early in the extrapolation constant was not likely to continue event applied 450 minutes after the increasing over 30 days, and start of the event even if it did, additional NPSH margin would be available later in the event as the pool temperature drops E4-18

Enclosure 4 Defense-in-Depth and Safety Margin Conservatism Credited as Topic Realistic Conditions Impact on Evaluation Safety Margin Strainer/Pump Strainer failure is assumed in all It is likely that the strainer Strainer structural failures are Failures cases where the head loss meets could withstand higher head over-predicted or exceeds the structural margin losses than predicted, and of the strainer even if a structural failure occurs, it may not result in a complete loss of functionality Strainer/Pump All gas voids formed by Due to the relatively low Gas void fractions at the Failures degasification were assumed to Froude number, gas voids are pumps are over-predicted transport to the pumps likely to accumulate in the strainer and vent back to the pool when the buoyancy of the accumulated air exceeds the strainer head loss Strainer/Pump Pump NPSH required was Small gas void fractions would Pump NPSH required is over-Failures adjusted for gas voids based on likely have a much smaller predicted and Pump NPSH very conservative guidance effect on NPSH required margin is under-predicted (Reference 13) when gas voids are present Core Failures The fiber penetration correlation The penetration of fiberglass Fiber penetration (and ignores effects of fiber and fines would be reduced by the subsequent accumulation on particulate interactions and accumulation of particulate the core) is over-predicted accumulation of pieces of and fiberglass pieces on the fiberglass strainer Core Failures The WCAP-16793-N P The peak cladding The peak cladding methodology for evaluating peak temperature and total temperature and total cladding temperature and total deposition thickness would be deposition thickness are over-deposition thickness due to lower predicted accumulation of debris on the fuel rods is conservative (Reference 14) E4-19 I_

Enclosure 4 Defense-in-Depth and Safety Margin Conservatism Credited as Topic Realistic Conditions Impact on Evaluation Safety Margin Core Failures Maximum boil-off flow rate with Debris transport to core inlet Debris accumulation on core additional 20% margin used to would be proportional to boil- inlet is over-predicted for cold calculate debris accumulation on off flow rate and actual boil-off leg breaks core inlet for cold leg breaks flow rate is likely to be lower Core Failures Fiber limits associated with core It is likely that significantly Core failures due to the blockage and boron precipitation larger quantities of debris accumulation of fiber debris are based on bounding tests and could accumulate in the core are over-predicted. analyses without resulting in core damage E4-20

Enclosure 4 Defense-in-Depth and Safety Margin For information, Table 4-2 shows the operating margin included in the analysis for various types of debris and exposed aluminum surface areas.

                               - - Descr1p" fion of O1peraf In!

T a bl e 4 2 M argm Actual Item Value Used Operating Margin Value Epoxy Unqualified 2700.6 lbm 2,729 lbm 28.4 lbm Coatings Alkyd Unqualified 30.6 lbm 591bm 28.4 lbm 0.753 ft 3 Coatings IOZ Unqualified 27.6 lbm 56 lbm 28.4 lbm Coatinos Latent Debris 60 lbm 2001bm 1401bm Miscellaneous 4 ft2 50 ft2 46 ft2 Debris Unsubmerged 741.3 ft 2 926.6 ft 2 185.3 ft2 Aluminum Metal Submerged 278.7 ft2 348.4 ft2 69.7 ft2 Aluminum Metal E4-21

4.0 References

1. Regulatory Guide 1.174, Revision 2, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," May 2011

2. Regulatory Guide 1.54, Revision 2, "Service Level I, II, and Ill Protective Coatings Applied to Nuclear Power Plants," October 2010

3. ANSI N101.4-1972, "Quality Assurance for Protective Coatings Applied to Nuclear Facilities," American National Standards Institute, Washington, DC

4. Regulatory Guide 1.97, Revision 2, "Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions during and Following an Accident," December 1980

5. EA-12-049, "Issuance of Order to Modify Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events," March 12, 2012

6. ASME Boiler and Pressure Vessel Code, Section XI, "Rules for lnservice Inspection of Nuclear Power Plant Components," 2010 Edition, July 1, 2010

7. ASME, Boiler and Pressure Vessel Code, Section Ill, "Rules for Construction of Nuclear Power Plant Components," 2010 Edition, July 1, 2010

8. Regulatory Guide 1.45, Revision 1, "Guidance on Monitoring and Responding to Reactor Coolant System Leakage," May 2008

9. Not used

10. NUREG/CR-7172, "Knowledge Base Report on Emergency Core Cooling Sump Performance in Operating Light Water Reactors," January 2014

11. NUREG/CR-6772, "GSl-191: Separate-Effects Characterization of Debris Transport in Water," August 2002 12.Not Used

13. Regulatory Guide 1.82, Revision 4, "Water Sources for Long-Term Recirculation Cooling Following a Loss-of-Coolant Accident," March 2012 14.WCAP-16793-NP-A, "Evaluation of Long-Term Cooling Considering Particulate, Fibrous and Chemical Debris in the Recirculating Fluid", Revision 2, July 2013.

E4-22

CAW-17~4565 AFFIDAVIT COMMONWEALTH OF PENNSYLVANIA: SS COUNTY OF BUTLER:

• I, James A. Gresham, am aµthorized to ex\:)cute tliis Affidavit onbehalfofWestinghouse Electric .
• Company LLC ("Westinghouse~') and. declare that the averments of fact set forth in this. Affidavit are true and correct to the best of my knowledge, information, arid belief.

Executed on: 1/l~f 1

                                                      *f J~e~ A. Gresham, Manager
                                                    .//      . .. .      .. .     .

Regulatory Compliance J

3 CAW-17-4565 (1) I am Manager, Regulatory Compliance, Westinghouse Electric Company LLC ("Westinghouse"), and as such, I have been specifically delegated the function of reviewing the proprietary information sought to be withheld from public disclosure in connection with nuclear power plant licensing and rule making proceedings, and am authorized to apply for its withholding on behalf of Westinghouse. (2) I am making this Affidavit in conformance with the provisions of 10 CFR Section 2.390 of the Nuclear Regulatory Commission's ("Commission's") regulations and in conjunction with the Westinghouse Application for Withholding Proprietary Information from Public Disclosure accompanying this Affidavit. (3) I have personal knowledge of the criteria and procedures utilized by Westinghouse in designating information as a trade secret, privileged or as confidential commercial or financial information. (4) Pursuant to the provisions of paragraph (b)(4) of Section 2.390 of the Commission's regulations, the following is furnished for consideration by the Commission in determining whether the information sought to be withheld from public disclosure should be withheld. (i) The information sought to be withheld from public disclosure is owned and has been held in confidence by Westinghouse. (ii) The information is of a type customarily held in confidence by Westinghouse and not customarily disclosed to the public. Westinghouse has a rational basis for determining the types of information customarily held in confidence by it and, in that connection, utilizes a system to determine when and whether to hold certain types of information in confidence. The application of that system and the substance of that system constitute Westinghouse policy and provide the rational basis required. Under that system, information is held in confidence if it falls in one or more of several types, the release of which might result in the loss of an existing or potential competitive advantage, as follows: (a) The information reveals the distinguishing aspects of a process (or component, structure, tool, method, etc.) where prevention of its use by any of L

4 CAW-17-4565 Westinghouse's competitors without license from Westinghouse constitutes a competitive economic advantage over other companies. (b) It consists of supporting data, including test data, relative to a process (or component, structure, tool, method, etc.), the application of which data secures a competitive economic advantage, e.g., by optimization or improved marketability. (c) Its use by a competitor would reduce his expenditure ofresources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing a similar product. ( d) It reveals cost or price information, production capacities, budget levels, or commercial strategies of Westinghouse, its customers or suppliers. (e) It reveals aspects of past, present, or future Westinghouse or customer funded development plans and programs of potential commercial value to Westinghouse. (f) It contains patentable ideas, for which patent protection may be desirable. (iii) There are sound policy reasons behind the Westinghouse system which include the following: (a) The use of such information by Westinghouse gives Westinghouse a competitive advantage over its competitors. It is, therefore, withheld from disclosure to protect the Westinghouse competitive position. (b) It is information that is marketable in many ways. The extent to which such information is available to competitors diminishes the Westinghouse ability to sell products and services involving the use of the information. (c) Use by our competitor would put Westinghouse at a competitive disadvantage by reducing his expenditure of resources at our expense.

5 CAW-17-4565 (d) Each component of proprietary information pertinent to a particular competitive advantage is potentially as valuable as the total competitive advantage. If competitors acquire components of proprietary information, any one component may be the key to the entire puzzle, thereby depriving Westinghouse of a competitive advantage. (e) Unrestricted disclosure would jeopardize the position of prominence of Westinghouse in the world market, and thereby give a market advantage to the competition of those countries. (f) The Westinghouse capacity to invest corporate assets in research and development depends upon the success in obtaining and maintaining a competitive advantage. (iv) The information is being transmitted to the Commission in confidence and, under the provisions of 10 CFR Section 2.390, is to be received in confidence by the Commission. (v) The information sought to be protected is not available in public sources or available information has not been previously employed in the same original manner or method to the best of our knowledge and belief. (vi) The proprietary information sought to be withheld in this submittal is that which is appropriately marked in "Vogtle Electric Generating Plant Supplemental Response to NRC Generic Letter 2004-02, Enclosure 2, 'Supplemental Response to NRC Generic Letter 2004-02' "(Proprietary), for submittal to the Commission, being transmitted by Letter GP-19572. The proprietary information as submitted by Westinghouse is that associated with resolution of and response to NRC Generic Letter 2004-02 and may be used only for that purpose. (a) This information is part of that which will enable Westinghouse to provide commercial support for resolution of and response to NRC Generic Letter 2004-02.

6 CAW-17-4565 (b) Further this information has substantial commercial value as follows: (i) Westinghouse plans to sell the use of similar information to its customers for the purpose of providing support for resolution of and response to NRC Generic Letter 2004-02. (ii) Westinghouse can sell support and defense of industry guidelines and acceptance criteria for plant-specific applications. (iii) The information requested to be withheld reveals the distinguishing aspects ofa methodology which was developed by Westinghouse. Public disclosure of this proprietary information is likely to cause substantial harm to the competitive position of Westinghouse because it would enhance the ability of competitors to provide similar technical evaluation justifications and licensing defense services for commercial power reactors without commensurate expenses. Also, public disclosure of the information would enable others to use the information to meet NRC requirements for licensing documentation without purchasing the right to use the information. The development of the technology described in part by the information is the result of applying the results of many years of experience in an intensive Westinghouse effort and the expenditure of a considerable sum of money. In order for competitors of Westinghouse to duplicate this information, similar technical programs would have to be performed and a significant manpower effort, having the requisite talent and experience, would have to be expended. Further the deponent sayeth not.

Proprietary Information Notice Transmitted herewith are proprietary and non-proprietary versions of a document, furnished to the NRC associated with resolution of and response to NRC Generic Letter 2004-02 and may be used only for that purpose. In order to conform to the requirements of 10 CFR 2.390 of the Commission's regulations concerning the protection of proprietary information so submitted to the NRC, the information which is proprietary in the proprietary versions is contained within brackets, and where the proprietary information has been deleted in the non-proprietary versions, only the brackets remain (the information that was contained within the brackets in the proprietary versions having been deleted). The justification for claiming the information so designated as proprietary is indicated in both versions by means oflower case letters (a) through (f) located as a superscript immediately following the brackets enclosing each item of information being identified as proprietary or in the margin opposite such information. These lower case letters refer to the types of information Westinghouse customarily holds in confidence identified in Sections (4)(ii)(a) through (4)(ii)(f) of the Affidavit accompanying this transmittal pursuant to 10 CFR 2.390(b)(l). Copyright Notice The reports transmitted herewith each bear a Westinghouse copyright notice. The NRC is permitted to make the number of copies of the information contained in these reports which are necessary for its internal use in connection with generic and plant-specific reviews and approvals as well as the issuance, denial, amendment, transfer, renewal, modification, suspension, revocation, or violation of a license, permit, order, or regulation subject to the requirements of 10 CFR 2.390 regarding restrictions on public disclosure to the extent such information has been identified as proprietary by Westinghouse, copyright protection notwithstanding. With respect to the non-proprietary versions of these reports, the NRC is permitted to make the number of copies beyond those necessary for its internal use which are necessary in order to have one copy available for public viewing in the appropriate docket files in the public document room in Washington, DC and in local public document rooms as may be required by NRC regulations if the number of copies submitted is insufficient for this purpose. Copies made by the NRC must include the copyright notice in all instances and the proprietary notice ifthe original was identified as proprietary. L_

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Vogtle Electric Generating Plant Supplemental Response to NRC Generic Letter 2004-02 Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Attachment 3 ALION Proprietary Information with Affidavit E5:A3-1

 .ALION SCIEN'C£ AND lECHNOLOGY ALION Science & Technology AFFIDAVIT We, Andy Roudenko, Project Manager and Martin Rozboril, Jr. Assistant Vice President Division Manager (AVPDM) state as follows:

(1) We, Andy Roudenko, Project Manager, and Martin Rozboril, Jr. AVPDM, Nuclear Services, ALION Science & Technology ("Alion") and have been delegated the function of reviewing the information described in paragraph (2) which is sought to be withheld, and have been authorized to apply for its withholding. (2) The infonnation sought to be withheld is contained in all revisions of ALION Science & Technology report "Erosion Testing of Small Pieces of Low Density Fiberglass Debris-Test Report," ALION-REP-ALION-1006-04, with the latest revision to date, Rev. 1, dated November 17, 20011. Information from this report was used to support analysis of post-LOCA debris transport in work designed to address GSI-191, Assessment of Debris Accumulation on PWR Sump Performance, issues at Southern Nuclear Operating Company, Vogtle Units 1 and 2. Specifically, the following Sections and Figures are to be withheld, on that basis that these unique attribute of the testing approach, test results and conclusions:

• Background
• Figure 1.1.1
• Figure 2.1.2
• Figure 2.1.3
• Figure 2.1.5
• Figure 2.1.6
• Figure 2.1.9
• Test Results, including Figures and Tables
• Data Analysis, including Figures and Tables
• Conclusions
• Appendices (3) In making this application for withholding of proprietary information of which it is the owner, Alion relies upon the exemption from disclosure set forth in the Freedom of Information Act ("FOIA"), 5 USC Sec. 552(b)(4), and the Trade Secrets Act, 18 USC Sec. 1905, and NRC regulations 10 CFR 9.17(a)(4), and 2.390(a)(4) for "trade secrets" (Exemption 4). The material for which exemption from disclosure is here sought also qualify under the narrower definition of "trade secret", within the meanings assigned to those terms for purposes of FOIA Exemption 4 in, respectively, Critical Mass Energy Project v. Nuclear Regulatory Commission, 975F2d87 l (DC Cir. 1992), and Public Citizen Health Research Group v. FDA, 704F2d1280 (DC Cir. 1983).

Page I of3 MAS Affidavit

AL I 0 N SCIENCE AtiD TECHNOLOGY (4) Some examples of categories of information which fit into the definition of proprietary information are:

a. Information that discloses a process, method, or apparatus, including supporting data and analyses, where prevention of its use by Alion's competitors without license from Alion constitutes a competitive economic advantage over other companies

b. Information which, if used by a competitor, would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing of a similar product;

c. Information which reveals aspects of past, present, or future Alion customer-funded development plans and programs, resulting in potential products to Alion;

d. Information which discloses patentable subject matter for which it may be desirable to obtain patent protection.

The information sought to be withheld is considered to be proprietary for the reasons set forth in paragraphs (4) a, and (4) b, above. (5) To address 10 CFR 2.390 (b) (4), the information sought to be withheld is being submitted to NRC in confidence. The information is of a sort customarily held in confidence by Alion, and is in fact so held. The information sought to be withheld has, to the best of my knowledge and belief, consistently been held in confidence by Alion, no public disclosure has been made, and it is not available in public sources. All disclosures to third parties including any required transmittals to NRC, have been made, or must be made, pursuant to regulatory provisions or proprietary agreements which provide for maintenance of the information in confidence. Its initial designation as proprietary information, and the subsequent steps taken to prevent its unauthorized disclosure, are as set forth in paragraphs (6) and (7) following. (6) Initial approval of proprietary treatment of a document is made by the manager of the originating component, the person most likely to be acquainted with the value and sensitivity of the information in relation to industry knowledge. Access to such documents within Alion is limited on a "need to know" basis. (7) The procedure for approval of external release of such a document typically requires review by the staff manager, project manager, principal scientist or other equivalent authority, by the manager of the cognizant marketing function (or their delegate), and by the Legal Operation, for technical content, competitive effect, and determination of the accuracy of the proprietary designation. Disclosures outside Alion are limited to regulatory bodies, customers, and potential customers, and their agents, suppliers, and licensees, and others with a legitimate need for the information, and then only in accordance with appropriate regulatory provisions or proprietary agreements. (8) The document identified in paragraph (2), above, is classified as proprietary because it contains "know-how" and "unique data" developed by Alion within our research and Page 2 of3 MAS Affidavit

AL I 0 N SCIENCE. AND TECHNOl.OGY development programs. The development of this document, supporting methods and data constitutes a major Al ion asset in this current market. (9) Public disclosure of the information sought to be withheld is likely to cause substantial harm to Al ion's competitive position and foreclose or reduce the availability of profit-making opportunities. The information is part of Alion's comprehensive BWR/PWR GSI-191 analysis base, and its commercial value extends beyond the original development cost. The value of the technology base goes beyond the extensive physical database and experimental methodology and includes development of the expertise to determine and apply the appropriate evaluation process. The research, development, engineering, analytical and experimental costs comprise a substantial investment of time and money by Alion. The precise value of the expertise to devise an evaluation process and apply the correct analytical methodology is difficult to quantify, but it clearly is substantial. Alion's competitive advantage will be lost if its competitors are able to use the results of the Alion experience to normalize or verify their own process or if they are able to claim an equivalent understanding by demonstrating that they can arrive at the same or similar conclusions. The value of this information to Alion would be lost if the information were disclosed to the public. Making such information available to competitors without their having been required to undertake a similar expenditure of resources would unfairly provide competitors with a windfall, and deprive Alion of the opportunity to exercise its competitive advantage to seek an adequate return on its large investment in developing these very valuable analytical tools. I declare under penalty of perjury that the foregoing affidavit and the matters stated therein are true and correct to the best of my knowledge, information, and belief. Executed on this 20th day of April 2017. Martin rif Digitally signed by Andy Roudenko Rozboril, Jr.

               ~I DN: cn=Andy Roudenko, o=Alion A- r K II t,~ Scie~ce an.d_T.echnology, ou=Nuclear f ~Services D1v1s1on,                                                       2017.04.20
                            ."       e1fuail=aroudenko@a1ionscience.com,
                          ;' r   _,.7=uS -'-:~                                                             15:06:17 -06'00'
                    .~ j?            Date: 2017.04.20 12:09:50 -07'00' Andy Roudenko                                                                     Martin Rozboril, Jr.

Project Manager Assistant Vice President ALION Science & Technology Division Manager, Nuclear Services ALION Science & Technology Page 3 of3 MAS Affidavit

Vogtle Electric Generating Plant - Units 1 & 2 Supplemental Response to NRC Generic Letter 2004-02 Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary)

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Table of Contents 1.0 Overall Compliance 2.0 General Description of and Schedule for Corrective Actions 3.0 Specific Information Regarding Methodology for Demonstrating Compliance 3.a Break Selection 3.b Debris Generation/Zone of Influence 3.c Debris Characteristics 3.d Latent Debris 3.e Debris Transport 3.f Head Loss and Vortexing 3.g Net Positive Suction Head 3.h Coating Evaluation 3.i Debris Source Term 3.j Screen Modification Package 3.k Sump Structural Analysis 3.1 Upstream Effects 3.m Downstream Effects - Components and System 3.n Downstream Effects - Fuel and Vessel 3.o Chemical Effects 3.p Licensing Basis 4.0 NRC Requests for Additional Information 5.0 References Attachments ES:A 1 General Electric Hitachi (GEH) Proprietary Information Affidavit E5:A2 Westinghouse Electric Corporation (WEC) Proprietary Information Affidavit E5:A3 ALION Proprietary Information Affidavit E5-1

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) 1.0 Overall Compliance: Provide information requested in GL 2004-02 Requested Information Item 2(a) regarding compliance with regulations. GL2004-02 Requested Information Item 2(a) Confirmation that the ECCS and CSS recirculation functions under debris loading conditions are or will be in compliance with regulatory requirements listed in the Applicable Regulatory Requirements section of this GL. This submittal should address the configuration of the plant that will exist once all modifications required for regulatory compliance have been made and this licensing basis has been updated to reflect the results of the analysis described above. Response to 1.0: This submittal proposes a risk-informed methodology for determining the design requirements to address the effects of loss-of-coolant accident (LOCA)-generated debris on emergency core cooling system (ECCS) and containment spray system (CSS) recirculation functions. The risk-informed analysis covers a full spectrum of postulated LOCAs, including double-ended guillotine breaks (DEGBs), for all pipe sizes up to and including the design-basis accident (OBA) LOCA, to provide assurance that the most severe postulated LOCAs are evaluated. Vogtle Electric Generating Plant (VEGP) conservatively relegates to failure the individual breaks that can generate and transport debris that are not bounded by VEGP analyzed limits. The results of the evaluation in Enclosure 3 show that the risk from the failures related to LOCA-generated debris is "very small" as the risk falls in Region Ill of RG 1.174. The methodology includes conservatisms in the plant-specific testing and in the assumption that all breaks that exceed the tested debris quantities are relegated to failure. Conservatisms in the VEGP approach are discussed in Enclosure 4, Defense-in-Depth and Safety Margin. Southern Nuclear Operating Company (SNC) is utilizing a risk-informed approach to the effects of LOCA debris for VEGP. The risk-informed approach replaces the existing deterministic approach described in the VEGP licensing basis and consequently requires an amendment to the VEGP Units 1 and 2 operating licenses to incorporate the revised methodology per the requirements of Title 10 of the Code of Federal Regulations (CFR) Section 50.59 (10 CFR 50.59). The proposed amendment to the operating license will be described in the future license amendment request (LAR). Exemptions to the overall requirements associated with 10 CFR 50.46(a)(1 ), GDC 35, GDC 38, and GDC 41 are required due to the change in methodology. The requests for exemption will be provided in the future LAR. In addition, SNC proposes to amend the VEGP Unit 1 and Unit 2 operating licenses to revise the Technical Specifications (TSs) for the ECCS and CSS. The proposed TS changes detailed in the future LAR will align the TSs with the risk-informed E5-2

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) methodology change. The licensing discussion is continued in the Response to 3.p of this enclosure. In the resolution of Generic Safety Issue (GSl)-191, "Assessment of Debris Accumulation on Pressurized Water Reactor (PWR) Sump Performance," VEGP implemented (or will implement) the following changes:

• The refueling water storage tank (RWST) High level was increased and the Low-Low level (initiation of semi-automatic switchover to recirculation) decreased to provide increased submergence of the sump strainers while maintaining adequate net positive suction head (NPSH) for the ECCS and containment spray (CS) pumps; allowing sufficient time for completion of operator actions for switchover to recirculation.
• To improve existing margins until all corrective actions can be implemented, VEGP installed larger sump strainers that increased the available screen area for each of the residual heat removal (RHR) strainers and the CS strainers. The hole diameters of the strainer perforated plates were reduced to lessen the potential for debris passing through the strainer and causing plugging and/or wear of the downstream ECCS and CS piping and equipment, and reactor vessel. In addition, Min-K insulation was removed from the containment bioshield area.
• Orifices were installed in the intermediate- and high-head ECCS lines, and the associated throttle valves were adjusted to operate at a minimum internal clearance greater than the size of debris that could pass through the strainers.

The opening size was increased to ensure that adequate clearance in the valves will prevent debris from causing excessive wear or plugging.

• Procedural and program controls are in place to ensure materials used in the containments will not result in an increase of the debris loading beyond the analyzed values. This includes controls for containment coatings, labels, and insulation.
• Extensive analysis has been performed in accordance with Nuclear Energy Institute (NEI) 04-07 guidance (Reference 2), the associated United States Nuclear Regulatory Commission (NRC) safety evaluation (SE) (Reference 3),

and other industry documents reviewed by the NRC. With few exceptions, VEGP has followed this guidance. Technical justification is available and provided for the few cases where other approaches were utilized.

• The emergency operating procedures are being revised to delay operator action to isolate the RHR pumps from the RWST. This ensures that water level in the RWST is drawn down to the Empty alarm level for all scenarios (it previously only reached the Empty alarm level for scenarios that actuated CS) and prevents most scenarios from resulting in partially submerged strainers.

E5-3

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary)

• To ensure full submergence for an increased number of postulated break scenarios, the Uhit 1 and Unit 2 RHR strainers require a reduction in height. The Unit 1 and Unit 2 design packages have been prepared.

Correspondence Background The following discussion contains correspondences issued by or submitted to the NRG prior to December 31, 2007, on the subject of GSl-191. The title of each letter is provided in the reference section of this enclosure. The NRG issued Bulletin 2003-01 on June 9, 2003 (Reference 5), asking for a 60-day response providing a description of any interim compensatory measures that have been implemented, or that will be implemented, to reduce the risk which may be associated with potentially degraded or nonconforming EGGS and GSS recirculation functions until an evaluation to determine compliance is complete. SNG provided the 60-day response in a letter dated August 7, 2003 (Reference 6). Supplemental letters dated October 29, 2004 (Reference 7), and July 22, 2005 (Reference 8), were provided by SNG in response to requests for additional information. The NRG issued Generic Letter (GL) 2004-02 on September 13, 2004 (Reference 1), requesting an initial 90-day response, a 12-month response, and for the guidance of the GL to be met by December 31, 2007. In December 2004, NEI issued NEI 04-07 (Reference 2) providing an evaluation methodology for the industry. The NRG provided the associated SE (Reference 3) on December 6, 2004. The NRG had already issued RG 1.82 Revision 3 (Reference 25) in November 2003. SNG provided the initial response for VEGP in a letter dated February 25, 2005 (Reference 10). SNG provided a follow-up response on August 31, 2005 (Reference 11 ), providing more details on how SNG would meet the GL 2004-02 requirements. The NRG issued a request for additional information on February 9, 2006 (Reference 12), with a 60-day response time. NEI worked with the NRG and recognized that much of the information needed to address the RAls would not be available until ongoing testing activities were completed. The NRG-issued letter dated March 28, 2006 (Reference 13), identified that the Request for Additional Information (RAI) answers could be provided as part of the supplemental response by the end of December 2007. An NRG letter dated January 4, 2007 (Reference 18), provided clarification that even if a licensee had an extension for modifications past 2007, the supplemental response was still due by December 31, 2007.

• SNG submitted an extension request in a letter dated June 22, 2006 (Reference 14),

for modification/installation of the Unit 1 EGGS flow orifices and for chemical effects testing. In a teleconference on June 30, 2006, with the NRG staff reviewer of the E5-4

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) June 22, 2006, extension request, SNC was asked to provide an update of on-going activities and a clarification as to what activities are driving the extension request. SNC provided the requested information in a response dated July 28, 2006 (Reference 15), which also requested an extension from December 31, 2007, to the spring 2008 outage. This extension request was approved in an NRC letter dated September 7, 2006 (Reference 16).

• The NRC issued a letter dated August 15, 2007, containing the content guide for the GL 2004-02 supplemental response due in December 2007. Additional information was provided by the NRC in a letter dated September 27, 2007, for chemical effects, protective coatings, and head loss testing. A revision to the content guide was issued by the NRC in a letter dated November 21, 2007 (Reference 108). The due date for the supplemental response was extended by an NRC letter dated November 30, 2007 to allow the supplemental response to be submitted by February 29, 2008.

An NRC letter dated November 8, 2007, provided guidance for requesting plant-specific extensions. Additional information was also provided in an NRC letter dated November 13, 2007, on how GSl-191 would be closed and how the closure would be documented for each site. SNC submitted a letter dated December 7, 2007 (Reference 19), requesting an extension for submittal of chemical effects testing results, downstream effects - components and systems, and downstream effects - fuel and vessel until June 30, 2008. This request was approved in an NRC letter dated December 19, 2007 (Reference 20). E5-5

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) 2.0 General Description of and Schedule for Correction Actions: Provide a general description of actions taken or planned, and dates for each. For actions planned beyond December 31, 2007, reference approved extension requests or explain how regulatory requirements will be met as per Requested Information Item 2(b). (Note: All requests for extension should be submitted to the NRC as soon as the need becomes clear, preferably no later than October 1, 2007.) GL 2004-02 Requested Information Item 2(b) A general description and implementation schedule for all corrective actions, including any plant modifications that you identify while responding to this generic letter. Efforts to implement the identified actions should be initiated no later than the first refueling outage starting after April 1, 2006. All actions should be completed by December 31, 2007. Provide justification for not implementing the identified actions during the first refueling outage starting after April 1, 2006. If corrective actions will not be completed by December 31, 2007, describe how the regulatory requirements discussed in the Applicable Regulatory Requirements section will be met until the corrective actions are completed. Response to 2.0: SNC has performed analysis to determine the susceptibility of the ECCS and CSS recirculation functions for VEGP to the adverse effects of post-accident debris blockage and operation with debris-laden fluids. These analyses conform, to the greatest extent practicable, to the NEI 04-07 methodology (Reference 2) as approved by the NRC SE dated December 6, 2004 (Reference 3). As of April 24, 2017, SNC has completed the following GL 2004-02 (Reference 1) actions, analyses, and modifications:

• Replaced Unit 1 and Unit 2 containment emergency sump screens during refueling outage 1R13 (Fall 2006), and refueling outage 2R12 (Spring 2007),

respectively

• Installed ECCS flow orifices in the intermediate and high-head ECCS lines that allow the ECCS throttle valves to be opened greater than the maximum expected strainer bypass debris size while maintaining the capability to ensure ECCS flow balance, mitigating downstream effects (2008)
• Completed inspection of containment per NEI 02-01 (Reference 41 ), "Condition Assessment Guidelines: Debris Sources Inside PWR Containment"
• Performed latent debris sampling and characterization ,
• Participated in the PWR Owners Group (PWROG) program to evaluate downstream effects related to in-vessel long-term cooling with results documented in WCAP-16793-NP-A (Reference 22)
• Removed Min-K insulation in the original zone of influence (ZOI) analyzed for GL 2004-02 from VEGP's containments based on preliminary head loss testing (without chemical effects) which determined that the removal of Min-K insulation resulted in a significant reduction in head loss across a debris-laden strainer E5-6

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary)

• Implemented programmatic and procedural changes to maintain acceptable configuration and to protect the newly established design and licensing basis
• Developed containment 30 CAD model of VEGP Unit 1 containment to include pipe welds, for both VEGP containments because Units 1 and 2 are virtually identical (CAD model of VEGP Unit 1 containment was used to determine a correlation between the containment pool volume and containment pool level)
• Completed detailed laser scans of the VEGP containments, which provide measurements for contingency insulation replacement for Units 1 and 2 (laser scans of both units were completed before January 1, 2013)
• Developed detailed debris generation and debris transport analyses and a computational fluid dynamics (CFO) model
• Developed a hydraulic model of the ECCS
• Performed detailed CS and RHR NPSH analysis
• Performed water level analysis
• Modified probabilistic risk assessment (PRA) to include strainer and core blockage events
• Quantified chemical precipitants using WCAP-16530 (with refinements)
• Performed chemical effects testing
• Completed RELAPS-30/MELCOR modeling similar to South Texas Project (STP) model (however, the results are not used as input for the base case analysis)
• Performed strainer head loss and fiber debris penetration testing
• Participated in the PWROG Comprehensive Analysis and Test Program for GSl-191 Closure
• Performed downstream wear and blockage analysis to WCAP-16406-P-A, Revision 1 (Reference 21)
• Performed detailed structural analysis of strainers
• Assembled base case final inputs for quantifying the conditional failure probabilities related to GSl-191 using the software package Nuclear Accident Risk-Weighted Analysis (NARWHAL) (see Enclosure 3, Section 13.1 for general description of the software).
• Completed NARWHAL sensitivity analyses
• Integrated NARWHAL results into VEGP PRA model to determine changes in core damage frequency (LiCDF) and changes in large early release frequency (LiLERF)
• Revised operating procedures to ensure that the RHR strainers are completely submerged for an increased number of postulated LOCA scenarios (operator action to isolate the RWST from the RHR pumps was delayed until the Empty level is reached to ensure sufficient injection of RWST water for breaks that activate CS)

The following risk-informed resolution path activities are planned by SNC to address GL 2004-02 and support closure of GSl-191 for VEGP.

• SNC is planning to modify the VEGP Unit 1 and Unit 2 RHR sump strainers to reduce the overall height by removing the top two strainer disks per stack from each of the RHR strainer assemblies.

E5-7

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary)

• SNC will submit an LAR for a risk-informed resolution to GL 2004-02 for VEGP within six months after receipt of the SE for WCAP-17788-P.
• SNC will submit any necessary revisions to the supplemental response to support closure of GL 2004-02 for VEGP Units 1 and 2 within six months after receipt of the SE for WCAP-17788-P.

Correspondence Background The following discussion contains correspondences issued by or submitted to the NRC beyond December 31, 2007, on the subject of GSl-191. The correspondences document VEGP's compliance with regulatory requirements per Requested Information Item 2(b) and include reference to approved extension requests. The title of each letter is provided in the reference section of this enclosure. SNC letter dated February 28, 2008, NL-07-1777 (Reference 95), provided SNC's supplemental response to GL 2004-02 for VEGP. SNC letter dated May 21, 2008, NL-08-0670 (Reference 96), provided a revised transmittal of SNC's supplemental response to GL 2004-02 for VEGP based on NRC's questions regarding the proprietary nature of information provided in SNC letter dated February 28, 2008 (Reference 95). SNC letter dated May 22, 2008, NL-08-0818 (Reference 97), requested an extension for the final response to GL 2004-02 for the completion of WCAP-16406-P-A and WCAP-16793-NP-A evaluations, and chemical effects testing and evaluation of test results. An extension was granted to SNC by the NRC to July 31, 2008, in a letter dated May 29, 2008, as stated in NL-08-1155 (Reference 98). SNC letters dated July 31, 2008, NL-08-1155 (Reference 98) and NL-08-1195 (Reference 99), provided the downstream effects results for components and in-vessel analyses and requested an extension for the GL 2004-02 supplemental response for chemical effects, respectively. SNC letter dated August 22, 2008, NL-08-1228 (Reference 100), provided the GL 2004-02 response for chemical effects. The letter also contained a revised answer to question 3.g.15 originally submitted in SNC letter dated May 21, 2008 (Reference 96). NRC letter dated September 17, 2008, NL-08-1497 (Reference 101), provided RAls from a partial review of prior SNC responses to GL 2004-02 pertaining to the reliance on results from testing at the VUEZ facility by Alien Science and Technology. As discussed with Mr. Jared S. Wermiel, Deputy Director of the Division for Engineering and Safety Systems, in a telephone call with SNC on September 11, 2008, the NRC identified several critical issues with the test protocol used in the testing at VUEZ. The NRC staff has stated that based on their review of information provided by Alien on the VUEZ testing, it is highly unlikely that SNC's ES-8

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) reliance on the VUEZ testing, performed to date to demonstrate strainer adequacy, will provide an adequate technical basis to resolve GL 2004-02. SNC letter dated November 7, 2008, NL-08-1583 (Reference 102), requested an extension, in accordance with SECY-06-0078, for completion of chemical effects testing and closeout activities for GL 2004-02 as a result of NRC's concern that SNC's reliance on the VUEZ performed testing would not provide an adequate technical basis to resolve GL 2004-02 for VEGP. NL-08-1583 also noted that the RAls issued by the staff on September 17, 2009 were from a partial review of SNC's responses to GL 2004-02 and did not represent a comprehensive set of RAls. In addition, SNC submitted milestone dates supporting a closeout of GSl-191 and a final response to the staff by November 20, 2009, predicated on a reasonable submittal of SNC test-for-success protocol with NRG review and comment cycle, and resolution of any issues associated with pending revision ofWCAP-16793-NP. NRG letter dated December 2, 2008, NL-08-1829 (Reference 87), provided RAls for prior SNC supplemental responses, letters dated February 28, 2008; May 21, 2008; July 31, 2008; and August 22, 2008 (References 95, 96, 98, and 100, respectively). The NRG requested RAI responses within 90 days. SNC letter dated February 10, 2009, NL-09-0159 (Reference 103), notified the NRG that a single response to the RAls issued by the NRG letter dated December 2, 2008, would be submitted once the new testing and analysis discussed in the November 7, 2008, letter was completed. SNC also notified the NRG that the planned completion date for the GL 2004-02 response was November 20, 2009, excluding issues related to WCAP-16793-NP. SNC letter dated November 19, 2009, NL-09-1839 (Reference 104), stated that the schedule for completion of GSl-191 activities for VEGP is contingent upon resolution of generic issues and their impact to the remaining 26 RAls for VEGP. Of the 29 RAls issued December 2, 2008, SNC discussed 26 satisfactorily with the staff during public telecoms between SNC and the staff on August 13, 2009, and October 13, 2009. The three remaining RAls concerned the following generic issues: Nukon ZOI, fibrous insulation erosion, and in-vessel downstream effects. During the telecom between SNC and the staff on October 13, 2009, the above generic issues were discussed as to how the outcome of said issues would impact the 26 RAls for VEGP. It was agreed to by the staff in this telecom that a written response to the resolved RAls was not required by November 20, 2009. NEI letter to NRG dated May 4, 2012 (Reference 91 ), highlighted the current industry status and recommended actions for closure of GSl-191 based on licensees providing a docketed submittal to the NRG by December 31, 2012, outlining a GSl-191 resolution path and schedule pursuant to the Commission direction in Staff Requirements Memorandum (SRM) SECY-10-0113 (Reference 90). E5-9

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) An NEI letter to the NRC dated November 15, 2012 (Reference 92), and a subsequent NRC review of the schedule (Reference 93) recommended that licensees delay submittal of GSl-191 resolution path and schedule until January 31, 2013. The letter also recommended an alternative option to submit 30 days following placement of both the Commission's response to SECY-12-0093 (Reference 70) and the NRC SE on WCAP-16793-NP (Reference 94) into the public record. The Commission approved the staff's recommendation in SRM-SECY-12-0093 (Reference 70) dated December 14, 2012, to allow licensees the flexibility to choose any of the three options discussed in the paper to resolve GSl-191. Further, the Commission encouraged the staff to remain open to staggering licensee submittals and the associated NRC reviews to accommodate the availability of staff and licensee resources. The SE forWCAP-16793-NP (Reference 94) was made publicly available by the NRC on April 16, 2013. SNC Letter dated May 16, 2013, NL-13-0953 (Reference 105), transmitted the VEGP Units 1 and 2 resolution path forward and schedule for resolution, summary of actions completed for GL 2004-02, and defense-in-depth and mitigation measures, using the industry template developed by NEI. VEGP provided a basis for continued operation in the interim while the industry and NRC collaborated on how to proceed towards resolution of the issue. VEGP is following the "STP Piloted Risk-Informed Approach for GSl-191," as submitted by the following South Texas Project Nuclear Operating Company (STPNOC) letters to the NRC (Reference 105). STPNOC letter to the NRC dated November 13, 2013, NOC-AE-13003043 (Reference 44), submitted Supplement 1 to the STP pilot risk-informed approach for GSl-191. STPNOC letter to the NRC dated August 20, 2015, NOC-AE-15003241 (Reference 45), submitted Supplement 2 to the STP pilot submittal. STPNOC letter to the NRC dated October 20, 2016, NOC-AE-16003401 (Reference 109), submitted Supplement 3 to the Revised STP pilot submittal. E5-10

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) 3.0 Specific Information Regarding Methodology for Demonstrating Compliance:

a. Break Selection The objective of the break selection process is to identify the break size and location that present the greatest challenge to post-accident sump performance.

1. Describe and provide the basis for the break selection criteria used in the evaluation.

Response to 3.a.1: The VEGP debris generation calculation followed the methodology of NEI 04-07 and associated NRG SE (References 2 and 3, respectively), with the exception that it analyzed a full range of breaks, not just the worst-case breaks as suggested by NEI 04-07. The purpose of the calculation is to obtain debris quantities for the range of possible break scenarios. The calculation evaluated debris generation quantities for breaks on every inservice inspection (ISi) weld within the Class 1 pressure boundary. Both DEGBs and partial breaks were considered. All break sizes analyzed are assumed to fall into one of three high-level categories: small-break LOCA (SBLOCA) - a break smaller than 2 inches, medium-break LOCA (MBLOCA)

   - a break greater than or equal to 2 inches and less than 6 inches, or large-break LOCA (LBLOCA) - a break greater than or equal to 6 inches with the largest break being a DEGB of the 31-inch crossover leg.

In the debris generation calculation, a three-dimensional CAD model of VEGP Unit 1 containment building was updated to work with ENERCON's Break Analysis Debris Generator (BADGER) software. Note that the Unit 1 containment was used to represent both containments because the VEGP units are virtually identical. BADGER was used to place ZOls representing possible breaks on every ISi weld identified in containment. See Figure 3.a.1-1 for weld locations and Table 3-8 in Enclosure 3 for a complete list of welds inside the first isolation valve. In Figure 3.a.1-1, welds labeled as "In" are on the RCS side of the first isolation valve and welds labeled as "Out" are downstream of the first isolation valve. As discussed in Enclosure 1 Section 3.2, non-pipe LOCAs were not explicitly analyzed. E5-11

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Figure 3.a.1 Weld Locations where Postulated LOCAs Occur It should be noted that, while DEGBs on main loop piping are typically bounding with regard to the volume of debris generated, small partial breaks are much more likely to occur. A partial break is any break smaller than a DEGB (i .e., sidewall break) . Partial breaks are described by equivalent break size and are represented by a hemispherical ZOI with a radius proportional to the equivalent break size (see Figure 3.a.1-2). E5-12

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Figure 3.a.1 Single Partial Break Zone of Influence

2. State whether secondary line breaks were considered in the evaluation (e.g. , main steam and feedwater lines) and briefly explain why or why not.

Response to 3.a.2: Although the probability is low, a secondary side break (SSB) inside containment could require ECCS recirculation in a feed and bleed scenario. Therefore , secondary side breaks from the steam generator feedwater lines and main steam lines were analyzed . Because secondary side breaks occur at a lower pressure and temperature than the primary side breaks , the ZOI size corresponding to the insulation destruction pressure would be smaller. The appropriate ZOI sizes were calculated based on the ANSI jet methodology described in Appendix I of NEI 04-07 Volume 2 (Reference 3) . The main steam and feedwater pressures, temperatures , and calculated ZOI sizes are presented in Table 3.a .2-1. Breaks were postulated in increments of least every 5 ft along each of the main steam and feedwater pipes. All secondary side breaks were assumed to be DEGBs. Only Nukon insulation was considered for the secondary-side breaks because there is no fire barrier within the vicinity of the main steam and feedwater lines , and the coatings quantities would be bounded by the primary side breaks. E5-13

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Table 3.a.2 Secondary-Side Line ZOI Summary Main Steam Lines Feedwater Lines Pipe Inner Pipe Inner Po (psia) To Diameter Po (psia) To Diameter (oF) (oF) (inch) (inch-) 985 545 24.0 1150 445 12.8 Mass Flux (lbm*ft2 *s*1 ) CT Mass Flux (lbm*ft*2 *s*1 ) CT 1,978 1.25 19,238 1.65 Insulation I Insulation I Destruction Pressure ZOI Radius (D) Destruction Pressure ZOI Radius (D) (psia) (psia) CoatinQs 140 3.0 Coatings I 40 2.8 lnteram / 10.2 7.9 lnteram I 10.2 7.2 Nukon I 6 10.5 Nukon / 6 11.3 Note: Cr is the thrust coefficient

3. Discuss the basis for reaching the conclusion that the break size(s) and locations chosen present the greatest challenge to post-accident sump performance.

Response to 3.a.3: Debris generation quantities were evaluated for breaks on every ISi weld within the Class 1 pressure boundary. Both DEGBs and partial breaks were considered. The welds are sufficiently close, with sufficient overlap in the ZOls to provide confidence that the debris load that' presents the greatest challenge to post-accident sump performance has been captured. The total quantity of each type of debris generated within a particular ZOI is unique for every break scenario. Therefore, the bounding break-specific debris loads contained in the BADGER database were used on a break-specific basis for the analysis. The results of the debris generation calculation are presented below .. When reading the tables in this section it should be noted that the individual quantities for fines, small pieces, large pieces, and intact blankets do not necessarily add up to the total fiber quantity within the ZOI because the minimum, maximum, and average values for each size do not necessarily come from the same break. All average values are based on an equal probability of all breaks and do not consider differences in weld-specific break frequencies or the lower frequencies associated with larger break sizes. These results reflect the following conservatisms:

• All debris sources within the reactor cavity were assumed to be available for destruction by all breaks within the reactor cavity, despite the likelihood that ZOls would be restricted by structures and restraints.

E5-14

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary)

• All qualified coatings on steel and concrete were analyzed as having the worst-case coating system for each surface.
• Main loop breaks in the steam generator (SG) compartments were grouped by loop and truncated collectively in a way that could result in conservative amounts of debris generated for some breaks.

Debris Generated by DEGBs Table 3.a.3-1 shows the location of the worst-case break for each debris type. Table 3.a.3-2, Table 3.a.3-3, and Figures 3.a.3-1 through 3.a.3-8 show the minimum, average, and maximum debris quantities by debris type for DEGBs and partial breaks upstream of the first isolation valve at VEGP Units 1 and 2. Table 3.a.3 Location of Maximum Debris by Debris Type Debris Type Worst-Case Location Amount 3 Nukon (ft ) SG Compartment 1/4 Side 2229.2 lnteram E-50 Series (lbm) SG Compartment 1/4 Side 59.8* IOZ (lbm) SG Compartment 2/3 Side 65.3 Epoxy (lbm) Reactor Cavity 220.4

     *The limiting quantity of lnteram E-50 is generated by a partial break. This is possible because partial breaks are analyzed as being centered on the edge of the pipe, whereas DEGBs are centered on the axis of the pipe (see Figure 3.a.1-2). Because of this, partial breaks can extend further than DEGBs up to the outside radius of the pipe.

E5-15

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Table 3.a.3 Debris Generated by DEGBs Upstream of First Isolation Valve Debris Quantity Generated Small Breaks Medium Breaks Large Breaks Debris Type Debris Size ( < 2" ( 2"- 6") ( > 6" Min Avg Max Min Avg Max Min Avg Max Fines (Individual 0.0 0.1 0.8 0.0 2.9 12.2 1.9 56.0 289.3 Fibers) Small Pieces 0.0 0.4 2.8 0.0 9.4 40.8 6.4 187.1 999.5 (< 6" a side) Large Nukon (ff) Pieces 0.0 0.3 1.5 0.0 6.6 25.6 2.8 106.1 549.6 (> 6" a side) Intact (Covered) 0.0 0.3 1.6 0.0 7.1 27.7 3.1 114.6 594.0 Blankets All Debris 0.0 1.2 6.6 0.1 26.1 104.1 14.6 464.0 2229.2 Within ZOI Fire Barrier Fiber 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.4 12.1 Debris (lbm) 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.8 28.3 IOZ Qualified 0.0 <0.1 0.1 0.0 <0.1 1.3 0.0 9.9 65.3 Coatings (lbm) Particulate Epoxy Qualified 0.0 <0.1 0.3 0.0 0.4 4.9 0.0 44.8 220.4 Coatings (lbm) E5-16

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Range of Debris with Average by Pipe Size, DEGB 10" TO 16" (116 WELDS) 15 165 649 1 -- - - - - - - '

    ~

Vi Cll 6" T08"(198WELDS) 5 46 127

   .~

c.. 2.S" T04" (158WELDS) Q 9 64

               < 2" (354 WELDS)  Q    1     7 Debris Generated (ft 3 )

Figure 3.a.3 Range of Nukon Debris Generated by DEGBs Upstream of First Isolation Valve E5-17

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Range of Debris with Average by Pipe Size, DEGB MAIN LOOP (56 WELDS) 0 4 1--~--~~~~~~~~~~~~~~~~----~..._ ___ 40 10" TO 16" ( 116 WELDS) IJ 0

!l iii CIJ 6" TO 8" (198 WELDS) 0 0 0 Q.

ii: 2.5" TO 4" (158 WELDS) 0 0 0

              < 2" (354 WELDS)   0      0 0 Debris Generated (lb)

Figure 3.a.3 Range of Fire Barrier Debris Generated by DEGBs Upstream of First Isolation Valve Note that Figure 3.a.3-2 shows the total amount of fi re barrier destroyed (fiber plus particulate) . E5-18

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Range of Debris with Average by Pipe Size, DEGB MAINLOOP(S6WELDS) 37 143 220 r--~--__;;...-~~~~~~~~~~.--~~~~~:------ 10" TO 16" (116 WELDS) 0 3 21

   .~

VI QJ 6"T08" (198WELDS) Q 1 6 c.. ii: 2.S" T04 " (158WELDS) Q 0 1

               < 2" (354 WELDS)  Q     0    0 Debris Generated (lb)

Figure 3.a.3 Range of Epoxy Coatings Debris Generated by DEGBs Upstream of First Isolation Valve ES-19

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Range of Debris with Average by Pipe Size, DEGB

    ~

iii QI 6" TO 8" (198 WELDS) 0 0 2 a. ii: 2.S" T04 " (158WELDS) Q 0 1

             < 2" (3S4 WELDS) 0  0  0 Debris Generated (lb)

Figure 3.a.3 Range of IOZ Coatings Debris Generated by DEGBs Upstream of First Isolation Valve E5-20

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Table 3.a.3 Debris Generated by Partial Breaks Upstream of First Isolation Valve Debris Quantity Generated Small Breaks Medium Breaks Large Breaks Debris Type Debris Size ( < 2" ( 2" - 6") ( > 6") Min Avg Max Min Avg Max Min Avg Max Fines (Individual 0.0 <0.1 0.2 0.0 1.0 8.9 0.0 38.4 223.7 Fibers) Small Pieces 0.0 <0.1 0.8 0.0 3.0 28.5 0.2 128.6 794.1 (< 6" a side) Large Pieces Nukon (ft3) 0.0 <0.1 0.3 0.0 2.3 24.4 0.0 72 .0 438.4 (> 6" a side) Intact (Covered) 0.0 <0.1 0.3 0.0 2.5 26.4 0.0 77 .8 473.7 Blankets All Debris 0.0 0.1 1.6 0.0 8.8 79.5 0.2 317.0 1783.0 Within ZOI lnteram E-50 Fiber 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.2 17.9 Series (lbm) 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.4 41 .9 IOZ Qualified 0.0 <0.1 0.1 0.0 <0.1 2.1 0.0 7.4 40.9 Coatings (lbm) Particulate Epoxy Qualified 0.0 <0.1 0.4 0.0 0.2 3.9 0.0 30.7 149.1 Coatings (lbm) E5-21

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Range of Debris with Average by Break Size 31* ~6 74_ 161 30.5"

                -                                        --                                           154 30*
                -                                     _9-7                                          1492 29.s*
                -                                   67_                                          144 29*
                -                               6.18                                                              m1:783 28.5"
                -                                                                                                -7 28"   -                              8_                               I

_68 21.s*

                 -                      4'                                                                  _64 27*
                -         r

_489 1594 2&*

                 -        I                                                     l 25"
                 -        I                                                     I

_404 24* 86 131 QI N 23" .., f)4: n1 v; 22* Ir -128 I

   .¥     21*                 _:t_                                        '10l3 Ill          -

QI

    ~

20"

                 -                                                   9-14 cc     19*                                                 -2 ,..

1s* 5' 55 17" 5'

16. 613 ~

15" 14" _] 12* 10*

                -           7

_;~1 8' 6' 4' < 10 5 2* .-0 2 21 0.5"" 0 0 2 Debris Generated (ft3 ) Figure 3.a.3 Range of Nukon Debris Generated by Partial Breaks Upstream of First Isolation Valve E5-22

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Range of Debris with Average by Break Size 31" .. 30.5" **3 30" ~ I 29.5" 29" ..... 28.5" I .... I I I

!l 28" I

                                                           *9 v;
  ..:.:: 27.5" I        I                                      60 !

nl cc

    ....C1I   27" I        I                                    ~

26" n I I

• u 25" I I ..,

24" ~o .., I I 34 23* I I 32 22* I I 22 21* 00 1 I I Debris Generated (lb) Figure 3.a.3 Range of Fire Barrier Debris Generated by Partial Breaks Upstream of First Isolation Valve Note that Figure 3.a.3-6 shows the total amount of fire barrier destroyed (fiber plus particulate) . Note that no fire barrier material is destroyed by partial breaks smaller than 21 inches. E5-23

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Range of Debris with Average by Break Size 31" 14 30.5' 1 30'

                 --                                             --                                                 139 29.5' f"-ll !                                        --                                                _34 29'
                                                              -                                            l28 28.5'                                                                                         ~"23 28' 27.5"
                 --                             I

_'1.1

                                                           -                              1'1.2 27" 26'
                                                        --                            _-08
                                                    '1                            99 25" 24'                                                              _7 Ql N

23"

                 -                                                    n Vi      22"    -                      -                          70
  .:.:. 21"
                                      -                       6 I'll cc Ql    20" 19'

_3 58 18" - 8 17' 16' 9:. 15' _5 14"

                 -        --          _2 12"
                 -      -    2. !'8 10"  ~-.

8"  !:o:-J. 7 6" c()I 0 4 4" =o 0 2 2" 0 0 0 0.5" 0 0 0 Debris Generated (lb) Figure 3.a.3 Range of Epoxy Coatings Debris Generated by Partial Breaks Upstream of First Isolation Valve E5-24

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Range of Debris with Average by Break Size 31* 2S 41 30.S" 30" 8 29.5" 29" s 2s.s

• 3 I

2s* 32 27.S" 1 I 27" I 26. 2s*

                     -         I   --                                      _6

_8 I - 24" s QI 23" 2' N Vi 22" 23

       ~

Ill 21*

                     -                                         ll QI co 20*

19*

                     -                              18

_'() ~ 1s

• 16 17" §:1 16" 4 i s* 13 14" 2 12*

10* 8" "=O 6" r=o-o 2 4" lrO 0 1 2" 0 0 0 0.5 0 0 0 Debris Generated (lb) Figure 3.a.3 Range of IOZ Coatings Debris Generated by Partial Breaks Upstream of First Isolation Valve

b. Debris Generation/Zone of Influence (excluding coatings)

The objective of the debris generation/ZOI process is to determine , for each postulated break location : (1) the zone within which the break jet forces would be sufficient to damage materials and create debris ; (2) the amount of debris generated by the break jet forces .

1. Describe the methodology used to determine the ZOls for generating debris .

Identify which debris analyses used approved methodology default values . For debris with ZOls not defined in the guidance report/SE , or if using other tha n default values , discuss method(s) used to determine ZOI and the basis for each . Response to 3.b.1: For DEGBs , the ZOI is defined as a spherical volume about the break in which the jet pressure is higher than the destruction/damage pressure for a certain type of insulation , coatings , or other materials impacted by the break jet. E5-25

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) In a PWR reactor containment building, the worst-case pipe break would be a DEGB. In a DEGB, jets of water and steam would blow in opposite directions from the severed pipe. One or both jets could impact obstacles and be reflected in different directions. To take into account the double jets and potential jet reflections, NEI 04-07 (Reference 2) proposes using a spherical ZOI centered at the break location to determine the quantity of debris that could be generated by a given line break. For any break smaller than a DEGB (i.e., a partial break), NEI 04-07, Volume 2 (Reference 3) suggests a hemispherical ZOI centered at the edge of the pipe. Because these types of breaks could occur anywhere along the circumference of the pipe, the partial breaks were analyzed using hemispheres at eight different angles that are 45 degrees apart from each other around the pipe. Since different insulation types have different destruction pressures, different ZOls must be determined for each type of insulation. Table 3.b.1-1 shows the primary side break equivalent ZOI radii divided by the break diameter (LID) for each representative material in the VEGP Units 1 and 2 containment buildings. See Table 3.a.2-1 for ZOI sizes for SSBs. Table 3.b.1 Primary Side Break ZOI Radii for VEGP Insulation Types Destruction Pressure ZOI Radius/Break Diameter Insulation Type losi) (LID) Unjacketed Nukon 6 17.0* Qualified Coatings Unknown 4.0** Fire Barrier Material Unknown 11.7***

• NRC SE for NEI 04-07 (Reference 3)
  • "Revised Guidance Regarding Coatings Zone of Influence for Review of Final Licensee Responses to Generic Letter 2004-02" ADAMS # ML100960495
    • The destruction pressure of the lnteram E-50 series fire barrier material at VEGP is unknown. However, its ZOI size was assumed to be 11.70 based on comparison to the robustness of Temp-Mat.

In some cases, if the ZOI for a particular material is very large (i.e., it has a low destruction pressure or is located on a large pipe); the radius of the sphere may extend beyond robust barriers located near the break. Robust barriers consist of structures, such as concrete walls that are impervious to jet flow and prevent further expansion of the jet. Insulation in the shadow of large robust barriers can be assumed to remain intact to a certain extent (Reference 3, Section 3.4.2.3). Due to the compartmentalization of containment in VEGP Units 1 and 2, the insulation on the opposite side of the compartment walls can be assumed to remain intact. However, the SG compartments share an opening where a break jet could extend, so this was accounted for by including destruction of some of the insulation in these areas. All ZOls were truncated to account for robust barriers and compartment openings per NEI 04-07 Volume 2 (Reference 3). ES-26

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Volumetric debris quantities were determined by measuring the interference between a ZOI and its corresponding debris source. This was done within the CAD model. No insulation debris would be generated outside of the ZOls (Reference 2). This practice is considered acceptable by the NRC as stated in the SE for NEI 04-07 (Reference 3, Section 3.4.3.2).

2. Provide destruction ZOls and the basis for the ZOls for each applicable debris constituent.

Response to 3.b.2: See the Response to 3.b.1.

3. Identify if destruction testing was conducted to determine ZOls. If such testing has not been previously submitted to the NRC for review or information, describe the test procedure and results with reference to the test report(s).

Response to 3.b.3: VEGP applied the ZOI refinement discussed in NEI 04-07 Volume 2 (Reference 3, Section 4.2.2.1.1 ), which allows the use of debris-specific spherical ZOls. No new destruction testing was used to determine the ZOls listed above.

4. Provide the quantity of each debris type generated for each break location evaluated. If more than four break locations were evaluated, provide data only for the four most limiting locations.

Response to 3.b.4: Using the ZOls listed in this section, the breaks selected in the Response to 3.a.1, and the size distribution provided in the Response to 3.c.1 of this enclosure, quantities of generated debris for each break case were calculated for each type of insulation. The quantities of debris generated for the four most limiting break cases are listed below in Table 3.b.4-1 as determined in the debris generation calculation. The quantities of debris generated for the four most limiting break cases that do not fail any of the acceptance criteria in the NARWHAL evaluation (see Enclosure 3 Section 13.0) are listed below in Table 3.b.4-2. See Table 3.h.5-1 in the Response to 3.h.5 for the quantity of qualified and unqualified coatings for the four most limiting break locations. See Response to 3.d.3 for the quantity of latent debris. E5-27

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprie~ary) Table 3.b.4-1: The Four Overall Worst-Case Breaks Break Location 11201-004-6-RB 11201-001-5-RB 11201-001-3-RB 11201-004-4-RB Break Size 29" 29" 29" 29" Break Type DEGB DEGB DEGB DEGB Fine 289.3 287.5 280.9 276.0 Nukon Small 999.5 991.4 962.9 938.6 (ft3) Large 451.6 454.0 460.1 473.8 Intact 487.9 490.5 497.1 511.9 Fire Barrier Fine 0.0 2.4 2.4 0.0 (ft3) Small 0.0 2.4 2.4 0.0 Fire Barrier Particulate 0.0 26.9 26.8 0.0 (lbm) Table 3.b.4-2: The Four Worst-Case Breaks that Do Not Fail Any Acceptance Criteria for the Single Train Failure Equipment Configuration Break Location 11201-004-4-RB 11201-001-3-RB 11201-003-5-RB 11201-002-5-RB Break Size 20" 23" 19" 16" Break Type Partial Partial Partial Partial Fine 48.4 47.2 50.7 52.3 Nukon Small 151.2 160.5 186.5 168.0 (ft3) Large 122.1 80.9 46.6 118.6 Intact 132.0 87.4 50.3 128.1 Fire Barrier Fine 0.0 0.0 0.0 0.0 (ft3) Small 0.0 0.0 0.0 0.0 Fire Barrier Particulate 0.0 0.0 0.0 0.0 (lbm)

5. Provide total surface area of all signs, placards, tags, tape, and similar miscellaneous materials in containment.

Response to 3.b.5: Labels, tags, stickers, placards, and other miscellaneous or foreign materials were evaluated via walkdown. As with latent debris, a foreign material walkdown was only performed for Unit 1. However, based on the similarity between units discussed previously, Unit 1 data is considered applicable for Unit 2. The amount of foreign materials found during the walkdown was 2.0 ft2 . However, for conservatism, a total surface area of 4.0 ft2 was assumed in the VEGP debris generation calculation, and 50 ft2 was used in the NARWHAL conditional failure probability (CFP) calculation. E5-28

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary)

c. Debris Characteristics The objective of the debris characteristics determination process is to establish a conservative debris characteristics profile for use in determining the transportability of debris and its contribution to head loss.

1. Provide the assumed size distribution for each type of debris.

Response to 3.c.1: A summary of the material properties of the debris types found within containment are listed in Table 3.c.1-1 below. Table 3.c.1 Debris Material Properties Characteristic Density Debris Distribution Size (lbm/ft3) (µm) 2.4 (bulk) Nukon See section below 7 159 (fiber) Fiber (30% of total mass): 54.3 (bulk) 1.5 (fiber) 50% Fines 175 (fiber) Fire Barrier 50% Smalls Particulate (70% of total mass): 151 (particulate) 1O (particulate) 100% Particulate Qualified 208 (IOZ) 100% Particulate 10 Coatings 107 (Epoxy) 208 (IOZ) Unqualified and Degraded 100% Particulate 109 (Epoxy) 10 Coatings 122 (Alkyd) Nukon Low-Density Fiberglass Insulation The bulk density of Nukon is 2.4 pounds mass per cubic foot (lbm/ft3), and the individual fiber density is 159 lbm/ft3, per NEI 04-07 (Reference 2). The characteristic diameter of the individual fibers is 7 micrometers (µm). A baseline analysis of Nukon includes a size distribution with two categories: 60 percent small fines, and 40 percent large pieces, per NEI 04-07 (Reference 2). The debris generation calculation uses a four-category size distribution based on the guidance in NEI 04-07 Volume 2 (Reference 3). This guidance provides an approach for determining a size distribution for low-density fiberglass using the air jet impact test (AJIT) data, with conservatism added due to the potentially higher level of destruction from a two-phase jet. Within the 170 ZOI, the size distribution varies based on the distance of the insulation from the break (i.e., insulation debris E5-29

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprieftary) generated near the break location consists of more small pieces than insulation debris generated near the edge of the ZOI). Consequently, the following equations were developed to determine the fraction of fines (individual fibers), small pieces (less than 6 inches), large pieces (greater than 6 inches), and intact blankets as a function of the average distance between the break point and the centroid of the affected debris measured in units of pipe diameters (C). (OD H 4D) = 0.2 Ffines(C) (4D H 15D) = -0.01364

• C + 0.2546 (15D H 17D) = -0.025
• c + 0.425 (OD H 4D) = 0.8 Fsmalls (C) ( 4D H 15D) = -0.0682
• C + 1.0724 (15D H 17D) = -0.025
• c + 0.425 (OD H 4D) = 0 Frarge(C) ( 4D H 15D) = 0.0393
• C - 0.157 (15D H 17D) = -0.215
• c + 3.655 (OD H 4D) = 0 FintactCC) (4D H 15D) = 0.0425
• C - 0.170 (15D H 17D) = 0.265
• c - 3.505

2. Provide bulk densities (i.e., including voids between the fibers/particles) and material densities (i.e., the density of the microscopic fibers/particles themselves) for fibrous and particulate debris.

Response to 3.c.2: See the Response to 3.c.1 for the material and bulk densities of the various types of debris.

3. Provide assumed specific surface areas for fibrous and particulate debris.

Response to 3.c.3: Specific surface areas could be calculated for each debris type based on the characteristic diameter described in the Response to 3.c.1. However, specific surface areas were not calculated or used for the Vogtle head loss evaluation.

4. Provide the technical basis for any debris characterization assumptions that deviate from NRG-approved guidance.

Response to 3.c.4: The lnteram E-50 Series fire barrier material was assumed to be comprised of 70 percent particulate and 30 percent fiber by weight. These values fall within the ranges given in the E-50 Material Safety Data Sheet (MSDS). It was assumed that E5-30

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) the fiber constituent of 3M E-50 would fail as 50 percent fines and 50 percent small pieces sized between % inch and 4 inches. This engineering judgment is based on observations from exploratory testing with a 1500 psi pressure washer (the same type used for NEI fiber preparation). This assumption is conservative because the observations indicate that it is a very robust material and is less likely to break up into fines than low density fiberglass (LDFG).

d. Latent Debris The objective of the latent debris evaluation process is to provide a reasonable approximation of the amount and types of latent debris existing within the containment and its potential impact on sump screen head loss.

1. Provide the methodology used to estimate the quantity and composition of latent debris.

Response to 3.d.1: The evaluation for latent debris at VEGP was performed in a manner consistent with the NRC NEI 04-07 SE approved methodology (Reference 3, Section 3.5.2.3). The total source term was determined through the collection of debris samples from multiple locations throughout the containment. Conservatism was added by sampling those areas that exhibited unusually large concentrations of dirt and dust. In addition to dirt and dust, foreign materials and other debris sources were surveyed and documented including lint, paint chips, fibers, pieces of paper (shredded or intact), plastic, tape, adhesive labels, and fines or shards of thermal insulation, fireproof barrier, or other materials that are already present in the containment prior to a postulated break in a high-energy line inside containment. Vertical, horizontal, and equipment surfaces were sampled for dirt and dust by wiping with muslin cloth. Sample areas were chosen by cognizant engineering personnel with the intent to produce bounding results. The containment was divided into categories from which a minimum of three samples were taken. Prior to collecting samples, the containment was surveyed through a series of walkdowns to locate desirable sample locations.

2. Provide the basis for assumptions used in the evaluation.

Response to 3.d.2: See Response to 3.d.3. E5-31

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary)

3. Provide results of the latent debris evaluation, including amount of latent debris types and physical data for latent debris as requested for other debris under c.

above. Response to 3.d.3: Latent debris includes dirt, dust, lint, paint chips, fines, and shards of loose thermal insulation fibers that could potentially transport to the sump strainers during recirculation. Latent debris can be introduced into containment several ways, including by deterioration of items such as insulation and coatings and by personnel tracking in particulate and fibers from outside containment. The quantity of latent debris is calculated in the debris generation calculation. A walkdown of VEGP Unit 1 was performed to measure quantities of latent debris, and the total quantity was calculated based on those samples. The total amount of latent debris calculated based on walkdown data was 60 lbm, but 200 lbm is assumed in the debris generation calculation. This conservatively bounds the 60 lbm of actual latent debris with ample operating margin. Table 3.d.3-1 lists the assumed latent fiber and particulate constituents and their material characteristics. Latent debris is assumed to consist of 15 percent fiber and 85 percent particulate by mass, per the NRC NEI 04-07 SE (Reference 3, page 50). Based on NEI 04-07 Volume 2 (Reference 3, Sections 3.5.2.3, 3.7.2.3.2.3), the size and density of latent particulate were assumed to be 17.3 µm and 168.6 lbm/ft3, respectively. Additionally, the bulk density and microscopic density of latent fiber were assumed to be 2.4 lbm/ft3 and 93.6 lbm/ft3, respectively. Latent fiber is assumed to have a characteristic size of 5.5 µm. This is reasonably conservative, as it is the smallest fiber diameter listed in Table 3-2 of the general reference for low-density fiberglass found in NEI 04-07 (Reference 2). Table 3.d.3 Latent Fiber and Particulate Constituents Latent Bulk Microscopic Characteristic Debris Density Density Size (lbm) (lbm/ft3 ) (lbm/ft3 ) (µm) Particulate (85%) 170 - 168.6 17.3 Fiber (15%) 30 2.4 93.6 5.5 Total 200

4. Provide amount of sacrificial strainer surface area allotted to miscellaneous latent debris.

Response to 3.d.4: There is no sacrificial strainer area allotted to miscellaneous latent debris in addition to that documented in the Response to 3.b.5. E5-32

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary)

e. Debris Transport The objective of the debris transport evaluation process is to estimate the fraction of debris that would be transported from debris sources within containment to the sump suction strainers.

1. Describe the methodology used to analyze debris transport during blowdown, washdown, pool-fill-up, and recirculation phases of an accident.

Response to 3.e.1: The methodology used in the transport analysis is based on the NEI 04-07 guidance and the associated NRC SE (Reference 3), as well as the refined methodologies suggested by the SE in Appendices Ill, IV, and VI (Reference 3). The specific effect of each of four modes of transport was analyzed in the debris transport calculations for each type of debris generated. These modes of transport are:

• Slowdown Transport - the vertical and horizontal transport of debris to all areas of containment by the break jet
• Washdown Transport - the vertical (downward) transport of debris by the containment sprays, break flow, and condensation
• Pool Fill-Up Transport - the transport of debris by break and containment spray flows from the RWST to regions that may be active or inactive during recirculation
• Recirculation Transport - the horizontal transport of debris from the active portions of the recirculation pool to the sump screens by the flow through the ECCS The logic tree approach was applied for each type of debris determined from the debris generation calculation. The logic tree shown in Figure 3.e.1-1 is slightly different from the baseline. This departure was made to account for certain non-conservative assumptions identified by the NRC SE (Reference 3), including the transport of large pieces, erosion of small and large pieces, the potential for washdown debris to enter the pool after inactive areas have been filled, and the direct transport of debris to the sump screens during pool fill-up.

E5-33

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Debris Size. Blo.wdo'lm Wash down PoolfUI CFD Recilcufation Erosion IFraction of Debris Transport Transpcrt Tra.'ISpOrt I Transport I at Sump Retained on Structures up pH iranspon Coot:U-.ment I WasJi,:.dDown Sed:mer.t 1ransport

                         """'                                     ru.*uve r-oot I    Sedimeru:

Lovrer Sump Screens Contai-n:nent lrocll,,.,r<<>1 I Erodes to Fines Retained on StrucWres Remains intact UPP"' 1ransport Contafr.ment I I Erodes to Fines Wash.E-d Down Se<f:ment I Remains intact P.ieo:s uanspon Actve r<Xn I I Erodes to Fines Sediment I Debris Remains intact Generaitio."11 LOl't'S'  ::iump .:-creens Conta1unent Inactive-Pool I Erodes to Flnie-s Retained on I Struch.Jras Hema1ns mtact upp..- 1ransport Conb:mnent I J Erodes to Fine; Washed Dm*.n I Sediment I Remarns mtact Large P:ie-ces I transport Active Pool I I Erodes to Fines SedJnent I Remains intact

                                    ~  ...

Coota:runent

                                                               ~ump.........,..,,.ens lnacti.\.'ePool Rebinedon Strucb.Jres UPP"'                                                  lransport Coobimnent                                           II Washed DD""*YTl Sed:ment I    1ranspm L.3111" Piereswith Actvo Pocl I

Jackef.ng  ;:,eOiment Intact Lower SumpS~ns ContainmE!nt lnactivePoc:4 Figure 3.e.1-1: Generic Debris Transport Logic Tree The basic methodology for the VEGP transport analysis is summarized below.

1. The CAD model was provided as input to determine break locations and sizes.

2. The debris generation calculation was provided as input into the calculation for debris types and sizes.

3. Potential upstream blockage points were qualitatively addressed.

E5-34

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary)

4. The fraction of debris blown into upper containment and lower containment for each compartment was determined based on the volumes of upper and lower containment.

5. The fraction of debris washed down by containment spray flow was determined along with the locations where the debris would be washed down.

6. The quantity of debris transported to inactive areas or directly to the sump strainers was calculated based on the volume of the inactive and sump cavities proportional to the water volume at the time these cavities are filled.

7. The location of each type/size of debris at the beginning of recirculation was determined based on the break location.

8. A CFO model was developed in Flow-30 to simulate the flow patterns that would develop during recirculation.

9. A graphical determination of the transport fraction of each type of debris was made using the velocity and turbulent kinetic energy (TKE) profiles from the CFO model output, along with the determined initial distribution of debris.

10. The initial recirculation transport fractions from the CFO analysis were gathered to determine the final recirculation transport fractions for input into the logic trees.

11. The quantity of debris that could experience erosion due to the break flow or spray flow was determined.

12. The overall transport fraction for each type/size of debris was determined by combining each of the previous steps into logic trees.

Potential Upstream Blockage Points Potential upstream blockage points were qualitatively addressed in the debris transport calculation. It was determined that there are not any upstream blockage points in the VEGP containment building. Upstream effects are discussed in the Response to 3.1. i I I I E5-35

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) CFO Model of Containment Recirculation Pool A diagram showing the significant parts of the CFO model is shown in Figure 3.e .1-

2. The sump mass sink and the various direct and runoff spray regions are highlighted .

Accumulator Inj ection Modeled Spray line Modeled Mass Drainage Through Source Loop 2 Annulus Modeled Modeled Spray Drainage Falling Pressurizer Surge Through Steam Line Modeled Generator Mass Source Compartm nts Loop 4 Modeled Sinks Mass Source CS Sump Strainer Module Mass Sinks Figure 3.e.1-2: Significant Features in CFO Model The key CFO modeling attributes/considerations included the following : Computational Mesh A rectangular mesh was defined in the CFO model that was fine enough to resolve important features , but not so fine that the simulation would take excessively long to run. A mesh spacing of 5 inch by 5 inch was used in the x and y directions and 3-inch to 4-inch mesh space was used in the z direction . ES-36

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Modeling of Containment Spray Flows For CFO cases with CS activated, various plan and section drawings, as well as the containment building CAD model, were considered. Spray water would drain to the pool through many pathways. Some of these pathways include the steam generator enclosures, the various openings in the operating deck, the annulus through the various open sections of grating, and the refueling canal drains. The sprays were introduced near the surface of the pool. Modeling of Break Flow The water falling from the postulated break would introduce momentum into the containment pool that influences the flow dynamics. This break stream momentum was accounted for by introducing the break flow to the pool at the velocity that a freefalling object would have if it fell the vertical distance from the location of the break to the surface of the pool. Modeling of the Sump Strainers Each sump strainer in VEGP consists of four columns of stacked disks with a solid plate on top. In the CFO model, each strainer was modeled with a plate above it to prevent flow from entering through the top of the strainers. The sump strainers were modeled as having flow across their surfaces proportional to the areas of the strainers. A negative flow rate was set for the sump mass sink, which tells the CFO model to draw the specified amount of water from the pool over the entire exposed surface area of the mass sink obstacle. Turbulence Modeling Several different turbulence-modeling approaches can be selected for a Flow-30 calculation. The approaches (ranging from least to most sophisticated) are:

• Prandtl mixing length
• Turbulent energy model
• Two-equation k-E model
• Renormalized group theory (RNG) model
• Large eddy simulation model The RNG turbulence model was determined to be the most appropriate for this CFO analysis. The RNG model has a large spectrum of length scales that would likely exist in a containment pool during emergency recirculation. The RNG approach applies statistical methods in a derivation of the averaged equations for turbulence quantities (such as TKE and its dissipation rate). RNG-based turbulence schemes rely less on empirical constants while setting a framework for the derivation of a range of models at different scales.

E5-37

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Steady-State Metrics The CFO models were started from a stagnant state at a defined pool depth and run long enough for steady-state conditions to develop. A plot of mean kinetic energy was used to determine when steady-state conditions were reached. Checks were also made of the velocity and turbulent energy patterns in the pool to verify that steady-state conditions were reached. Debris Transport Metrics The metrics for predicting debris transport during recirculation are the TKE necessary to keep debris suspended, and the flow velocity necessary to tumble sunken debris along the floor or lift it over a curb. Debris transport metrics have been derived or adopted from data. The metrics utilized in the VEGP transport analysis originate from the sources below.

• NUREG/CR-6772 Tables 3.1 and 3.2 (Reference 37)
• NUREG/CR-6808 Figure 5.2 and Tables 5-1 and 5-3 (Reference 39)

Graphical Determination of Debris Transport Fractions for Recirculation The following steps were taken to determine what percentage of a particular type of debris could be expected to transport through the containment pool to the emergency sump screens. Detailed explanations of each bullet are provided in the paragraphs below.

• Colored contour velocity and TKE maps were generated from the Flow-30 results in the form of bitmap files indicating regions of the pool through which a particular type of debris could be expected to transport.
• The bitmap images were overlaid on the initial debris distribution plots and imported into AutoCAD with the appropriate scaling factor to convert the length scale of the color maps to feet.
• Closed polylines were drawn around the contiguous areas where velocity and TKE were high enough that debris cou.ld be carried in suspension or tumbled along the floor to the sump strainers for uniformly distributed debris.
• The areas within the closed polylines were determined using an AutoCAD querying feature.
• The combined area within the polylines was compared to the initial debris distribution area.
• The percentage of a particular debris type that would transport to the sump strainers was determined based on the above comparison.

Plots showing the TKE and the velocity magnitude in the pool were generated for each case to determine areas where specific types of debris would be transported. The limits on the plots were set according to the minimum TKE or velocity metrics necessary to move each type of debris. The overlying yellow areas represent ES-38

Enclosure 5 Suppl*emental Response to NRC Generic Letter 2004-02 (Non-Proprietary) regions where the debris would be suspended, and the red areas represent regions where the debris would be tumbled along the floor (see Figure 3.e.1-4). The yellow TKE portion of the plots is a three-dimensional representation of the TKE. Since the TKE is a three-dimensional representation, the plots do not show the TKE at any specific elevation. Rather, any debris that is shown to be present in this yellow area will transport, regardless of the elevation of TKE in the pool. The velocity portion of the plots represents the velocity magnitude just above the floor level (1.5 inches), where tumbling of sunken debris could occur. Directional flow vectors were also included in the plots to determine whether debris in certain areas would be transported to the sump strainers or transported to less active regions of the pool where it could settle to the floor (blue regions). The following figures and discussion are presented as an example of how the transport analysis was performed for a generic small debris type. This same approach was used for other debris types analyzed at VEGP. As shown in Figure 3.e.1-3, the small debris (depicted by green shading) was initially assumed to be uniformly distributed between the break location and the sump strainers. The break location in this scenario is a break in the annulus on the pressurizer surge line. E5-39

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Figure 3.e.1-3: Distribution of Small Debris in Lower Containment Figure 3.e.1-4 shows that the turbulence (yellow regions) and the velocity (red regions) in the pool (blue regions) are high enough to transport the generic small debris to the sump strainers during recirculation . The initial distribution area (Figure 3.e.1-3) was overlaid on top of the plot showing tumbling velocity , TKE , and flow vectors (Figure 3.e.1-4) to determine the recirculation transport fraction (Figure 3.e.1-5). ES-40

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Figure 3.e.1-4: TKE and Velocity with Limits Set at Suspension/ Tumbling of Small Generic Debris E5-41

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Figure 3.e.1-5: Floor Area where Small Generic Debris Would Transport to the Sump Strainers (hatched area) This same analysis was applied for each type of debris at VEGP . Recirculation-pool transport fractions were identified for each debris type associated with the location of its initial distribution . This includes a recirculation transport fraction for debris blown to lower containment, debris washed down inside the secondary shield wall , and debris washed down into the annulus. Erosion Discussion Due to the turbulence in the recirculation pool and the force of break and spray flow, Nukon debris may erode into smaller pieces , making transport of this debris to the strainer more likely. Results of the Drywell Debris Transport Study indicate that debris exposed to containment sprays above the recirculation pool undergo an erosion fraction of less than 1%. Therefore , a 1% erosion fraction for debris held up on gratings and other miscellaneous structures at Vogtle was used . E5-42

Enclosure 5 Supplemental Response to NRC G'eneric Letter 2004-02 (Non-Proprietary) Erosion Test Discussion To estimate erosion that would occur in the recirculation pool at Vogtle, generic 30-day erosion testing was performed. ((

                           ))1 Input Parameters The flow conditions used for the testing were based on prototypical plant flow conditions. The target velocity selected for the erosion testing was ((
                    )) 1 based on the tumbling velocity required to transport small pieces of LDFG (Reference 37). Since small pieces of LDFG would transport at higher velocities, the non-transporting small pieces of LDFG on a containment pool floor would be exposed to a velocity less than ((              ))1. Typically, in regions where the velocity is lower than 0.12 ft/s, the pool is relatively quiescent, and the turbulence levels are very low. Based on a review of the average turbulence levels for various plants in the quiescent "non-transport" regions, a target turbulence of ((
    ))1 was selected for the erosion testing.

To prevent potential contamination of the samples from the minerals in tap water, deionized (DI) water was selected for the erosion testing. In prototypical plant conditions, the containment pool water is borated and buffered. Based on observations during chemical effects testing, chemical precipitates tend to accumulate on exposed fiberglass. This effect can mask the actual erosion. However, by using pure water, this phenomenon was eliminated in the erosion testing. Erosion Test Durations Table 3.e.1-1 shows the test matrix for the erosion testing. The length of the pre-test was selected based on the longest period of time that any one filter was installed during the primary test and post-test, or 5 days. The length of the primary test was based on a full 30-day mission time. The length of the post-test was based on the filter measurements during the primary test as well as the initial filter measurements during the post-test, or 5 days. Since the primary test measurements showed that the majority of erosion occurs within the first 10 days, a 10-day test was determined to be an adequate length of time to accomplish the purposes of the post-test. This test was run for a total of 13 days. 1 Alien trade secret ES-43

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Table 3.e.1-1 Erosion Test Durations Test Duration Description Quantify weight change of a filter under test conditions Pre-Test 5 days without anv fiben:1lass in the flume. Measure overall weight loss of fiber clumps after 30 days exposure to flow. Also, determine time dependent Primary Test 30 days erosion curve by measuring weight change of filters throughout. Determine repeatability of primary test, and confirm that there are no unknown long-term phenomena that Post-Test 13 days resulted in a non-conservative weight gain of the fiber samples durinQ the later staQes of the primarv test.

2. Provide the technical basis for assumptions and methods used in the analysis that deviate from the approved guidance.

Response to 3.e.2: The methodology used in the transport analysis is based on and does not deviate from the NRC approved NEI 04-07 guidance and the associated NRC SE (Reference 3) for refined analyses, as well as the refined methodologies suggested by the SE in Appendices Ill, IV, and VI.

3. Identify any computational fluid dynamics codes used to compute debris transport fractions during recirculation and summarize the methodology, modeling assumptions, and results.

Response to 3.e.3: To assist in the determination of recirculation transport fractions, several CFO simulations were performed using Flow-30, a commercially available software package. Seven breaks were investigated that included single- and two-train recirculation with sprays both on and off to ensure a conservative representation of the post-LOCA containment-sump flow velocities. Breaks were also evaluated inside and outside the secondary shield wall to determine which scenario(s) would maximize debris transport. The simulation results include a series of contour plots of velocity and TKE. These results have been combined with settling and tumbling velocities from the GSl-191 literature to determine the recirculation transport fractions for all debris types present in the VEGP containment building. See Response to 3.e.1 for additional discussion of the CFO results. E5-44

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary)

4. Provide a summary of, and supporting basis for, any credit taken for debris interceptors.

Response to 3.e.4: No credit was taken for debris interceptors.

5. State whether fine debris was assumed to settle and provide basis for any settling credited.

Response to 3.e.5: No credit was taken for settling of fine debris.

6. Provide the calculated debris transport fractions and the total quantities of each type of debris transported to the strainers.

Response to 3.e.6: The following debris transport fractions listed in Table 3.e.6-1 through Table 3.e.6-14 are inputs to the NARWHAL CFP calculation. Note that these fractions result in 'the bounding quantity of debris transported to the strainer. The debris transport quantities are provided in Tables 3.e.6-15 and 3.e.6-16. E5-45

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Slowdown Transport Table 3.e.6-1 shows the bounding blowdown transport fractions as a function of break location and debris type. Table 3.e.6-1: Slowdown Transport Fractions Transport Fraction To Upper To Lower Remaining Break Location Debris Type Containment Containment in (UC) (LC) Compartment Fines (all) 80% 20% 0% Steam Generator Small Nukon & 39% 61% 0% Compartments Fire Barrier Large Nukon 0% 100% 0% Fines (all) 80% 20% 0% Small Nukon & Reactor Cavity 39% 61% 0% Fire Barrier Large Nukon 0% 100% 0% Fines (all) 80% 20% 0% Pressurizer Small Nukon & 69% 9% 22% Compartment Fire Barrier Large Nukon 0% 0% 100% Fines (all) 80% 20% 0% Annulus - Small Nukon & Pressurizer Surge 35% 18% 47% Fire Barrier Line Lan:ie Nukon 0% 25% 75% Fines (all) 80% 20% 0% Annulus - Small Nukon & Accumulator 17% 83% 0% Fire Barrier Injection Line Larqe Nukon 0% 100% 0% E5-46

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Washdown Transport Table 3.e.6-2 shows the bounding washdown transport fractions as a function of containment spray activation and debris type. Note that these transport fractions do not depend on the location of the break. Table 3.e.6-2: Washdown Transport Fractions Transport Fraction Sprays Debris Type Washed Down Washed Down Washed Down Initiated? in Annulus Inside SSW RFC Drains Fines (all) 53% 37% 10% Small Nukon & Yes 43% 37% 10% Fire Barrier Large Nukon 0% 0% 10% Nukon Fines/ 10% No Latent Debris All Others 0% 0% 0% Pool-Fill Transport Table 3.e.6-3 shows the bounding pool fill transport fractions as a function of debris type. Table 3.e.6-3: Pool fill Transport Fractions Pool Fill Transport Fraction Debris Type Elevator Cavity Each ECCS Sump Fines (all) 2% 0.75% Small Nukon & Fire Barrier 2% 0.75% Large Nukon 2% 0.75% Unqualified Coatings 0% 0% Recirculation Transport For the recirculation transport fractions, seven different cases were evaluated in the debris transport calculation. These cases are listed below:

• Case 1: LBLOCA in SG Compartment Loop 4, Sprays not Activated, Two Trains Operational
• Case 2: LBLOCA in Pressurizer Surge Line in Annulus, Sprays not Activated, Two Trains Operational
• Case 3: LBLOCA in Accumulator Injection Line in Annulus, Sprays not Activated, Two Trains Operational
• Case 4: LBLOCA in SG Compartment Loop 4, Sprays not Activated, One Train Operational E5-47

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary)

• Case 5: LBLOCA in SG Compartment Loop 4, Sprays not Activated, Two Trains Operational, High Water Level *
• Case 6: LBLOCA in SG Compartment Loop 2, Sprays Activated, Two Trains Operational
• Case 7: LBLOCA in SG Compartment Loop 4, Sprays Activated, Two Trains Operational The bounding recirculation transport fractions for fine debris are shown in Table 3.e.6-4.

Table 3.e.6-4: Recirculation Transport Fractions for Fine Debris Washed Inside Washed Washed Break Case Sump Secondary Shield In Down Recirculation Wall Annulus RFC RHRA 50% NA NA NA Case 1 RHR B 50% NA NA NA RHRA 50% NA NA NA Case 2 RHR B 50% NA NA NA RHRA 50% NA NA NA Case 3 RHR B 50% NA NA NA Case4 RHR B 100% NA NA NA RHRA 50% NA NA NA Case 5 RHR B 50% NA NA NA CSA 21% 21% 21% 21% RHRA 29% 29% 29% 29% Case 6 CS B 21% 21% 21% 21% RHR B 29% 29% 29% 29% CSA 21% 21% 21% 21% RHRA 29% 29% 29% 29% Case 7 CS B 21% 21% 21% 21% RHR B 29% 29% 29% 29% E5-48

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) The bounding recirculation transport fractions for small fiber debris are shown in Table 3.e.6-5. Table 3.e.6-5: Recirculation Transport Fractions for Small Fiber Debris Washed Inside Break Washed In Washed Case Sump Secondary Shield Recirculation Annulus Down RFC Wall RHRA 0% NA NA NA Case 1 RHR B 0% NA NA NA RHRA 9% NA NA NA Case 2 RHR B 33% NA NA NA RHRA 0% NA NA NA Case 3 RHR B 0% NA NA NA Case 4 RHRB 0% NA NA NA RHRA 0% NA NA NA Case 5 RHR B 0% NA NA NA CSA 24% 23% 28% 0% RHRA 0% 0% 9% 0% Case 6 CS B 8% 8% 19% 100% RHR B 9% 10% 7% 0% CSA 20% 17% 25% 0% RHRA 0% 0% 10% 0% Case 7 CS B 8% 9% 6% 100% RHR B 26% 25% 20% 0% E5-49

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) The bounding recirculation transport fractions for large fiber debris are shown in Table 3.e.6-6. Table 3.e.6-6: Recirculation Transport Fractions for Large Fiber Debris Washed Inside Washed Break Washed Case Sump Secondary Shield In Recirculation Down RFC Wall Annulus RHRA 0% NA NA NA Case 1 RHR B 0% NA NA NA RHRA 4% NA NA NA Case 2 RHR B 20% NA NA NA RHRA 0% NA NA NA Case 3 RHR B 0% NA NA NA Case4 RHR B 0% NA NA NA RHRA 0% NA NA NA Case 5 RHR B 0% NA NA NA CSA 0% NA NA NA RHRA 0% NA NA NA Case 6 CS B 0% NA NA NA RHR B 0% NA NA NA CSA 0% NA NA NA RHRA 0% NA NA NA Case 7 CS B 0% NA NA NA RHR B 0% NA NA NA E5-50

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Overall Debris Transport Transport logic trees were developed for each size and type of debris generated. These trees were used to determine the total fraction of debris that would reach the sump strainers in each of the postulated cases. The bounding overall transport fractions are presented in Table 3.e.6-7 through Table 3.e.6-14. Note that the near annulus breaks represent breaks that are within close proximity to the strainers in the annulus, and that the far annulus breaks represent breaks that are far away from the strainer in the annulus. The values below are slightly different than what is calculated in NARWHAL. This is because the total transport fractions are entered in NARWHAL and the time-dependent fiber accumulation on the strainers is calculated based on the flow split. Additionally, some fiber penetrates the strainers and accumulates on the core, which also contributes to this slight difference. Table 3.e.6-7: Overall Transport Fractions for an SG Compartment/ Reactor Cavity Break, Two Trains Operational, CS On Debris Type CSA RHRA CSB RHRB Total Nukon Individual Fibers 21% 29% 21% 29% 100% Nukon & Fire Barrier Small Pieces 20% 5% 13% 25% 63% Nukon Large Pieces 3% 4% 3% 4% 14% Nukon Intact Pieces 0% 0% 0% 0% 0% Fire Barrier Fines & Particulate 21% 29% 21% 29% 100% Qualified Coatings (IOZ, Epoxv) 21% 29% 21% 29% 100% Unqualified Epoxy Coatings Particulate 21% 29% 21% 29% 100% Unqualified IOZ CoatinQs Particulate 21% 29% 21% 29% 100% Unqualified Alkyd Coatings Particulate 21% 29% 21% 29% 100% Latent Dirt/Dust Particulate & Fiber 21% 29% 21% 29% 100% Table 3.e.6-8: Overall Transport Fractions for an SG Compartment/ Reactor Cavity Break, One Train Operational, CS Off Debris Type CSA RHRA CSB RHRB Total Nukon Individual Fibers NA NA NA 27% 27% Nukon & Fire Barrier Small Pieces NA NA NA 6% 6% Nukon Large Pieces NA NA NA 10% 10% Nukon Intact Pieces NA NA NA 0% 0% Fire Barrier Fines & Particulate NA NA NA 27% 27% Qualified Coatings (IOZ, Epoxy) NA NA NA 27% 27% Unqualified Epoxy Coatings Particulate NA NA NA 47% 47% Unqualified IOZ Coatings Particulate NA NA NA 60% 60% Unqualified Alkyd Coatings Particulate NA NA NA 100% 100% Latent Dirt/Dust Particulate & Fiber NA NA NA 31% 31% E5-51

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Table 3.e.6-9: Overall Transport Fractions for an SG Compartment/ Reactor Cavity Break, Two Trains Operational, CS Off Debris Type CSA RHRA CSB RHRB Total Nukon Individual Fibers NA 14% NA 14% 28% Nukon & Fire Barrier Small Pieces NA 3% NA 3% 6% Nukon Large Pieces NA 6% NA 6% 12% Nukon Intact Pieces NA 0% NA 0% 0% Fire Barrier Fines & Particulate NA 14% NA 14% 28% Qualified Coatings (IOZ, Epoxy) NA 14% NA 14% 28% Unqualified Epoxy Coatings Particulate NA 23% NA 23% 46% Unqualified IOZ Coatings Particulate NA 30% NA 30% 60% Unqualified Alkyd Coatings Particulate NA 50% NA 50% 100% Latent Dirt/Dust Particulate & Fiber NA 16% NA 16% 32% Table 3.e.6-10: Overall Transport Fractions for a Pressurizer Compartment Break, Two Trains Operational, CS Off Debris Type CSA RHRA CSB RHRB Total Nukon Individual Fibers NA 14% NA 14% 28% Nukon & Fire Barrier Small Pieces NA 1% NA 3% 4% Nukon Large Pieces NA 0% NA 0% 0% Nukon Intact Pieces NA 0% NA 0% 0% Fire Barrier Fines & Particulate NA 14% NA 14% 28% Qualified Coatings (IOZ, Epoxy) NA 14% NA 14% 28% Unqualified Epoxy Coatings Particulate NA 23% NA 23% 46% Unqualified IOZ Coatings Particulate NA 30% NA 30% 60% Unqualified Alkyd Coatings Particulate NA 50% NA 50% 100% Latent Dirt/Dust Particulate & Fiber NA 16% NA 16% 32% E5-52

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Table 3.e.6-11: Overall Transport Fractions for a Near Annulus Break, Two Trains Operational, CS Off Debris Type CSA RHRA CS B RHRB Total Nukon Individual Fibers NA 14% NA 14% 28% Nukon & Fire Barrier Small Pieces NA 2% NA 6% 8% Nukon LarQe Pieces NA 2% NA 6% 8% Nukon Intact Pieces NA 0% NA 0% 0% Fire Barrier Fines & Particulate NA 14% NA 14% 28% Qualified Coatings (IOZ, Epoxy) NA 14% NA 14% 28% Unqualified Epoxy Coatings Particulate NA 23% NA 23% 46% Unqualified IOZ Coatings Particulate NA 30% NA 30% 60% Unqualified Alkyd CoatinQs Particulate NA 50% NA 50% 100% Latent Dirt/Dust Particulate & Fiber NA 16% NA 16% 32% Table 3.e.6-12: Overall Transport Fractions for a Far Annulus Break, Two Trains Operational, CS Off Debris Type CSA RHRA CS B RHRB Total Nukon Individual Fibers NA 14% NA 14% 28% Nukon & Fire Barrier Small Pieces NA 5% NA 5% 10% Nukon LarQe Pieces NA 6% NA 6% 12% Nukon Intact Pieces NA 0% NA 0% 0% Fire Barrier Fines & Particulate NA 14% NA 14% 28% Qualified Coatings (IOZ, Epoxy) NA 14% NA 14% 28% Unqualified Epoxy Coatings Particulate NA 23% NA 23% 46% Unqualified IOZ Coatings Particulate NA 30% NA 30% 60% Unqualified Alkyd Coatings Particulate NA 50% NA 50% 100% Latent Dirt/Dust Particulate & Fiber NA 16% NA 16% 32% E5-53

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Table 3.e.6-13: Overall Transport Fractions for a Near Annulus Break, Two Trains Operational, CS On Debris Type CSA RHRA CSB RHRB Total Nukon Individual Fibers 21% 29% 21% 29% 100% Nukon & Fire Barrier Small Pieces 21% 3% 10% 27% 61% Nukon Large Pieces 3% 4% 3% 4% 14% Nukon Intact Pieces 0% 0% 0% 0% 0% Fire Barrier Fines & Particulate 21% 29% 21% 29% 100% Qualified Coatings (IOZ, Epoxy) 21% 29% 21% 29% 100% Unqualified Epoxy Coatings Particulate 21% 29% 21% 29% 100% Unqualified IOZ Coatings Particulate 21% 29% 21% 29% 100% Unqualified Alkyd Coatings Particulate 21% 29% 21% 29% 100% Latent Dirt/Dust Particulate & Fiber 21% 29% 21% 29% 100% Table 3.e.6-14: Overall Transport Fractions for a Far Annulus Break, Two Trains Operational, CS On Debris Type CSA RHRA cs 8 RHRB Total Nukon Individual Fibers 21% 29% 21% 29% 100% Nukon & Fire Barrier Small Pieces 20% 5% 13% 25% 63% Nukon Large Pieces 3% 4% 3% 4% 14% Nukon Intact Pieces 0% 0% 0% 0% 0% Fire Barrier Fines & Particulate 21% 29% 21% 29% 100% Qualified Coatings (IOZ, Epoxy) 21% 29% 21% 29% 100% Unqualified Epoxy Coatings Particulate 21% 29% 21% 29% 100% Unqualified IOZ Coatings Particulate 21% 29% 21% 29% 100% Unqualified Alkyd Coatings Particulate 21% 29% 21% 29% 100% Latent Dirt/Dust Particulate & Fiber 21% 29% 21% 29% 100% E5-54

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Transported Debris Quantities The transported debris quantities for the most limiting break cases identified in Tables 3.b.4-1 and 3.b.4-2 are shown below and were derived using the debris transport fractions provided in this section for a single train failure case. The debris transport quantities for the four most limiting break cases are listed below in Table 3.e.6-15 as determined in the NARWHAL CFP calculation. The quantities of debris transported for the four most limiting break cases that do not fail any of the strainer or core acceptance criteria are listed in Table 3.e.6-16. Note that the fiber quantity includes fines, small pieces, large pieces, intact pieces, and latent fiber debris. To calculate the transported quantities of debris presented in the following tables, the blowdown, washdown, pool-fill, and recirculation data (Table 3.e.6-1 through Table 3.e.6-6) are input into NARWHAL. However, the NARWHAL CFP calculation takes into account certain factors that the transport calculation does not consider in order to calculate the time-dependent arrival of debris on the strainer. For example, it takes into account various factors such as the RHR strainer switching over to recirculation before the CS strainer, and the flow split between the strainers. Therefore, the calculation of debris transported to the strainer in the NARWHAL CFP calculation is not a straightforward one. All breaks listed in the tables below occur on the hot leg and activate containment sprays since the break size for each break is greater than 15". The recirculation transport fractions for Case 7 (LBLOCA in SG Compartment Loop 4, Sprays Activated, Two Trains Operational) from the transport calculation were conservatively input into NARWHAL for the single train failure with containment sprays activated. This was done since there was not a CFO case in the transport calculation that examined one train failure with containment sprays activated, and it is conservative since the turbulence and the velocities in the pool during recirculation for two train operation with containment sprays activated are maximized. E5-55 I ,

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Table 3.e.6-15: Transported Debris for the Four Overall Worst-Case Breaks Break Location 11201-004-6-RB 11201-001-5-RB 11201-001-3-RB 11201-004-4-RB Break Size 29" 29" 29" 29" Break Type DEGB DEGB DEGB DEGB Fiber (ft3) RHR 639.8 635.4 617.9 605.8 cs 279.5 277.6 272.7 267.4 Particulate RHR 2, 166.5 2, 166.3 2,159.7 2,156.1 (lbm) cs 935.0 934.9 941.5 940.0 Calcium Phosphate RHR 73.5 73.5 73.5 73.5 (lbm) cs 42.3 42.2 42.1 42.1 Sodium Aluminum RHR 86.2 86.1 85.8 85.8 Silicate (lbm) cs 0.5 0.5 0.5 0.5 Fire Barrier Particulate RHR 0.0 20.3 20.2 0.0 (lbm) cs 0.0 8.8 8.8 0.0 Fire Barrier RHR 0.0 1.0 1.0 0.0 Fiber (ft3) cs 0.0 0.4 0.4 0.0 ES-56

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Table 3.e.6-16: Transported Debris for the Four Worst-Case Breaks that Do Not Fail the Acceptance Criteria Break Location 11201-004-4-RB 11201-001-3-RB 11201-003-5-RB 11201-002-5-RB Break Size 20" 23" 19" 16" Break Type Partial Partial Partial Partial Fiber (ft3 ) RHR 107.2 107.1 107.6 107.5 cs 46.9 46.7 46.4 46.6 Particulate RHR 2,103.5 2, 106.3 2, 109.9 2, 108.4 (lb) cs 917.0 918.2 910.5 909.9 Calcium Phosphate RHR 46.4 38.3 34.0 47.8 (lbm) cs 16.4 14.1 12.8 16.8 Sodium Aluminum RHR 54.3 52.5 51.5 54.6 Silicate (lbm) cs 0.3 0.3 0.3 0.3 Fire Barrier Particulate RHR 0.0 0.0 0.0 0.0 (lbm) cs 0.0 0.0 0.0 0.0 Fire Barrier RHR 0.0 0.0 0.0 0.0 Fiber (ft3 ) cs 0.0 0.0 0.0 0.0 ES-57 I I

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary)

f. Head Loss and Vortexing The objectives of the head loss and vortexing evaluations are to calculate head loss across the sump strainer and to evaluate the susceptibility of the strainer to vortex formation.

1. Provide a schematic diagram of the emergency core cooling system (ECCS) and containment spray systems (CSS).

Response to 3.f.1: See Figure 3.f.1-1 and Figure 3.f.1-2 for ECCS and CSS schematics, respectively. E5-58

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) IRC OA:C I I RWST CL1 CL2 CL3 cu CL1 IRC I ORC HUA I I-A 1 UO CL2 I IH111~ CL3 v c cu T 1111 TO R EGEN HX Figure 3.f.1-1 Emergency Core Cooling System Schematic ES-59

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) CO TAl NWENT SPRAY NO ZZLES C S PUMP A Fl

                                                                                                         ! 2!

207 I I CNMT IRC I ORC EMERGEN CY I CS PUMP B I

• • • • • • • SUMPS r*****: r******

~TSPI r-. ,. -. hsPi I ITS Pi I EN CAP SUL ATIO N

                                                                  ':_E_S~~'::_S I    .           1 II             I   *-*-*-*-*-*-*-*.I

• ii - . - . - .- *<-. - .-. T . - ~ - ' .llf!lB. _; 'GD3 B jj GUARDP IPE  :  ;-* *-i

~-*-*-*-*-*-*-*-*-*-*-*J I o;;; * - * - * - * - * - * - * - * - * - *** -. '002A _j SGD3A I -- * -

• Figure 3.f.1-2 Containment Spray System Schematic ES-60

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary)

2. Provide the minimum submergence of the strainer under small-break loss-of-coolant accident (SBLOCA) and large-break loss-of-coolant accident (LBLOCA) conditions.

Response to 3.f.2: The sump strainers are fully submerged during recirculation in all cases except reactor nozzle breaks (see Table 3.g.1-3). The highest elevation of the RHR strainer disk is 53-1 /4 inches (or 4.438 ft) above the containment floor (see Figure 3.f.2-1 ). The submergence of the highest elevation point of the RHR strainer is conservatively taken to be its minimum submergence. The height of the 16-disk RHR strainer bounds (i.e., is greater than) the height of the 14-disk CSS strainer, and the minimum submergence of the RHR strainers is always less than the minimum submergence of the CS strainers. 4.438 f1 I lotl *

• loot Ittl *
• It ol Figure 3.f.2-1: Side View of 16-Disk RHR Strainer The RHR and CS sump strainers are fully submerged under an LBLOCA that is not a reactor nozzle break. As shown in the Response to 3.g.1, the minimum LBLOCA water level during recirculation for a break that is not a reactor nozzle break is 4 .536 ft, and the minimum submergence of the RHR strainer is 0.098 ft.

The RHR sump strainers are also fully submerged for an LBLOCA at a reactor nozzle when CS is not activated . As shown in the Response to 3.g.1, the minimum water level during recirculation for this case is 4.977 ft, and the minimum submergence of the RHR strainer is 0.539 ft. The RHR and CS sump strainers are fully submerged under an SBLOCA. As shown in the Response to 3.g .1, the minimum SBLOCA water level during recirculation is 5.186 ft. Therefore, the minimum submergence of the RHR strainer is 0.748 ft. The RHR and CS sump strainers are not fully submerged for an LBLOCA caused by a reactor nozzle break that actuates CS . As shown in the Response to 3.g.1 , the E5-61

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) minimum water level for this case (3.054 ft) occurs at the start of recirculation. This pool level is 1.384 ft below the top of the RHR strainer. It should be noted that the water level increases to 4. 788 ft when the sump recirculation switchover is complete. This corresponds to a strainer submergence of 0.35 ft. According to Regulatory Guide (RG) 1.82 (Reference 107), the total strainer head loss of a partially submerged strainer should be less than half the submerged height of the strainer. This ensures the average hydrostatic head of the submerged portion of the strainer will be greater than the head loss through the debris bed. This requirement was used when evaluating partially-submerged strainers for VEGP.

3. Provide a summary of the methodology, assumptions, and results of the vortexing evaluation. Provide bases for key assumptions.

Response to 3.f.3: Summary of Vortex Tests In 2009, vortex testing was performed on a prototypical strainer module to observe the size, shape, and location of vortices that may develop as both the flow rate through the strainer and the submergence of the strainer module were varied. The vortex tests were performed during the head loss test described in the Response to 3.f.4. Both clean screen and debris laden vortex tests were performed. See Figure 3.f.4-1 for the layout of the test strainer and test tank. Two vortex tests were conducted at clean strainer conditions, as summarized below.

• The first clean strainer vortex test was started at a submergence level of 3.625 inches and an average approach velocity of 0.0258 ft/s. No vortexing was observed. The average approach velocity was then increased to 0.0355 ft/s. No vortexing was observed. The water level was then reduced to 1.825 inches below the top of the strainer. Again, no vortexing was observed.
• A second clean strainer vortex test was started with a strainer submergence of 4.175 inches and an approach velocity of 0.0306 ft/s. This approach velocity was maintained throughout the duration of this test. The water level in the tank was reduced to just below the top of the strainer. Some pump noise was audible and a small surface swirl was visible in the front right corner of the tank but no persistent vortices were observed.

Debris laden vortex tests were performed at the end of the thin bed and full-load head loss tests after adding all conventional and chemical debris. The test loop setup was the same as that used for the clean screen vortex test and the approach velocity for all debris laden vortex tests was 0.0136 ft/s. It should be noted that, during the course of the thin bed and full-load tests, the tank water level was ES-62

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) maintained at 3.675 +/- 0.5 inches above the strainer, and no appreciable vortices were visually observed.

• Water level was slowly reduced at the end of the thin bed test. Air ingestion was not observed until the water level was 0.25 inches below the top of the strainer.
• For the first full debris load test, when the water level was reduced to approximately 3 inches above the strainer, air-entrainment vortices were observed. The vortices became persistent when the water level reached 2.25 inches above the strainer.
• At the end of the second full debris load test, when the water level was reduced to approximately 3 inches above the strainer, air-entrainment vortices were observed. The vortices were not persistent until the water level reached 1.5 inches below the top of the strainer.

Vortexing of Plant Strainers For reference in the discussion below, the average approach velocity of the 16-disk RHR strainers in the plant is 0.0122 ft/s, and the average approach velocity for the 14-disk CS strainers in the plant is 0.0098 ft/s. These approach velocities are calculated using the strainer flow rates and surface areas shown in Table 3.f.3-1. These approach velocities are well bounded by that used during the clean strainer (0.0355 ft/s) and debris-laden (0.0136 ft/s) vortex tests. T a bl e 3 .. - f 3 1 Pl an t St ram . er A verage A~pproac hVe Ioc1"f1es RHR Strainer CS Strainer Flow Rate (gpm) 3,700 2,600 2 Surface Area (ft ) 677.6 590 Approach Velocity (ft/s) 0.0122 0.0098 Table 3.g.1-3 presents the minimum strainer submergences for different breaks at different times following the accident. As shown in the table, for the following four cases, the strainer submergence is greater than 3 inches after the start of sump recirculation. The debris-laden vortex test showed that, even with all debris loaded to the strainer, no vortices were observed for submergences greater than 3 inches. It is reasonable to conclude that vortexing will not occur for these four cases because at the start of recirculation 3 , the strainer is expected to be clear of debris.

• SBLOCA without CS

       * . MBLOCA without CS 3

The start of recirculation for the breaks that actuate containment spray is the time when the RWST level reaches the Low-Low level alarm and the sump suction valves for the RHR pumps open. For the breaks that do not actuate containment spray, start of recirculation is when the switchover of the RHR pump suctions from the RWST to sump is completed at the Empty level alarm. ES-63

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary)

• LBLOCA without CS
• Reactor nozzle break without CS For an LBLOCA with CS, the minimum strainer submergence is 0.098 ft (or 1.2 inches) at the start of recirculation when the strainer is still clean. Since the clean strainer vortex showed no vortices even for a partially submerged strainer, it is concluded that vortexing is not a concern for this case at the start of recirculation.

Table 3.g.1-3 shows that, for an LBLOCA with CS , the minimum strainer submergence increases to 1.803 ft when the switchover to sump recirculation is completed, which occurs approximately 20 minutes after start of recirculation. Since this submergence is much higher than the 3-inch limit identified by the debris-laden vortex test and only small amount of debris is expected to be transported to the strainer during the period of time it would take to reach a submergence of 3 inches, vortexing will not be an issue at debris-laden conditions for an LBLOCA with CS. Lastly, for LBLOCA reactor nozzle breaks with CS, Table 3.g.1-3 shows that the strainer is partially submerged at the start of recirculation. After that, the minimum strainer submergence increases over time and is equal to 0.35 ft (or 4.2 inches) when switchover to sump recirculation is completed, which is approximately 20 minutes after the start of recirculation. Similar to the discussion presented above for the LBLOCA with CS, the reactor nozzle break with CS is also bounded by the debris-laden vortex test with respect to formation of vortices. Based on the discussion above, vortexing is not a concern for any of the analyzed break scenarios.

4. Provide a summary of methodology, assumptions, and results of prototypical head loss testing for the strainer, including chemical effects. Provide bases for key assumptions.

Response to 3.f.4: Head loss tests were performed to measure the head losses caused by conventional debris (fiber and particulate) and chemical precipitate debris generated and transported to the sump strainers following a LOCA. The test program used a test strainer, debris quantities, and flow rates that were prototypical to VEGP. Different I . test cases were performed with the thin bed and full debris load protocols, following the 2008 NRC Staff Review Guidance (Reference 111 ). The results of the head loss tests provided a matrix of head loss data for various combinations of conventional and chemical debris loads. This matrix was used in the NARWHAL CFP calculation to determine the debris head loss for the debris load associated with each postulated break. ES-64

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Test Setup The test strainer assembly consists of seven stacked disks that are duplicates of the disks in the plant strainer. The top surface of the top disk and bottom surface of the bottom disk are solid steel rather than perforated plate . This results in a total of six disks contributing to the effective surface area of the test strainer. The test strainer was placed in a corner of a 6 ft tall by 6 ft wide by 10 ft long test tank on top of a horizontal plenum that simulated the plenum configuration present in the plant. The gaps between the test strainer and the surrounding walls of the test tank simulated the configuration of the plant strainer. See Figure 3.f.4-1 for an illustration of the test strainer and tank. Figure 3.f.4-1: Isometric of Head Loss Test Strainer Assembly inside Test Tank A schematic piping diagram of the test loop is provided in Figure 3.f.4-2 below. The test loop had a recirculation pump that took suction from the plenum underneath the test strainer and returned the water back into the test tank . The return flow exit into the tank was located such that the turbulence from the flow did not affect the debris bed on the test strainer. A flow element was used to measure the flow rate through the loop . Flow control valves and heating and cooling loops were used to control the test flow rate and water temperature . The test water was maintained at temperatures of at least 80 degrees F throughout the tests. ES-65

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) T11nk H atcr

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 ._ ____ ___I TT-VOi Test                     Simulated Strainer             Containment Floor 0              Module                    and Walls FS*VOO TT.V02 Tank Chiller     Flow
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2500 gpm Pump Figure 3.f.4-2: Piping Diagram of Head Loss Test Loop Test Parameters and Scaling The test strainer replicates all hydraulic dimensions of the plant strainer except for the number of strainer disks and the number of gaps between disks . The test debris quantities and test flow rate were scaled from plant values based on the ratio between the numbers of gaps between disks of the test strainer to that of the plant strainer. This is analogous to scaling the debris loads and flow rate based on the ratio of the test strainer surface area to the plant strainer surface area . The surface area of the test strainer was calculated to be 65 .57 ft2 . This surface area of the test strainer was scaled from the RHR strainer surface area based on the number of disks. The post-modification RHR strainer at the plant consists of four stacks of 15.5 disks (the top surface of the top disk for each stack is solid steel , so the top disk counted as ~ a disk) . The test strainer has a similar configuration as the plant strainer except the test strainer has only seven disks with the top surface of the top disk and bottom surface of the bottom disk being solid steel. Therefore, the active surface area of the test strainer is equivalent to six disks . This simple scaling method is reasonable because the test strainer disks and spacer rings were fabricated to the same dimensions as the disks and spacer rings installed in the plant. ES-66

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) To scale debris loads from the test to the plant, the debris load is multiplied by the ratio of the plant strainer area to the test strainer area (shown above). The 16-disk RHR strainers have a surface area of 677.6 ft2, and the 14-disk CS strainers have a surface area of 590 ft2 (see Table 3.f.3-1 ). This scaling is used when determining the strainer debris limits at the plant scale in Table 3.f.5-1. During VEGP head loss testing, a nominal test flow rate of 400 gpm was used. Therefore, using the test strainer surface area shown above, the average approach velocity of the test strainer was 0.0136 ft/s, which bounds the approach velocities of the current plant RHR and CS strainers, as shown in Table 3.f.3-1. In the NARWHAL CFP calculation, the measured strainer head losses from the 2009 testing were corrected from the testing conditions (e.g., strainer approach velocity and water temperature) to plant conditions of interest using the flow sweep data collected during testing. See response to 3.f.10. Debris Materials and Preparation The following materials were used as conventional debris for head loss testing: Nukon, lnteram E-54A, green silicon carbide powder, and silica sand. The method of preparation prior to introduction to the test tank for each material is discussed below. Nukon fines were used as surrogate for latent fiber, as recommended in NEI 04-07 and associated NRC SE (References 2 and 3, respectively). Nukon was also used to represent fines and small pieces of LDFG insulation debris. To prepare Nukon fines, Nukon fiberglass sheets were first shredded and inspected to ensure that the shredded Nukon met the size distribution requirements defined in NUREG/CR-6808 (Reference 39). Afterward, the required quantity was weighed out and boiled for 10 minutes to remove the binder. The boiled fiber was then placed in a bucket of water that was within +/-10 degrees F of the testing water temperature. The fiber was mixed thoroughly with a paint mixer attached to an electric drill until a homogeneous slurry was formed. Prepared fiber fines consisted of Class 1-3 fibers as defined in NUREG/CR-6224 (Reference 29). For small pieces of Nukon, the preparation was similar. However, the prepared small pieces consisted of interwoven strands of fiber, equivalent to or smaller than the Class 4 small fiber clusters, as defined in NUREG/CR-6224. lnteram fire blanket was processed (double-shredded) through a leaf shredder, similar to the manner in which the fiber debris was shredded. After shredding, the lnteram was added to buckets with sufficient water to suspend the debris. The buckets were then stirred to wet and suspend the lnteram debris in the bucket. Silica sand was used as a surrogate for latent particulate debris and was prepared by Performance Contracting, Inc. (PCI). The size distribution of the silica sand was ES-67

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) prepared to be consistent with that of latent particulate debris provided in the NRC SE for NEI 04-07 (References 3). Green silicon carbide powder was used as a surrogate for both qualified and unqualified coatings. Since the density of the green silicon carbide powder and that of the actual coating are different, the mass of the surrogate material added was adjusted such that the volumes of the surrogate material and actual coatings debris were matched. Per NEI 04-07 and the associated NRC SE (References 2 and 3, respectively), the coatings particulate debris was assumed to be 10 µm diameter spheres. The majority of the silicon carbide surrogate had a size distribution range from 4 µm to 20 µm, a median size of 10.25 µm, and a mean size of 10.46 µm. The required amount of silicon carbide and silica sand was weighed out and placed in a bucket of water with a temperature within +/-10 degrees F of the testing temperature. The particulate was mixed with water using an electric paint stirrer until no agglomeration or clumping was observed. Before introducing the particulate to the test tank, all particulate batches were mixed once again with an electric paint stirrer to create a thin slurry. Two types of chemical debris surrogates were used for the head loss testing: sodium aluminum silicate (SAS) and calcium phosphate. The chemical debris was prepared in accordance with WCAP-16530-NP-A (Reference 73). The 1-hour settling volume for each batch of chemical precipitates was determined at the time the batch was produced. The chemical precipitate settling time was also measured within 24 hours from the time the surrogate was to be used. Chemical precipitates that did not meet the settling requirements were discarded and not used for testing. Debris Introduction Debris was added at the side of the tank adjacent to the return flow exit into the tank and away from the simulated sump floor and walls. This allowed for an even and representative debris accumulation on the test strainer. A sparger system was installed on the return flow exit in the tank away from the strainer to aid in suspension of debris. Two mechanical mixers were also installed at the tank corners opposite from the test strainer. During the thin-bed test, all of the debris was added over the mixers while the recirculation pump was running. Additional manual agitation was applied to help the lnteram arrive near the test strainer. During the later full-load tests, the lnteram was introduced to the floor area immediately adjacent to the test strainer to aid the transport. The agitation from the sparger, mechanical mixers, and manual agitation helped keep the debris suspended in the water, and near-field settling was not credited. The debris bed formed on the strainer was not affected by debris addition or agitation in the tank. ES-68

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) After conventional debris introduction was completed for each test, chemical precipitate debris was added to the test tank. Calcium phosphate was first introduced in batches, and the head loss was allowed to stabilize between batches. SAS batches were added last. Head Loss Test Cases and Results Three head loss tests were performed for VEGP: one thin bed test and two full debris load tests. The two full-load tests targeted the same flow conditions and debris loads. For the thin bed test (VOG-1-TB), all of the particulate debris (including lnteram, coatings surrogate, and latent particulate surrogate) was introduced to the test tank at the beginning of the test. Once all of the particulate debris was added and allowed to circulate through the test loop, fine fiber batches were incrementally added in batch sizes equivalent to a 1/8-inch theoretical uniform debris bed thickness. Only fiber fines were used for the thin bed test. Thin-bed formation was observed visually, via head loss and turbidity measurement. The head loss was allowed to stabilize (less than or equal to 1 percent change over a 1-hour period) after the final batch of fiber fines was added. Once the head loss stabilized, chemical precipitates were incrementally added. The debris batch composition and size for the thin bed test are summarized in Table 3.f.4-1. Table 3.f.4-1: Debris Batches Added for the Thin Bed Head Loss Test Test Nukon Test !Test Silicon Test Test Calcium Quantity lnteram Carbide Dirt/Dust Phosphate Test SAS Batch (lbm) Quantity Quantity Quantity Quantity Quantity (lbm) (lbm) (lbm) (L) (L) Fines Smalls VOG-1.2-TB-P 0 0 29.15 358.42. 5.3 0 0 VOG-1.3-TB-F1 1.49 0 0 0 0 0 0 VOG-1.4-TB-F2 1.49 0 0 0 0 0 0 VOG-1.5-TB-F3 1.49 0 0 0 0 0 0 VOG-1.6-TB-F4 1.49 0 0 0 0 0 0 VOG-1.7-TB-F5 1.49 0 0 0 0 0 0 VOG-1.10-TB-CP1 0 0 0 0 0 160.24 0 VOG-1.11-TB-CP2 0 0 0 0 0 160.24 0 VOG-1.12-TB-CP3 0 0 0 0 0 160.24 0 VOG-1. 13-TB-NAS 1 0 0 0 0 0 0 122.86 VOG-1.14-TB-NAS2 0 0 0 0 0 0 122.86 VOG-1.15-TB-NAS3 0 0 0 0 0 0 122.86 Total 7.45 0 29.15 358.42 5.3 480.72 368.58 E5-69

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Table 3.f.4-2 summarizes the stabilized head losses after adding each batch of debris during the thin bed test. Table 3.f.4-2: Thin Bed Head Loss Test Results Stabilized Head Loss Batch (ft-H20) VOG-1.2-TB-P 0.171 VOG-1.3-TB-F1 0.226 VOG-1.4-TB-F2 0.262 VOG-1.5-TB-F3 0.308 VOG-1.6-TB-F4 0.368 VOG-1.7-TB-F5 0.625 VOG-1.10-TB-CP1 1.02 VOG-1.11-TB-CP2 1.54 VOG-1.12-TB-CP3 1.65 VOG-1.13-TB-NAS 1 2.12 VOG-1.14-TB-NAS2 2.27 VOG-1.15-TB-NAS3 2.56 For the full-load tests (VOG-2-FL-B and VOG-2-FL-82), the particulate and fiber debris was introduced simultaneously in equal batches maintaining the same fiber to particulate ratio until the full conventional debris load was reached. This debris addition sequence resulted in a homogenous debris bed accumulation. Each debris batch consisted of Nukon fiber fines and small pieces, and particulate debris. After the final batch of conventional debris was introduced, the head loss was allowed to stabilize with a less than 1 percent change over a 1-hour period. Chemical precipitates were then incrementally added. The debris batch composition and size for the full-load tests are summarized in Table 3.f.4-3. I , E5-70

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Table 3.f.4-3: Debris Batches Added for the Full-load Head Loss Tests Test Nukon Test Test Silicon Test Test Calcium Test SAS Quantity lnteram Carbide Dirt/ Dust Phosphate Quantity Batch (lbm) Quantity Quantity Quantity Quantity (L) Fines Smalls (lbm) (lbm) (lbm) (L) VOG-2.2-FL-F1 4.58 2.04 7.29 89.61 1.32 0 0 VOG-2.3-FL-F2 4.58 2.04 7.29 89.61 1.32 0 0 VOG-2.4-FL-F3 4.58 2.04 7.29 89.61 1.32 0 0 VOG-2.5-FL-F4 4.58 2.04 7.29 89.61 1.32 0 0 VOG-2.6-FL-CP1 0 0 0 0 0 160.24 0 VOG-2. 7-FL-CP2 0 0 0 0 0 160.24 0 VOG-2.8-FL-CP3 0 0 0 0 0 160.24 0 VOG-2.9-FL-NAS1 0 0 0 0 0 0 122.86 VOG-2.1 O-FL-NAS2 0 0 0 0 0 0 122.86 VOG-2.11-FL-NAS3 0 0 0 0 0 0 122.86 Total 18.32 8.16 29.16 356.64 5.28 480.72 368.58 As discussed above, two full-load head loss tests were performed. However, the first full-load test VOG-2-FL-B reported higher head losses. Table 3.f.4-4 summarizes the stabilized head losses after adding each debris batch during test VOG-2-FL-B. The measured head losses from the full-load test are much higher than the thin-bed test. Table 3.f.4-4: Bounding Full-load Head Loss Test Results Stabilized Head Loss Batch (ft-H20) VOG-2.2-FL-F1 0.276 VOG-2.3-FL-F2 1.06 VOG-2.4-FL-F3 2.42 VOG-2.5-FL-F4 5.46 VOG-2.6-FL-CP1 5.29 VOG-2.7-FL-CP2 6.22 VOG-2.8-FL-CP3 6.57 VOG-2.9-FL-NAS 1 7.16 VOG-2.1 O-FL-NAS2 7.24 VOG-2.11-FL-NAS3 11.81 E5-71

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary)

5. Address the ability of the design to accommodate the maximum volume of debris that is predicted to arrive at the screen.

Response to 3.f.5: The 2009 head loss test program evaluated debris loads based on the Nukon debris quantities calculated using a 7D ZOI from WCAP-16710-P, which was later rejected by the NRG. The current Nukon debris quantities were calculated with a 17D ZOI in BADGER, and this results in fiber debris loads greater than that tested. Therefore, the debris quantities used in the 2009 test program do not bound what is predicted for some of the breaks using BADGER. Debris limits were implemented in the NARWHAL analysis for each strainer in operation for a given scenario. The debris limits were applied to individual strainers, not to the total amount of transported debris. If the debris on the strainer exceeded any of the debris limits, a failure was recorded for that postulated break. The debris limits are based on the maximum quantity of conventional and chemical debris that was tested in 2009. Table 3.f.5-1 shows the debris limits for each of the debris types at the test scale and plant scale for one RHR strainer. Note that the debris limits at the plant scale are determined by multiplying the debris limits at the test scale by the ratio of the RHR strainer area (677.6 ft2) to the test strainer area (65.57 ft 2 ), as discussed in the response to 3.f.4. Note that the calcium phosphate debris and SAS debris are converted from volume to mass using concentrations of 5 g/L and 11 g/L, respectively. Table 3.f.5-1: Debris Limit Failure Criteria Debris Limit at Debris Limit Debris Type Plant Scale for One at Test Scale RHR Strainer Fiber 11.03 ft3 113.98 ft 3 Particulate 363.72 lbm 3758.68 lbm 4 Fire Barrier (particulate) 20.41 lbm 210.92 lbm Fire Barrier (fiber) 3.65 ft3 37.72 ft 3 Calcium Phosphate 5.30 lbm 54.77 lbm Sodium Aluminum Silicate 8.94 lbm 92.39 lbm 4 Fire barrier debris was treated as 70% particulate and 30% fiber. E5-72

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary)

6. Address the ability of the screen to resist the formation of a "thin bed" or to accommodate partial thin bed formation.

Response to 3.f.6: The "thin-bed effect" is defined as the relatively high head losses associated with a low-porosity (or high particulate to fiber ratio) debris bed formed by a thin layer of fibrous debris that can effectively filter particulate debris. The 2009 VEGP head loss testing included a test for thin-bed effects. During this test, the full particulate load was added into the test tank first, followed by fiber fines in batches equivalent to a 1/8-inch theoretical uniform bed thickness. This batching schedule allowed the formation of a debris bed with a high particulate to fiber ratio. As a result, any thin-bed effects, should they occur, would be captured by the measured head losses. As discussed in Section 3.f.10, head loss testing results from the thin bed test are used by the NARWHAL CFP calculation for postulated debris loads less than or equal to 3.1 ft3 of fiber at test scale. See Section 3.f.10 for additional discussion of how the total head loss is determined.

7. Provide the basis for strainer design maximum head loss.

Response to 3.f.7: There are several failure criteria evaluated by NARWHAL based on the head loss across the strainer: strainer structural margin, strainer debris limits, strainer partial submergence limits, void fraction limits, flashing, and pump NPSH. A postulated break that exceeds one or more of these criteria for the RHR strainers/pumps is considered to be a failure of the ECCS system. Each of the failure criteria was evaluated at each time step within the NARWHAL model to determine if an ECCS system failure would occur. Strainer Structural Margin Limits The strainer structural margin for each strainer is 24.0 ft. The head loss across each of the RHR and CS strainers due to conventional and chemical debris loading is compared to this value to ensure that the structural margin was not exceeded. See Section 3.k.1 for additional information on how the structural margin was determined. Debris Limits See Section 3.f.5 for discussion of conventional and chemical debris limits used in NARWHAL. ES-73

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Unsubmerged Strainer Limits If the strainers are partially submerged, the NARWHAL CFP calculation assumed that the strainer would fail if the head loss across the debris bed and strainer is equal to or greater than half of the submerged strainer height per RG 1.82 (Reference 25). Note that the pump NPSH and strainer structural limits are also applicable for a partially submerged strainer. The NARWHAL CFP calculation tracks time-dependent accumulation of debris on the strainer. When the strainer is partially submerged, the evaluation only credits the active (i.e., submerged) portion of the strainers for flow and debris accumulation. Void Fraction Limits A pump failure due to degasification was recorded if the steady state gas void fraction at the pump is greater than 2 percent by volume. Note that bubble compression was not credited. Flashing Failure Limits A flashing failure was recorded for a postulated break if, at any time during sump recirculation, the pressure downstream of the strainer was lower than the vapor pressure at the sump temperature. The pressure downstream of the strainer was calculated by NARWHAL based on the strainer submergence, containment pressure and head loss across the strainer. Note that a small increase in containment pressure was credited in the flashing analysis, see Section 3.f.14 for additional information. Pump NPSH Limits A pump failure was recorded if the head loss across the strainer exceeded the clean strainer NPSH margin. (i.e., margin that is available for debris laden head loss). It should be noted that because the safety injection (SI) pumps and centrifugal charging pumps (CCPs) take suction from the RHR pumps during recirculation, only the NPSH margins of the RHR and CS pumps were calculated in NARWHAL. See Section 3.g.16 for details of the NPSH margin used in NARWHAL.

8. Describe significant margins and conservatisms used in head loss and vortexing calculations.

Response to 3.f.8: Vortexing Testing Testing was conducted to determine if vortexing is expected to occur. As discussed in the Response to 3.f.3, the vortex tests were performed at both clean strainer and debris-laden conditions. ES-74

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) All vortex tests used strainer approach velocities higher than those expected for the plant strainer (0.0122 ft/sand 0.0098 ft/s for the RHR and CS strainers, respectively, see Table 3.f.3-1 ). The clean strainer vortex tests used strainer approach velocities up to 0.0258 - 0.0355 ft/s. For the debris laden vortex tests, a strainer approach velocity of 0.0136 ft/s was used. As shown in the response to 3.f.3, plant strainer minimum submergence at the start of the recirculation is compared with the submergence limit established by the debris-laden vortex tests. It should be noted that these tests were performed after all conventional and chemical debris has been added to the test tank. This is conservatively bounding because, at the start of recirculation, the strainer is expected to be clear of debris. Strainer Head Loss The quantity of latent debris used to determine the strainer head loss is 200 lbm, but the actual amount of latent debris documented for the plant is only 60 lbm. Similarly, the amount of miscellaneous debris used in the analysis is 50 ft2 , but, as stated in Response to 3.b.5, the amount of miscellaneous debris was conservatively assumed in the debris generation calculation to be 4 ft2, which bounds the 2 ft2 identified in containment during the walkdown. When correcting the debris head loss from the test conditions (e.g., water temperature and strainer approach velocity) to plant conditions, head loss coefficients from both the full debris load test and thin bed test were applied in the analysis. The coefficients that result in the higher head loss are used to calculate the debris head loss. Additionally, specific flow sweep data points were excluded from the analysis if the use of the points would result in lower corrected head loss values. See section 3.f.10 for additional discussion. Finally, the rule-based approach described in Section 3.f.10 conservatively applies the maximum head loss result from the thin bed test for all postulated breaks with minimal fiber debris generation. Similarly, all postulated breaks that result in debris at the strainer greater than the thin bed test and less than or equal to the maximum amount of debris tested in the full load test have the maximum head loss from the full load test.

9. Provide a summary of methodology, assumptions, bases for the assumptions, and results for the clean strainer head loss calculation.

Response to 3.f.9: The clean strainer head loss was calculated to be 4.40 in-H20 at the RHR pump runout flow rate of 4,500 gpm. This flow rate conservatively bounds the flow rate of the RHR and CS strainers. ES-75

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) The clean strainer head loss was calculated by modeling flow through one 18-disk strainer stack per the following steps:

• The cross-sectional areas of flow for various parts of the strainer stack were calculated from the physical dimensions of the strainer components.
• Loss coefficients were calculated for the flow paths through the strainer components based on flow path geometry. Loss coefficients were determined for the perforated plate, wire cloth, and converging cross flow from the flow through the disk with flow through the core tube.
• A system of mass balance and energy balance equations were iteratively solved to calculate the flow and resulting pressure drop for each disk in the stack.
• The difference between the initial pressure and the pressure of the fluid before entering the plenum was calculated and reported as the head loss through the strainer stack.
• The head loss inside the sump pit below the strainer stacks was also included.

For each pit, the flow through the four strainer stacks combines inside the space below the strainer assembly and upstream of the ECCS or CS suction pipe openings. Head losses due to flow exiting the strainer stacks and turning inside the pit were accpunted for based on conservative loss coefficients and velocities. Several assumptions were used when applying the above methodology to determine the clean strainer head loss. The temperature of water was assumed to be 120 degrees F, the strainer was assumed to be fully submerged (i.e., flow is through all disks), and head loss along the outside face of the disks, elevation head, and coupling effects were ignored. Additionally, friction loss between and within the strainer disks was ignored as it is negligible compared to the screen and perforated plate losses. The surface friction loss in the strainer core was also ignored because the radial influx of water from the strainers disks and spacers prevents significant flow and friction loss along the surface of the strainer core. The flow turn inside the pit after exiting a strainer stack was conservatively assumed to be confined in a 90° steel mitre bend. It is acceptable to use the clean screen head loss calculated for an 18-disk strainer as the clean screen head loss for a 16-disk strainer for the following reasons:

• The flow distribution in the clean screen head loss calculation shows that the vast majority of flow is through the first six disks closest to the pump suction.

In fact, only 0.1 percent of the total flow comes from the top two disks.

• A flow rate of 3,700 gpm through the RHR strainer is used in the NARWHAL analysis rather than the pump runout flow rate of 4,500 gpm used in the clean screen head loss calculation. The clean screen head loss at 3,700 gpm is less than the clean screen head loss at 4,500 gpm.

ES-76

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary)

10. Provide a summary of methodology, assumptions, bases for the assumptions, and results for the debris head loss analysis.

Response to 3.f.10: The total head loss across the strainer is the sum of the clean strainer head loss, the conventional debris head loss, and the chemical head loss. The conventional and chemical head loss values were based on VEGP-specific head loss test results that were corrected from the test conditions (i.e., strainer approach velocity and water temperature) to plant conditions. Additionally, the test results were extrapolated to the end of the 30-day strainer mission time. Debris Head Loss Correction A head loss correction factor (based on the strainer approach velocity and pool temperature) was implemented into NARWHAL to scale the measured head losses from test conditions to plant conditions. For each time step for which conventional and chemical head losses are evaluated, the head loss value is corrected based on the plant flow rate through the strainer and the pool temperature. The correction was performed based on the debris bed characteristics obtained through flow sweeps conducted during head loss tests. For the 2009 test program, flow sweeps were performed at the end of the thin bed and full-load tests. To account for the uncertainty in the flow sweeps for each test, the resulting correction parameters from both tests were applied at each time step, and the maximum resulting head loss was returned. Table 3.f.10-1 and Table 3.f.10-2 show the flow sweep data for the thin bed and full load tests, respectively. Note that the thin bed test flow sweep was conducted at a water temperature of 86 degrees F, and that the full load test was conducted at a water temperature of approximately 93 degrees F. The flow sweep data was used to determine the correction parameters, which w then used to scale measured head losses. The volumetric flow rates documented in Tables 3.f.10-1 and 3.f.10-2 were converted to approach velocities using the test strainer area of 65.57 ft2. Table 3.f.10-1: Thin Bed Test Flow Sweep Data Flow Rate Head Loss Approach Velocity (gpm) (ft-H20) (Ws) 403 2.6 0.0137 371 2.37 0.0126 200 1.31 0.0068 436 2.87 0.0148 395 2.56 0.0134 E5-77

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Table 3.f.10-2: Full Load Test Flow Sweep Data Flow Rate Head Loss Approach Velocity (gpm) (ft-H20) (ft/s) 393 11 .81 0.0134 369 10.61 0.0125 200 3.47 0.0068 446 11 .57 0.0152 395 8.81 0.0134 Figure 3.f.10-1 and Figure 3.f.10-2 show the debris head loss as a function of approach velocity for the thin bed and full load test flow sweeps , respectively. In addition , the data was fit with a second -order polynomial in the following form : Here, K1 and K2 are the fitting coefficients (as shown in the figures) , and v is the strainer approach velocity in fUs . Note that the polynomial was forced through the origin , because the head loss would be zero at an approach velocity of zero. Also , note that for the full load test, the test data points represented by the two orange points were excluded from the curve fit. This produces a conservative curve fit because it results in a higher predicted head loss for lower approach velocities. 3.5 3 y = 240.89x2 +187 .72x R' =0.9995 2.5

        -~      2                                                                                                    .. **

1:1! **** ***** _g ..*.... **

         ~    1.5 I                                                       ........ .**********

1 .... *** 0.5 1 . .* ********************** 0 .* ************* - 0 0 .002 0.004 0.006 0.008 0.01 0.012 0.014 0.016 Average Approach Velocity {ft/s) Figure 3.f.10-1: Head Loss Fitting Coefficients for Thin Bed Test E5-78

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) 14 y = 57252x 2 +123.56x 12 2 R = 0.9999 ..*** *

         ~

3 VI VI 10 8

         ~C1I 6                                                                                       ... ***

I .** 4 2 ... **** 0 . ........... *** 0 0.002 0.004 0.006 0.008 0.01 0.012 0.014 0.016 Average Approach Velocity (ft/s) Figure 3.f.10-2: Head Loss Fitting Coefficients for Full Load Test With these curves defined , the head loss fitting parameter can be calculated . NARWHAL accepts a, b , and flow sweep head loss as inputs into the following correction equation . X _ a X µ X Vstrainer + b xp X v~tra i ner HL - liPHL Nomenclature : XHL = Head loss correction factor Vstrainer = Approach velocity of the strainer at plant condition , ft/s a = Coefficient determined from flow sweep curve fitting parameter K 2 and water viscosity at test temperature , K2 /µ b = Coefficient determined from flow sweep curve fitting parameter K1 and water density at test temperature , K1 /p

µ                 =       Viscosity of water at plant condition , lbm/(ft-s) p                 =       Density of water at plant condition , lbm/ft 3 LlPH L            =       Head loss at the test approach velocity and temperature , and the flow sweep debris load , ft-H20 The thin bed head loss test was conducted at a temperature of 86 degrees F, which corresponds to a water density of 62 .16 lbm/ft3 and a water viscosity of 0.000536 lbm/(ft-s). The head loss correction coefficients were calculated as follows for the thin bed test flow sweep data :

ES-79

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) K2 187.725 52 *ft a= - = = 350,223 .9 - -

     µ     0.000536 lbm                        lbm ft

• 5 52 Ki 240.89 ft 52 . ft 2 b=-= =3.875--

p 62.161~1;'1 lbm LlPHL = 2.6 ft The full load head loss test was conducted at a temperature of 93 degrees F, which corresponds to a water density of 62.08 lbm/ft3 and a water viscosity of 0.000494 lbm/(ft-s) . The head loss correction coefficients are calculated as follows for the full load test flow sweep data : K2 123.56 5 52 *ft a= - = lb = 250,121.5 u;---

     µ    0.000494 ~                             m ft

• 5 52 Ki 57,252 ft 52 . ft2 b=- = lbm = 922.229 lbm p 62.08 ft3 LlPHL = 11.81 ft By substituting the values of a, b, and LlPHL into the formula above , a head loss correction factor XHL can be calculated for each set of flow sweep data. As stated earlier, the two correction factors calculated from the thin-bed and full debris load flow sweeps were multiplied by the total debris head loss at each time step and the higher resulting head loss was returned. Note that the total debris head loss is the sum of the conventional debris head loss, the chemical head loss , and the extrapolation constant (where applicable) . The clean screen head loss is not corrected to different strainer approach velocities or pool temperatures.

Debris Head Loss Extrapolation To address extrapolation of the head loss tests to the end of 30-day mission time , a head loss extrapolation constant was applied . The extrapolation constant was determined using the raw test data from the end of the head loss test. The raw test data was smoothed using a locally weighted least-squares method. The first order derivative of the smoothed data was reviewed to ensure that the slope of the data was trending towards zero, suggesting that the head loss profile was stabilizing . ES-80

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) A natural logarithmic function was fitted to the smoothed data and the function was shifted upwards , which bounded any peaks observed after the last chemical addition . The curve fit also had a similar slope (i .e. , rate of increase in head loss over time) as the recorded head losses at the end of the test. This head loss extrapolation was performed for the thin -bed test and both full debris load tests. Figure 3.f.10-3 shows the recorded head losses and the logarithmic curve fit before and after the adjustment for the first full debris load test. Note that the test data used for the extrapolation analysis was recorded at least 12 hours after adding the last batch of chemical debris. The extrapolation constant was calculated for all three tests . Since the constant of the first full debris load test (3.89 ft-H 20) is larger than the other tests , this value was conservatively used for all NARWHAL analysis . Note that this extrapolation constant is at the testing condition and is corrected to plant conditions using the same approach as the debris head losses (see discussions earlier in this response). The extrapolation constant was applied at 7.5 hours after the accident. 14 13.5 13

                                ..                     Adjusted head loss correlation 12.5      ***                  Head Loss" 12.9 + 0.42591.ln(Time) 12 Natural log curve flt to smoothed data
                                                                        .~                       -
                                                                  ~
                                                            ~

y = m1 + rn2 ' ln( mO) 11 Value Erra mi 10.644 0.028743 m2 0.42591 0.019719 10.5 Chl&q 3.9565 NA thed data R 0.90907 NA 10

• 0 2 3 4 5 6 7 8 Time [hrj Figure 3.f.10-3: Logarithmic Curve Fit of Bounded Test Data Used for 30-Day Head Loss Extrapolation Clean Strainer Head Loss The clean strainer head loss varies as a function of strainer approach velocity.

However, the bounding clean strainer head loss of 4.40 ft at 4 ,500 gpm was used for all cases in NARWHAL. No flow or temperature correction was applied to the clean strainer head loss. ES-81

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Conventional Debris Head Loss NARWHAL uses a rule-based approach to calculate head loss based on the results of head loss testing. As shown in Table 3.f.10-3, if the fiber debris load at the strainer is less than the tested quantity from the thin bed test (3.1 ft3 at test scale) , the maximum thin bed conventional debris head loss was returned. If the quantity was greater than what was tested in the thin bed test, the conventional head loss of the full-load test was returned . For a given time step, NARWHAL scaled the plant debris load to the test scale based on the active plant strainer surface area before determining the conventional debris head loss . Table 3.f.10-3: Conventional Head Loss Values Fiber Debris Load at Test Scale Head Loss (ft3) (ft-H20) s;3 .1* 0.625

                           >3 .1*                                      5.46 3                                              3
     *The 3.1 ft of fiber at test scale corresponds to 32.04 ft at the plant scale for one RHR strainer.

Chemical Head Loss A conservative chemical head loss model was implemented in the NARWHAL CFP calculation . The head loss effects of calcium phosphate and SAS were each analyzed separately from the 2009 head loss test results. Table 3.f.10-4 shows the head loss applied to each strainer once precipitate starts to accumulate on the strainer. Table 3.f.10-4: Chemical Head Loss Values Quantity Head Loss Chemical Precipitate (lbm) (ft-H20) Calcium Phosphate >O 1.11 Sodium Aluminum Silicate >O 5.24 Note that chemical head loss is not applied until a 0.45-inch thick theoretical uniform fiber debris bed has formed on the strainer. This approach is reasonable because , for fiber quantities smaller than this , large areas of open screen are present on the strainer. This is supported by the 2009 thin bed head loss test data . During the thin bed test, particulate debris was added to the test tank before fiber debris was batched in. Figure 3.f.10-4 shows a negligible increase in debris head loss for fiber loads up to and including 5.96 lbm . This fiber load corresponds to a uniform fiber bed thickness of 0.45 inches. ES-82

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) 0 .3 0£ 0.25

                 '+-

0

                !E.. 0 .2
                  ~

_,0 "C 0.15 ro (I) I

                 .!:   0 .1
                  ~

ro (I)

0.05 c

0 0.11 0.23 0 .34 0.45 0.57 Theoretical Fiber Bed Thickness (inches) Figure 3.f.10-4: Increase in Head Loss as a Function of Fiber Bed Thickness 11 . State whether the sump is partially submerged or vented (i.e ., lacks a complete water seal over its entire surface) for any accident scenarios , and describe what failure criteria in addition to loss of NPSH margin were applied to address potential inability to pass the required flow through the strainer. Response to 3.f.11: As shown in Table 3.g.1-3 , for some of the postulated breaks (specifically, reactor cavity breaks with CS actuated) , the strainers could be partially submerged at the start of recirculation for a short period but become fully submerged before the switchover to recirculation is completed . When strainers are not fully submerged, the unsubmerged strainer head loss failure criterion discussed in Section 3.f.7 was used. The NARWHAL CFP calculation showed that the head loss during the time when the RHR strainer is partially submerged does not challenge the failure criterion which states that head loss cannot be greater than half of the submerged strainer height for any of the break scenarios . The calculation evaluated the most limiting break in terms of fiber at the RHR strainer during the time that the RHR strainer is partially submerged . The DEGB at Weld 11201-001-1-RB (located in the reactor cavity on the hot leg) with a single train failure configuration resulted in the most amount of fiber on the RHRA strainer for all breaks that have partially submerged strainers during recirculation . For this break, the strainer is partially submerged for a total of 11 minutes. The amount of debris accumulated during this period of time did not challenge the failure criterion for partially submerged strainers. ES-83

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary)

12. State whether near-field settling was credited for the head-loss testing , and if so ,

provide a description of the scaling analysis used to justify near-field credit. Response to 3.f.12: No near-field settling was credited in head loss testing . Sufficient turbulence was provided in the tank to ensure that all debris had an opportunity to collect on the surfaces of the test strainer, while not disturbing the debris bed formation . Additionally, manual stirs were applied as necessary to prevent debris from settling after introduction . Two mechanical stirrers were required to suspend the debris due to the strainer configuration and flow rate , one in the pit below the strainer, and another within the area underneath the strainer, bounded by the simulated containment walls and floo r. A sparger system was installed on the return line to aid in suspension of debris. Additionally, a sump pump and attendant tub ing were used to provide flow from beneath the simulated containment floor to ensure that particulate debris did not accumulate there . Hand-stirring and manual adjustment of the mechanical stirrers was performed as necessary during the add itions of the fibrous and particulate debris , with much care and consideration given to avoid disturbing the bed or otherwise artificially influencing the bed formation .

13. State whether temperature/viscosity was used to scale the results of the head loss test to actual plant conditions. If scaling was used , provide the basis for concluding that boreholes or other differential-pressure induced effects did not affect the morphology of the test debris bed .

Response to 3.f.13 : Head loss values were scaled from test conditions to plant conditions using both temperature (i.e ., viscosity and density as a function of temperature) as well as strainer approach velocity. As shown in the response to 3.f.10, these equations were derived from VEGP-specific flow sweep data . During the 2009 head loss testing , flow sweeps were conducted at the end of each test to characterize the flow through a prototypical debris bed . Therefore , any boreholes and other differential-pressure induced effects on bed morphology were captured and properly accounted for when scaling the head loss . In addition , as stated in the NARWHAL CFP calculation , two sets of flow sweep data were collected following the thin -bed and full debris load tests . To account for the uncertainties in the flow sweeps , the resulting correction parameters from both tests were applied at each time step and the higher resulting head loss was used . E5-84

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary)

14. State whether containment accident pressure was credited in evaluating whether flashing would occur across the strainer surface, and if so , summarize the methodology used to determine the available containment pressure.

Response to 3.f.14: The NARWHAL software was used to evaluate the potential for flashing due to the pressure drop across the strainer and debris bed . For a given break, a flashing failure was recorded if, at any time during sump recirculation , the pressure downstream of the strainer was lower than the vapor pressure at the sump temperature . The pressure downstream of the strainer was calculated by NARWHAL based on the strainer submergence , containment pressure-, and head loss across the strainer. Note that, for flashing analysis , the strainer submergence is evaluated from the top of the strainer. As discussed in Enclosure 3, Section 6.7, up to 3.5 psi of accident pressure was credited in order to preclude flashing. This approach is reasonable , since, as shown below, even the smallest margin in the containment pressure for preventing flashing is higher than the 3.5 psi credited in the analysis. The margin in containment pressure for preventing flashing immediately downstream of the strainer is evaluated for time-dependent post-accident containment and sump conditions . For each given set of conditions , the sump pool temperature is obtained from the design basis profile evaluated for a double-ended reactor coolant pump (RCP) suction break with minimum safeguards. The strainer head loss at each given pool temperature is taken from the NARWHAL CFP calculation . The post-accident containment pressure is from the design basis profile evaluated for a double-ended RCP suction break with maximum safeguards. The minimum containment pressure that is required to prevent flashing is calculated by adding the strainer head loss to the water vapor pressure . Afterwards , this minimum required containment pressure is compared with the expected post-accident containment pressure to determine the margin . The evaluation contains the following conservatisms : When calculating the minimum containment pressure required to prevent flashing , the submergence of the strainer is conservatively neglected . Including the submergence would reduce the minimum pressure required and increase the margin .

• For sump temperatures above 212 degrees F, the strainer head loss is conservatively assumed to be the same as that at 212 degrees F. In reality, the head loss adjusted to the actual temperature would be lower due to the lower water viscosities at higher temperatures.

When determining the post-accident containment pressures from the Vogtle FSAR chart, the values are rounded down, which results in conservatively smaller margins. ES-85

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) As shown in Table 3.f.14-1 , the minimum margin in the containment pressure to prevent flashing is over 6 psi at 3,000 seconds after the accident. Therefore , crediting 3.5 psi of accident pressure in the NARWHAL CFP calculation for the flashing evaluation is reasonable and conservative . Figure 3.f.14-1 compares the design basis post-accident containment pressure with the minimum containment pressure required to prevent flashing . The vertical difference between the two curves represent the margin in containment pressure for preventing flashing . Table 3.f.14-1: Margin in Containment Pressure for Preventing Flashing based on OBA Curves Min Cont. Margin in Sump Pool Vapor Strainer Accident Time Pressure Req'd Containment Temperature Pressure Head Pressure (s) (oF) to Prevent Pressure (psia) Loss (ft) (psia) Flashing (psia) (psi) 1,800 251 30.66 5.515 32.90 39.4 6.5 3,000 249 29.47 5.515 31 .72 37.8 6.1 3,400 248 28.70 5.515 30.95 38.4 7.4 7,020 212 14.81 5.515 17.10 33.4 16.3 7,980 205 12.88 5.544 15.19 33.4 18.2 10,020 195 10.49 5.589 12.83 31.4 18.6 19,980 165 5.42 5.729 7.85 27.4 19.5 30 ,000 153 4.08 9.006 7.92 25.4 17.5 60 ,000 140 2.96 9.128 6.86 22.9 16.0 90 ,060 133 2.47 13.615 8.30 21 .9 13.6 500,460 120 1.74 13.826 7.68 19.4 11 .7 50

                                                                  - Min Containment Pressure to Prevent Flashing 40 +---.c--- - - - - + - - - - 1
                                                                  ---.- Deisgn Basis Accident ro                                                       Containment Pressure
               - ~ 30 + - - -....;:.;;111-----'"""";::--- - - - - - + - - - - - - - I 10 -+-------~'                 ___,,__ _ _ _ _-+------~

0 -+---'--'-~'-'--'-'-'-1---L---'-..__,_-'--'--'-'+-_...._..___._,_......_,__,~ 1.E+3 1.E+4 1.E+S 1.E+6 Time (s) Figure 3.f.14-1: Margin in Containment Pressure for Preventing Flashing based on OBA Curves E5-86

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) To demonstrate that margin also exists for breaks that have a lower pressure than the OBA profile, the same approach was also applied to the best-estimate post-accident containment pressure and sump pool temperature profiles to derive the margins in containment pressure for preventing flashing. Note that the best-estimate curves have lower values than the design basis curves because the thermal hydraulic modeling for the best-estimate cases used less conservative inputs . Figure 3.f.14-2 compares the best-estimate containment pressure curve for a double-ended guillotine cold leg break with the minimum pressure required for preventing flashing evaluated using the corresponding sump pool temperature profile. The vertical difference between the two curves is the margin for preventing flashing . The results show that the minimum margin is 9.3 psi at approximately 30,000 seconds after the accident. Note that, for the cold leg break, sump recirculation starts at 3,398 seconds after the accident. 40 ~---- --~--- 1 ----- Min Pressure to Prevent Flash ing

                                                                                                         ==;-1 i I 30 i--"
                      """""'o::--- -                                       - - - Best Estimate CLB          Y i
                                     ---+-------1
                    ~                                                    L         Containment Press ~

1

            ~ 20   r-------1-__::=--...:;;;;;:::::::::::::j:::::::=-----1 Ill Ill
            <ll 0...

10 L ............... !.... - .... ___ .. ,~- ------...... _.,-.. t I 0 ....._._"'"" 1.E+2 1.E+3 1.E+4 1.E+5 lime (s) Figure 3.f.14-2: Margin in Containment Pressure for Preventing Flashing based on Best-Estimate Cold Leg Break Curves Similar evaluations were also performed using the best-estimate hot leg break curves. The results are shown in Figure 3.f.14-3. The minimum margin is 8.7 psi at approximately 2,500 seconds after the accident. Note that, for the hot leg break, sump recirculation starts at 2,263 seconds after the accident. ES-87

Enclosure 5 Supplemen'tal Response to NRC Generic Letter 2004-02 (Nori-Proprietary)

• Min Pressure to Prevent Flashing

                                                 -       Best Estimate HLB Contai nment Pressure 0  +-~~~~~~-'-+--~~~~~~--'--'-<>----~~~~~~~

1.E+2 1.E+3 1.E+4 1.E+5 Time (s) Figure 3.f.14-3: Margin in Containment Pressure for Preventing Flashing based on Best-Estimate Hot Leg Break Curves In summary, for the best-estimate containment pressure and sump temperature curves , the minimum margin in containment pressure for preventing flashing is at least 8.7 psi. Therefore , crediting 3.5 psi of accident pressure in the NARWHAL CFP calculation for the flashing evaluation is reasonable and conservative. Note that containment pressure and sump temperature are intrinsically related. While the best-estimate containment pressures are lower than the design basis case , the corresponding pool temperatures are also lower, which results in lower pressures required for preventing flashing. The resulting margins in containment pressure for the best-estimate curves are either comparable or actually greater than those derived based on the design basis curves . ES-88

Enclosure 5 Supplemental Response to NRC Generic Letter 2004;02 (Non-Proprietary)

g. Net Positive Suction Head The objective of the NPSH section is to calculate the NPSH margin for the ECCS and CSS pumps that would exist during a LOCA considering a spectrum of break sizes.

1. Provide applicable pump flow rates, the total recirculation sump flow rates, sump temperature(s), and minimum containment water level.

Response to 3.g.1: Pump/ Sump Flow Rates ECCS and CS pump design flow rates used in the NARWHAL model are presented in Table 3.g.1-1. The total recirculation sump flow rates are provided in Table 3.g.1-

2. Note that the SI pumps and CCPs piggyback off of the RHR system during recirculation. Additionally, each of the RHR and CS pumps has its own dedicated sump and strainer.

Table 3.g.1-1: Applicable Pump Flow Rates Pump Design Flow Rate (gpm) RHR 3,700 SI 425 cc 150 cs 2,600 Table 3.g.1-2: Total Recirculation Sump Flow Rates Sump Design Flow Rate (gpm) RHR 3,700 cs 2,600 Minimum Water Level Minimum sump pool levels were calculated in the NARWHAL CFP calculation and in a bounding hand calculation. The NARWHAL calculation performs comprehensive evaluation of GSl-191 phenomena in a self-consistent and time-dependent manner. For each accident evaluated, the entire duration of RWST injection and sump recirculation was divided into smaller time steps. The minimum sump pool volume was calculated for each time step by subtracting the transitory and geometric hold up volumes from the total quantity of water in containment. The NARWHAL CFP calculation evaluated the NPSH margin for each break scenario. Impact on the results due to variabilities in the inputs was evaluated by sensitivity analyses. ES-89

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) The minimum water level hand calculation evaluated bounding minimum sump pool volumes and levels which were used as inputs in the vortexing evaluation (see the Response to 3.f.3) and chemical precipitate debris hand calculation (see the Response to 3.o.1 ). Table 3.g.1-3 summarizes the results of the minimum water level hand calculation. The short-term water level values (prior to 60 hours post-LOCA) and long-term water level values (at 60 hours post-LOCA) differ due to the transient inputs, such as RWST injection, reactor cavity hold-up, and containment temperature (which influences vapor hold-up). The submergence values in Table 3.g.1-3 were calculated by subtracting the RHR strainer height (4.438 ft, discussed in Response to 3.f.2) from the water level above the containment floor. A negative value for strainer submergence indicates the strainer is partially submerged. The submergence values in Table 3.g.1-3 bound the minimum submergence of the CS strainers because the CS strainers are shorter than the RHR strainer and switchover of CS pumps to recirculation occurs after the RHR pumps. The VEGP sump recirculation switchover evaluation showed that, for breaks that do not actuate CS, the ECCS pumps continue drawing all of their flow from the RWST until the Empty level setpoint is reached. In other words, for these breaks, recirculation will not start until the time labeled as "completion of switchover" in Table 3.g.1-3. Therefore, for the breaks that do not actuate CS, no strainer submergence values are shown at the time when the sump suction valves open. For those breaks that do actuate CS, the VEGP sump recirculation switchover evaluation showed that the ECCS pumps start to draw flow from the sump as soon as the sump suction valves open. Therefore, for these breaks, sump recirculation begins when the sump suction valves open at the RWST Low-Low level. ES-90

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Table 3.g.1-3: Minimum Sump Pool Water Levels from Hand Calculation Time of Pool Strainer Break Case Time Occurrence Height Submergence Description (sec) (ft) (ft) 1 Sump Suction Valves Open 1,929 4.536 0.098 LBLOCAwith Completion of Switchover 3,288 6.241 1.803 Containment Spray 5.5 Hours 19,800 6.058 1.620 60 Hours 216,000 5.311 0.873 Sump Suction Valves Open 18, 109 4.108 NIA SBLOCA without Completion of Switchover2 24, 137 5.739 1.301 Containment Spray 5.5 Hours 19,800 6.003 1.565 60 Hours 216,000 5.186 0.748 Sump Suction Valves Open 6,038 4.648 N/A MBLOCA without Completion of Switchover2 8,048 6.353 1.915 Containment Spray 5.5 Hours 19,800 6.318 1.880 60 Hours 216,000 5.501 1.063 Sump Suction Valves Open 3,655 4.692 N/A LBLOCA without Completion of Switchover2 5,104 6.407 1.969 Containment Spray 5.5 Hours 19,800 6.318 1.880 60 Hours 216,000 5.501 1.063 Sump Suction Valves Open 1 1,929 3.054 -1.384 Reactor Nozzle Break LBLOCA Completion of Switchover 3,288 4.788 0.350 with Containment 5.5 Hours 19,800 4.971 0.533 Spray 60 Hours 216,000 5.039 0.601 Sump Suction Valves Open 3,655 3.235 N/A Reactor Nozzle Break LBLOCA Completion of Switchover2 5,104 4.977 0.539 without Containment 5.5 Hours 19,800 5.161 0.723 Spray 60 Hours 216,000 5.229 0.791 Notes: 1 Beginning of recirculation for the breaks that actuate CS is when the RWST level reaches Low-Low setpoint and the sump suction valves for the RHR pumps open. 2 Beginning of recirculation for the breaks that do not actuate CS is when the switchover of the RHR pump suctions from the RWST to sump is completed. E5-91

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 {Non-Proprietary) Sump Temperature The VEGP NARWHAL CFP calculation used the design-basis sump temperature profile calculated for a double-ended pump suction LOCA with minimum safeguards. Note that the minimum safeguards temperature profile shown in Figure 3.g.1-1 is conservatively higher than the temperature profile for the maximum safeguards case. 300

  -u. 250 L

100 . *-*-* *~**- "" - . **-

                                                ~      ',,,_._  ......,.... *-*-***- --

1

                                                                      '~                     ~

50 '  ! I 1.E-01 1.E+01 1.E+03 1.E+05 1.E+07 Time (s) Figure 3.g.1-1: Sump Temperature for Double-Ended Pump Suction Break with Minimum Safeguards As discussed above, the recirculation duration was divided into smaller time steps. When applying the sump temperature profile, the value that is closest to the current time step is used. Consider an example where the current time-step is 220 seconds and the profile has values corresponding to 219 and 229 seconds. NARWHAL would return the value at 219 seconds because it is closer to the current time step.

2. Describe the assumptions used in the calculations for the above parameters and the sources/bases of the assumptions.

Response to 3.g.2: Pump/Sump Flow Rate As discussed in the VEGP NARWHAL CFP calculation, the RHR flow rate was assumed to be 3, 700 gpm. The design flow rate for the RHR pumps is 3,000 gpm. Using a higher flow rate is generally conservative in terms of recirculation timing, ES-92

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) flashing calculations, and head loss correction as performed in the NARWHAL CFP calculation. This 3,700-gpm flow rate is also consistent with the value used in the design-basis NPSH calculation and the single train value used in the ECCS system head curve. In the NARWHAL CFP calculation, the flow rates for the SI pumps and CCPs are their design flow rates based on the SI system description. Similarly, the CS pump flow rate is also the design flow rate from the CS system description. Using the design flow rates for these pumps is reasonable because they are more closely aligned to what would be expected in post-LOCA mitigation than run-out flow rates when taking into consideration system pipe losses. In the NARWHAL CFP calculation, the same pump flow rates were used consistently for all breaks regardless of break size. Using higher flow rates for the smaller breaks is conservative in terms of NPSH and flashing failures because recirculation starts sooner when the pool temperature is higher. Minimum Water Level As stated in the response to 3.g.1, minimum sump pool water levels were calculated in both the NARWHAL CFP calculation and a hand calculation. The major assumptions used in these evaluations are listed as follows.

1. The density of the inventory of the RWST, the reactor coolant system (RCS), and the SI accumulators is assumed to be the same as pure water. This is a reasonable assumption because the concentration of boric acid in the water is extremely small, with a maximum of 1,900 ppm for the RCS; 2,600 ppm for the RWST; and 2,600 ppm for the accumulators.

2. It is assumed that SBLOCAs will not result in rapid, full depressurization of the RCS; therefore, the SI accumulators will not inject when evaluating the minimum water levels for SBLOCAs. This is a conservative assumption because this will minimize the pool volume.

3. It is assumed that MBLOCAs and LBLOCAs will result in full depressurization of the RCS; therefore, during recirculation, the RCS will retain water up to the elevation of the break. This is a reasonable assumption because these breaks result in rapid cooling from the SI accumulators, which are triggered through RCS depressu rization.

4. The hand calculation reported minimum sump water levels for three break size categories, defined as follows. This definition matches that used in the Vogtle PRA model.

a. An SBLOCA is defined as a break smaller than 2 inches.

b. An MBLOCA is defined as a break greater than or equal to 2 inches, less than 6 inches.

c. An LBLOCA is defined as a break greater than or equal to 6 inches with the largest break being a double-ended guillotine break of the crossover leg.

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Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary)

5. It is assumed that a reactor nozzle break will cause the entire reactor cavity (up to the seal ring) to fill with water before any water reaches the containment floor.

This conservatively maximizes the transient reactor cavity hold-up, thereby minimizing the pool level.

6. By maximizing the vapor hold-up in the atmosphere of containment, water is withheld from the pool, thereby conservatively minimizing the pool level. To

   'maximize the vapor hold-up in the atmosphere of containment, three complementary assumptions were made.

a. The relative humidity within containment post-LOCA was maximized. A post-LOCA relative humidity of 100 percent was used to saturate the atmosphere in containment completely, thereby maximizing the change in water vapor in the air from pre-LOCA to post-LOCA conditions.

b. The relative humidity within containment pre-LOCA was minimized. A pre-LOCA relative humidity of 0 percent was used to obtain a pre-LOCA vapor hold-up of 0 gal, thereby maximizing the change in water vapor in the air from pre-LOCA to post-LOCA conditions.

c. The transient values used for containment temperature are maximum values, which result in maximum vapor pressures. These maximum vapor pressures maximize the vapor hold-up in the air.

7. The containment sprays were assumed to only be activated for hot leg breaks greater than 15 inches, which includes all partial breaks and DEGBs greater than 15 inches on the hot legs. However, no failures on the cold or intermediate legs were assumed to actuate containment sprays. This assumption is consistent with the results of best-estimate thermal-hydraulic modeling for a range of potential break sizes on the hot and cold leg piping. This modeling showed that a hot leg DEGB resulted in containment pressures exceeding the CS actuation setpoint of 21.5 psig, while all other evaluated breaks (including a cold leg DEGB and partial 15 inch breaks on both the hot and cold legs) did not. Assuming that hot leg breaks greater than 15 inches activate CS is reasonable because it represents what was learned from the best-estimate thermal hydraulic modeling.

It is recognized that there is some uncertainty in which breaks initiate CS. Sensitivity runs were therefore performed on actuation limits and spray duration using NARWHAL, as summarized in Enclosure 3.

8. The accumulators were assumed to not inject for any secondary side break. This is a reasonable assumption because secondary side breaks do not result in rapid depressurization of the RCS, which would trigger accumulator injection.

Sump Temperature The sump temperature profile used for the GSl-191 analysis was from the design-basis containment analysis for a DEGB on the crossover leg with minimum safeguards and 11.12 percent fan cooler degradation. This analysis was performed for evaluating post-LOCA containment integrity to support the VEGP Units 1 and 2 measurement uncertainty recapture power uprate program. Note that, in addition to the crossover leg break, the containment analysis also modeled the containment response following a main steam line break (MSLB). The results showed that the ES-94

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) crossover leg break with minimum safeguards resulted in higher sump temperatures than the MSLB and the crossover leg break with maximum safeguards.

3. Provide the basis for the required NPSH values, e.g., 3 percent head drop or other criterion.

Response to 3.g.3: The NPSH required (NPSHR) values were taken from the bounding pump vendor curves. These curves were obtained by the pump manufacturer through testing in accordance with the Hydraulic Institute guidelines in effect at the time. Typically, the 3 percent head drop criterion was used in pump NPSH testing.

4. Describe how friction and other flow losses are accounted for.

Response to 3.g.4: The verification of adequate NPSH margin to the RHR and CS pumps from the containment sump was performed using the NARWHAL model. For each time step, NARWHAL calculates the pump NPSH available (NPSHA), NPSHR, and strainer head loss using the inputs of that time step (e.g., sump water level, sump temperature, and pump flow rates). Note that the calculated NPSHA accounted for the piping head loss from the sump to pump suction but not the strainer head loss. If the NPSH margin, determined by subtracting NPSHR from NPSHA, is less than the total strainer head loss, a failure is recorded. The total strainer head loss was calculated by combining the clean strainer and debris bed head losses, and extrapolation constant as necessary. The head loss of the suction piping between the strainer exit and the pump suction was accounted for when calculating NPSHA. The piping frictional loss was calculated using the standard Darcy formula with the friction factor determined from an empirical equation. The head losses of the components (e.g., valves, elbows, reducers, and tee junctions) on the pump suction piping were calculated using the loss coefficients from standard industry handbooks.

5. Describe the system response scenarios for LBLOCAs and SBLOCAs.

Response to 3.g.5: In response to a LOCA, the RHR pumps, SI pumps, and CCPs automatically start upon receipt of an SI signal. These pumps take suction from the RWST and inject to the RCS cold legs. This system line-up is referred to as the ECCS injection phase. The CS pumps start automatically when the containment pressure reaches the high-pressure setpoint for CS actuation. The CS pumps also take suction from the RWST during the injection phase. When the RCS depressurizes to approximately 600 psia, all four accumulators begin to inject borated water into the RCS loops. E5-95

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Before the RWST inventory is depleted, the suction source of the pumps must be switched to the recirculation sumps. The surrip suction valves for the ECCS pumps open automatically when the RWST level reaches the Low-Low setpoint. The switchover for the CS pumps starts manually when the RWST level reaches the Empty setpoint. The switchover is complete when the suction valves from the RWST for all pumps are manually closed, which occurs between the RWST Empty and Dead Volume levels. For the breaks that do not actuate CS, the switchover to sump recirculation for the ECCS pumps follows the same logic. Approximately 7.5 hours following an accident, the ECCS line-up is modified for simultaneous cold and hot leg recirculation. For this operating mode, the SI pumps and CCPs continue taking suction from th.e RHR pump discharge. The RHR and SI pumps are aligned to supply flow to the RCS hot legs, but the CCPs continue supplying flow to the cold legs. The response sequence described above is typical for the ECCS and CSS following a LOCA. The differences between the responses to an LBLOCA and an SBLOCA are:

• Depending on the size of the break, the RCS pressure may stabilize at a value that does not allow injection from the SI accumulators and/or the RHR pumps.
• For an SBLOCA, the containment pressure will likely remain below the actuation setpoint for the CSS.

For an SBLOCA, the outflow from the RWST may be sufficiently low that the plant may be taken to a safe shutdown condition before the RWST level reaches the Low-Low setpoint. As a result, sump recirculation may not be required.

6. Describe the operational status for each ECCS and CSS pump before and after the initiation of recirculation.

Response to 3.g.6: Residual Heat Removal Pumps In the event of a LOCA, both RHR pumps are started automatically on receipt of an SI signal. During the injection phase, the RHR pumps take suction from the RWST and supply flow to the RCS cold legs. When the RWST level reaches the Low-Low setpoint, the suction valves to the sump automatically open. The RHR pumps could take suction simultaneously from the RWST and the containment sumps. After the RWST level reaches the Empty setpoint, the suction valves to the RWST are manually closed. Afterwards, the RHR pumps take suction from the sumps only. The RHR pumps continue to supply flow to the RCS cold legs and to the SI pumps and CCPs. ES-96

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Centrifugal Charging Pumps In the event of a LOCA, both CCPs start automatically on receipt of an SI signal and take suction directly from the RWST during the injection phase. The CCPs supply flow to the RCS cold legs. After switching to the sump recirculation phase, flow to the CCPs is provided from the RHR pump discharge. The CCPs continue supplying flow to the RCS cold legs during simultaneous cold leg and hot leg recirculation. Safety Injection Pumps In the event of a LOCA, both SI pumps start automatically on receipt of an SI signal. During the injection phase, these pumps take suction from the RWST and deliver water to the RCS cold leg. Similar to the CCPs, flow to the SI pumps is supplied from the containment emergency sump via the RHR pumps during the recirculation phase. Containment Spray System Pumps The CS pumps can be actuated manually from the control room or automatically on receipt of two out of four containment pressure (high-3) signals. These signals start the CS pumps and open the discharge valves to the spray headers. During the injection phase, the CS pumps take suction from the RWST. As discussed in the Response to 3.g.5, the pump suction is manually switched to the containment recirculation sump when the RWST level reaches the Empty setpoint.

7. Describe the single failure assumptions relevant to pump operation and sump performance.

Response to 3.g. 7: As described in Enclosure 3, Sections 4.0, 6.3, and 14.1, the VEGP risk-informed evaluation considered many different equipment configurations and wasn't limited to the worst single failure. The high likelihood configuration calculation used the VEGP PRA model of record, which accounts for human reliability analysis (HRA), and identified the following twelve equipment failure combinations.

1. No Equipment failure

2. RHR Pump 8 failure

3. RHR Pump A failure

4. Charging Pump A failure

5. Charging Pump 8 failure

6. SI Pump 8 failure

7. SI Pump A failure

8. Train 8 failure

9. Train A failure

10. CS Pump 8 failure ES-97

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary)

11. CS Pump A failure

12. Both CS Pumps failure Note that the VEGP inputs and NARWHAL methodology allow for this list to be reduced based on train symmetry. A pump failure for one train is analytically identical to a pump failure for the other train. Therefore, the following seven equipment configurations were analyzed in the NARWHAL CFP calculation.

1. No Equipment failure

2. RHR Pump B failure

3. Charging Pump B failure

4. SI Pump B failure

5. Train B failure

6. CS Pump B failure

7. Both CS Pumps failure It was assumed that all random equipment failures evaluated occur at the beginning of recirculation. This is a conservative assumption because it results in a quicker switchover to recirculation when compared to failure at the beginning of the event.

Additionally, for CS pump and/or RHR pump failure cases, it results in more debris accumulation on the remaining active strainers. The CCP B and SI pump B failure cases are identical to the no equipment failure case. This is because the failure is applied at the start of recirculation. The flow rate through the RHR strainers is not affected by the charging pump failure because the RHR pump provides the same flow rate regardless of which piggybacked pump fails. The CS pump B failure case is similar to the no equipment failure case. It only affects hot leg breaks greater than 15 inches. Thus, this case only has a slight effect on the results even though there is one less active strainer during recirculation.

8. Describe how the containment sump water level is determined.

Response to 3.g.8: As discussed in the response to 3.g.1, the post-LOCA minimum sump pool level was determined in both the NARWHAL CFP calculation and a hand calculation. The two calculations used the methodology described below:

1. A correlation was first developed for the relationship between the containment water level and the water volume using a 3-D CAD model.

2. The quantity of water added to containment from the RWST, RCS, and SI accumulators was calculated.

3. The quantity of water that is diverted from the containment sump by the following effects was evaluated:

• Hold-up within the reactor cavity and in-core tunnel.
• RCS hold-up volume required to fill the RCS steam space.

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Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary)

• Water volume required to fill the CS pump discharge piping that is empty pre-LOCA.
• Water in transit from the containment spray nozzles and the break to the containment sump.
• Steam hold-up in the containment atmosphere.
• Miscellaneous hold-up volumes throughout containment, such as containment sumps, the elevator pit, and containment floor drains.

4. Given the net mass of water added to the containment floor based on Items 2 and 3 listed above, the post-LOCA containment water level is calculated using the correlation developed in Item 1.

As discussed earlier, the NARWHAL CFP calculation used self-consistent inputs and evaluated time-dependent pool volumes and water levels for each postulated break. The hand calculation determined bounding minimum containment water levels for LBLOCA, MBLOCA, and SBLOCA and provided inputs for evaluating chemical precipitate debris quantities and vortexing. While the NARWHAL CFP calculation determines the water level at each time step within the simulation, the hand calculation only reported water levels at a few different times after the accident, as shown in Table 3.g.1-3.

9. Provide assumptions that are included in the analysis to ensure a minimum (conservative) water level in determining NPSH margin.

Response to 3.g.9: The assumptions provided in the Response to 3.g.2 ensure that minimum (conservative) containment water levels are calculated in the VEGP containment water volume calculation.

10. Describe whether and how the following volumes have been accounted for in pool level calculations: empty spray pipe, water droplets, condensation, and holdup on horizontal and vertical surfaces. If any are not accounted for, explain why.

Response to 3.g.10: As described in the Response to 3.g.8, the following volumes are treated within the VEGP containment water volume calculation as hold-up volumes that remove water from the containment pool: CS discharge piping (initially empty spray piping), water in transit from both the containment spray nozzles and the break itself, and the water droplets on containment walls. ES-99

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary)

11. Provide assumptions (and their bases) as to what equipment will displace water resulting in higher pool level.

Response to 3.g.11: The volumes occupied by structures, equipment, and equipment supports, etc. will displace water and result in a higher pool level. Examples of such equipment and or structures include concrete walls, accumulator tanks, piping, and cable trays. These volumes were accounted for in the VEGP containment water volume calculation. The 3D CAD model of containment was used to determine the correlation between the containment pool volume and water level. Smaller equipment, cables, and instruments are excluded from the CAD model and therefore provide some conservatism in the resulting water levels.

12. Provide assumptions (and their bases) as to what water sources provide pool volume and how much volume is from each source.

Response to 3.g.12: The following design inputs provided the basis for water sources and their volumes to determine the minimum containment water level for VEGP:

• The VEGP TS minimum initial RWST level was used for the initial RWST water level. As discussed in 3.g.1, when evaluating the minimum containment water level, the RWST level at the beginning of sump recirculation is either at the Low-Low level (minimum volume of water injected from the RWST at this level is 435,522 gal) or at the Empty level (the minimum volume of water injected from the RWST at this level is 580,497 gal).
• Four SI accumulators have a minimum volume of 6,555 gal/accumulator. The total minimum volume of the SI accumulators is therefore 26,220 gal. This volume is not credited in the SBLOCA cases because the RCS pressure is assumed to remain above their injection pressure as stated in Response to 3.g.2.
• The inventory of the RCS is assumed to remain relatively constant during normal operations. This is a reasonable assumption because during full power operation, the RCS remains at a fixed volume and remains at constant temperature and pressure. Due to the small volume of the RCS as compared to the RWST and its negligible variation in water volume, a best estimate value is representative. The best estimate RCS liquid volume is that associated with the total RCS liquid volume at hot full power conditions: 86,729 gal. The RCS represents both a source of water and a hold-up volume. The mass of water held up in the RCS may be more or less than the initial RCS mass depending on the elevation of the break (e.g., for a break at the top of the pressurizer, the vapor space of the pressurizer would be filled).

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Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary)

13. If credit is taken for containment accident pressure in determining available NPSH, provide description of the calculation of containment accident pressure used in determining the available NPSH.

Response to 3.g.13: Containment accident pressure was not credited in the VEGP analysis for pump NPSH. Using the VEGP NARWHAL model, pump NPSH margin was calculated at each time step using inputs from that time step. The containment pressure is assumed to be equal to the saturation pressure at the sump temperature for sump temperatures greater than 210.96 degrees F. Note that the temperature of 210.96 degrees F corresponds to the saturation temperature at the VEGP TS minimum containment pressure of -0.3 psig. For sump temperatures below 210.96 degrees F, the minimum containment pressure of-0.3 psig (or 14.396 psia) was used as the containment pressure to calculate the pump NPSHA.

14. Provide assumptions made which minimize the containment accident pressure and maximize the sump water temperature.

Response to 3.g.14: Containment Pressure As discussed in the Response to 3.g.13, the VEGP TS minimum containment pressure is -0.3 psig, which corresponds to a saturation temperature of 210.96 degrees F. When calculating pump NPSH margin, the containment pressure was minimized by using the minimum containment pressure of -0.3 psig for sump temperatures below 210.96 degrees F. When the sump temperature is higher than 210.96 degrees F, the containment pressure was assumed to be equal to the saturation pressure at the sump temperature, which is necessary to maintain the sump in a liquid phase. No accident pressure was credited for NPSH calculations. As stated in the NARWHAL CFP calculation, an accident pressure of 3.5 psi was used when evaluating flashing and degasification. However, a sensitivity case was run to show that it is unnecessary to credit accident pressure for degasification analysis. The sensitivity case without crediting any accident pressure resulted in

• degasification failures for 12 large hot leg breaks (greater than 28.5 inches). These 12 breaks already failed in the base case due to exceeding the debris limits failure criterion. As a result, the additional degasification failures without crediting any accident pressure will not increase the overall risk.

Sump Temperature As discussed in the Response to 3.g.1, the VEGP NARWHAL model used the design-basis sump temperature profile calculated for a double-ended pump suction LOCA with minimum safeguards. Note that the minimum safeguards temperature n ES-101

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) profile is conservatively higher than the temperature profile for the maximum safeguards case during recirculation through the sump strainers. The time-dependent sump temperature profile applied to all breaks is included as Figure 3.g.1-1.

15. Specify whether the containment accident pressure is set at the vapor pressure corresponding to the sump liquid temperature.

Response to 3.g.15: See the Responses to 3.g.13 and 3.g.14.

16. Provide the NPSH margin results for pumps taking suction from the sump in recirculation mode.

Response to 3.g.16: The RHR and CS pump NPSH margins were evaluated using the VEGP NARWHAL model. Table 3.g.16-1 provides a summary of the minimum NPSH margins for the RHR pumps in recirculation mode at various sump temperatures between 120 degrees F and 212 degrees F. The RHR pump NPSH margins shown in Table 3.g.16-1 are based on one of the four breaks within Table 3.b.4-2, all of which have the same NPSH margins due to the rule-based approach used in calculating head loss. ES-102

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Table 3.g.16-1 Limiting NPSH Margin vs. Sump Temperature NPSH Margin Net NPSH Margin Pool Strainer Before Subtracting After Subtracting Temperature Head Loss (oF) Strainer Head Strainer Head (ft-H20) Loss (ft-H20) Loss (ft-H20) 212 22.681 5.515 8 17.2 205 27.215 5.544 8 21.7 195 33.572 5.589 8 28.0 165 45.300 5.7298 39.6 153 46.599 9.006b 37.6 140 48.265 9.128b 39.1 133 46.810 13.615c 33.2 120 48.397 13.826c 34.6 a This includes clean strainer, conventional and chemical debris (calcium phosphate) head losses. b This includes clean strainer, conventional and chemical debris (calcium phosphate) head losses, and extrapolation constant. c This includes clean strainer, conventional and chemical debris (calcium phosphate and SAS) head losses and extrapolation constant. Although the minimum net NPSH margins shown in Table 3.g.16-1 are for RHR Pump A, they are bounding for all of the RHR and CS pumps. The RHR pumps are expected to have less NPSH margins than the CS pumps because of the higher head losses of the RHR strainers associated with the higher RHR pump flow rate and greater strainer debris loads. An NPSH evaluation was not performed for the SI pumps and CCPs because these pumps take suction from the RHR pumps during recirculation. Table 3.g.16-1 shows the NPSH margins before and after subtracting the total strainer head losses. The total strainer head losses include the clean strainer head loss, conventional debris (particulate and fiber) head loss, and chemical debris (calcium phosphate and SAS) head loss, as appropriate. Bounding head loss values, as shown in the Response to 3.f.10, were used in the evaluation. The head losses were also extrapolated to the end of the 30-day mission time as described in the Response to 3.f.10. As shown in the table, the minimum net NPSH margin for any given sump temperatures is over 17 ft. Therefore, adequate NPSH margin is available for Unit 1 and Unit 2 RHR and CS pumps to ensure their design functions. E5-103

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary)

h. Coatings Evaluation The objective of the coatings evaluation section is to determine the plant-specific ZOI and debris characteristics for coatings for use in determining the eventual contribution of coatings to overall head loss at the sump screen.

1., Provide a summary of type(s) of coating systems used in containment, e.g., Carboline CZ 11 Inorganic Zinc primer, Ameren 90 epoxy finish coat. Response to 3.h.1: The types of coating and systems used in containment are presented in Table 3.h.1-1. Qualified Coatings Table 3.h.1 Coatings Systems Used in Analyses OFT Density Substrate Layer Type (mil) (lbm/ft3) 1st Coat Carbozinc 11 5 208 2nd Coat Ameren 90 6 99.6 Steel Surfaces 3rd Coat Ameren 90 6 99.6 Total 17 1st Coat K&L 4129 1.5 69.0 2nd Coat K&L 4000 25 107.2 Concrete Surfaces 3rd Coat K&L D-Series 9 98.0 Total 35.5 Unqualified Coatings Unqualified coatings could include coatings within containment that do not have a specified preparation, application, or inspection compliant with plant specifications; previously qualified coatings that have noticeably deteriorated; coatings inaccessible for inspection; and coatings applied by vendors on vendor-supplied items that cannot be qualified. There are several types of unqualified coatings applied over numerous substrates within containment, including various types of epoxy, inorganic zinc, and alkyds. ES-104

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary)

2. Describe and provide bases for assumptions made in post-LOCA paint debris transport analysis.

Response to 3.h.2: The following assumptions related to coatings were made in the NARWHAL model:

• It was assumed that 100 percent of unqualified coatings were in the containment pool at the start of recirculation. This is a conservative assumption since no credit is taken for retention of unqualified coatings in upper containment regardless of the failure time or if containment sprays are initiated.
• It was assumed that the unqualified and degraded qualified coatings in VEGP have a recirculation transport fraction of 100%. This is consistent with the debris transport calculation, and is conservative since settling of this debris is not credited.

3. Discuss suction strainer head loss testing performed as it relates to both qualified and unqualified coatings. Identify surrogate material and what surrogate material was used to simulate coatings debris.

Response to 3.h.3: Silicon carbide was used to simulate both qualified and unqualified coatings debris. See the Response to 3.f.4 for detailed information on coating surrogates and the amount added to the test.

4. Provide bases for the choice of surrogates.

Response to 3.h.4: See the Response to 3.f.4. ES-105

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary)

5. Describe and provide bases for coatings debris generation assumptions. For example, describe how the quantity of paint debris was determined based on ZOI size for qualified and unqualified coatings.

Response to 3.h.5: The following assumptions related to coatings were made in the debris generation calculation:

• Qualified coatings within the ZOls were assumed to fail as 1Oµm diameter spheres; qualified coatings outside the ZOls were assumed to remain intact.

This is based on the guidance of NEI 04-07.

• It was assumed that the FN-8 qualified coatings system was applied to steel structures, including columns, equipment supports and grating. The FN-14/19 system was applied to all concrete surfaces within containment. Using these two systems is conservative because they have the largest number of coats and the largest final dry film thickness of all field coating systems present within containment for their respective substrates. Both field coating systems, including the type, dry-film thickness, and density are presented in Table 3.h.1-1.

The masses of unqualified coatings in containment are quantified based on detailed logs maintained over the life of the plant. The entire quantity of unqualified coatings, as shown in Tables 3.h.5-1 and 3.h.5-2, are assumed to fail for all breaks. The amount of coating debris generated at VEGP is shown Tables 3.h.5-1 and 3.h.5-2. Table 3.h.5-1: Coatings Debris for the Four Overall Worst-Case Breaks Break Location 11201-004-6-RB 11201-001-5-RB 11201-001-3-RB 11201-004-4-RB Break Size 29" 29" 29" 29" Break Type DEGB DEGB DEGB DEGB Qualified Epoxy 50.4 50.2 50.2 47.5 (lbm) Qualified IOZ 43.8 43.7 43.6 41.3 (lbm) Unqualified 2728.7 2728.7 2728.7 2728.7 Epoxy (lbm) Unqualified 58.9 58.9 58.9 58.9 Alkyd (lbm) Unqualified IOZ 55.7 55.7 55.7 55.7 (lbm) ES-106

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Table 3.h.5-2: Coatings Debris for the Four Worst-Case Breaks that Do Not Fail the Strainer Acceptance Criteria Break Location 11201-004-4-RB 11201-001-3-RB 11201-003-5-RB 11201-002-5-RB Break Size 20" 23" 25" 17" Break Type Partial Partial Partial Partial Qualified Epoxy 6.7 8.8 15.6 6.3 (lbm) Qualified IOZ 5.8 7.7 13.6 5.5 (lbm) Unqualified 2728.7 2728.7 ' 2728.7 2728.7 Epoxy (lbm) Unqualified 58.9 58.9 58.9 58.9 Alkyd (lbm) Unqualified IOZ 55.7 55.7 55.7 55.7 (lbm)

6. Describe what debris characteristics were assumed, i.e., chips, particulate, size, distribution, and provide bases for the assumptions.

Response to 3.h.6: In accordance with the guidance provided in NEI 04-07 (Reference 2) and the associated NRC SE (Reference 3), all coating debris was treated as particulate and therefore transported entirely to the sump strainer. See the Response to 3.h.1, 3.h.2, and 3.h.5 for additional description of debris characteristics.

7. Describe any ongoing containment coating conditions assessment program.

Response to 3.h.7: SNC conducts condition assessments of coatings inside containment every outage under the site work control system. As localized areas of degraded coatings are identified, those areas are evaluated and scheduled for repair or replacement as necessary. The periodic condition assessments and resulting repair and replacement activities assure that the amount of coatings that may be susceptible to detachment from the substrate during a LOCA event is minimized. ES-107

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary)

i. Debris Source Term The objective of the debris source term section is to identify any significant design and operational measures taken to control or reduce the plant debris source term to prevent potential adverse effects on the ECCS and CSS recirculation functions.

Provide the information requested in GL 2004-02 Requested Information Item 2(f) regarding programmatic controls taken to limit debris sources in containment. GL 2004-02 Requested Information Item 2(f) A description of the existing or planned programmatic controls that will ensure that potential sources of debris introduced into containment (e.g., insulations, signs, coatings, and foreign materials) will be assessed for potential adverse effects on the ECCS and CSS recirculation functions. Addressees may reference their responses to GL 98-04, "Potential for Degradation of the Emergency Core Cooling System and the Containment Spray System after a Loss-of-Coolant Accident Because of Construction and Protective Coating Deficiencies and Foreign Material in Containment," to the extent that their responses address these specific foreign material control issues. In responding to GL2004-02 Requested Information Item 2(f), provide the following:

1. A summary of the containment housekeeping programmatic controls in place to control or reduce the latent debris burden. Specifically for RMI/low-fiber plants, provide a description of programmatic controls to maintain the latent debris fiber source term into the future to ensure assumptions and conclusions regarding inability to form a thin bed of fibrous debris remain valid.

Response to 3.i.1: SNC procedure, "Containment Exit Inspection," provides detailed guidance for containment inspection to ensure no loose debris (e.g., rags, trash, clothing, etc.) is present in the containment that could be transported to the containment sump and cause restriction of pump suctions during LOCA conditions. This procedure contains an extensive checklist detailing all areas of containment that must be inspected for cleanliness prior to plant startup after each outage. SNC procedure, "Containment Entry," establishes guidance to inventory and control items carried into containment during non-outage entries. This procedure ensures that no loose debris (e.g., rags, trash, clothing, etc.) is present in the containment, which could be transported to the containment sump and cause restriction of pump suctions during LOCA conditions. E5-108

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary)

2. A summary of the foreign material exclusion programmatic controls in place to control the introduction of foreign material into the containment.

Response to 3.i.2: SNC procedure, "Foreign Material Exclusion Program," establishes the administrative controls and personnel responsibilities for the foreign material exclusion (FME) program. The procedure describes methods for controlling and accounting for material, tools, parts, and other foreign material to preclude their uncontrolled introduction into an open or breached system during work activities. This procedure also provides guidance for establishing and maintaini,ng system cleanliness, recovering from an intrusion of foreign material, and re-establishing system cleanliness requirements.

3. A description of how permanent plant changes inside containment are programmatically controlled so as to not change the analytical assumptions and numerical inputs of the licensee analyses supporting the conclusion that the reactor plant remains in compliance with 10 CFR 50.46 and related regulatory requirements.

Response to 3.i.3: An enhancement to the screening guidelines and considerations for the design input process, which is part of the design change procedure, has introduced a requirement to review the impact of a proposed change on the documentation that forms the design basis for the response to GL 2004-02. The specific areas that are addressed are:

• Insulation inside containment
• Fire barrier material inside containment
• Coatings inside containment
• Inactive volumes in containment
• Labels inside containment
• Buffer changes (iodine and pH control)
• Structural changes (i.e., choke points) in containment
• Downstream effects (piping components downstream of the ECCS sump strainers)

Inclusion in the design input process ensures all design changes consider these attributes during the design process. ES-109

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary)

4. A description of how maintenance activities including associated temporary changes are assessed and managed in accordance with the Maintenance Rule, 10 CFR 50.65.

Response to 3.i.4: Maintenance activities, including temporary changes, are subject to the provisions of 10 CFR 50.65(a)(4), as well as VEGP TSs. SNC fleet procedures also provide guidance. For instance, the 50.59 review process procedure provides details on maintenance activities and temporary modifications, while the on-line work management process procedure establishes administrative controls for performing on-line maintenance of structures, systems, components (SSCs) to enhance overall plant safety and reliability. Further guidance is also available in the temporary configuration change procedure.

5. If any of the following suggested design and operational refinements given in the guidance report (guidance report, Section 5) and SE (SE, Section 5.1) were used, summarize the application of the refinements.

a. Recent or planned insulation change-outs in the containment which will reduce the debris burden at the sump strainers.

Response to 3.i.5.a: All of the Min-K insulation located inside the steam generator compartments (original ZOI analyzed for GL 2004-02) was removed from VEGP Unit 1 and Unit 2 containments during refueling outage 1R13 (Fall 2006) and refueling outage 2R12 (Spring 2007). There are no known quantities of Min-K in VEGP Unit 1 and Unit 2 containments outside of the secondary shield wall (outer wall of the steam generator compartments). However, Min-K was only used as insulation in penetrations, which are difficult to inspect. This leaves the containment wall as the only ,. possible location of remaining Min-K. As shown in Figure 3.a.1-1, there is only one line outside of the steam generator compartments with analyzed breaks (i.e. welds inside the first isolation valve). If a worst case ZOI of 28.60 is assumed for this 2" line, the ZOI radius would be about 4.8 feet. Using the strainer as a reference dimension (square with each side approximately 5 feet), it is apparent that the ZOI could not reach the containment wall. Therefore, Min-K is not considered when analyzing the sump strainers. ES-110

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary)

b. Any actions taken to modify existing insulation (e.g., jacketing .or banding) to reduce the debris burden at the sump strainer.

Response to 3.i.5.b: This suggested design and operational refinement was not used in the VEGP evaluation.

c. Modifications to equipment or systems conducted to reduce the debris burden at the sump strainers.

Response to 3.i.5.c: This suggested design and operational refinement was not used in the VEGP evaluation.

d. Actions taken to modify or improve the containment coatings program.

Response to 3.i.5.d: No specific actions were taken to modify or improve the containment coatings program; however, enhancements were made to the screening guidelines and considerations for the design input process to ensure that all design changes consider GL 2004-02 attributes during the design process. The specific areas that are addressed are listed in the Response to 3.i.3.

j. Screen Modification Package The objective of the screen modification package section is to provide a basic description of the sump screen modification.

1. Provide a description of the major features of the sump screen design modification.

Response to 3.j.1: The currently installed strainers for RHR and CS consist of four parallel, vertically stacked, modular disk strainer assemblies that are connected to a plenum installed over each sump. Each RHR strainer assembly consists of 18 stacked disks that are 30 inches long by 30 inches wide, and the height of the disk portion of the strainer is

53. 75 inches. The RHR strainer assemblies are 59.6 inches tall, measured from the containment floor. The four RHR strainer assemblies provide approximately 765 ft2 of perforated plate surface area and 179 ft2 of circumscribed surface area per sump.

Each CS strainer assembly consists of 14 stacked disks that are 30 inches long by 30 inches wide, and the height of the disk portion of the strainer is 41. 75 inches. The CS strainer assemblies are 47.6 inches tall measured from the containment floor. Figure 3.j.1-1 below shows a picture of one CS strainer. Each of the four CS ES-111

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) strainer assemblies provides approximately 590 ft2 of perforated plate surface area and 139 ft2 of circumscribed surface area. Subsequent risk-informed analysis has led to the proposed modification of the Unit 1 and Unit 2 RHR sump strainer assemblies. The RHR strainers will be modified to reduce the overall height approximately 6 inches by removing the two top disks per disk stack. The modified RHR strainer assembly will consist of 16 stacked disks, and the disk portion of the strainer is approximately 47.75 inches high (53.75 in. - 6 in. =47.75 in.). As shown in the response to 3.f.2, the overall height of the modified RHR strainer is 53.25 inches, measured from the containment floor to the highest strainer disks. The four RHR modified strainer assemblies provide approximately 677 .6 ft2 of perforated plate surface area and 159 ft2 of circumscribed surface area per sump as calculated below. 4modules (4 sides ) (30 in)(47.75 in)(l ftz) (circumsc:ibed area)) module side = 1 59 ft2 ( (144 in 2 )

• All of the analyses shown in this submittal were performed for the modified strainer configuration. Operating procedures are being revised, in addition to the planned physical modification, to ensure that the RHR strainers are completely submerged for an increased number of postulated LOCA scenarios.

ES-112

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Figure 3.j.1-1 Containment Spray Strainer

2. Provide a list of any modifications , such as reroute of piping and other components ,

relocation of supports , addition of whip restra ints and missile shields, etc., necessitated by the sump strainer modifications. Response to 3.j.2: The following modifications were necessitated by those of the sump strainer:

• Installation of new and replacement of existing ECCS flow orifices to allow new ECCS throttle valve settings.
• Cage assembly vortex suppressors installed in the sumps removed .
• Temperature elements for the Units 1 and 2 RHR sumps replaced and relocated.
• Two conduit interferences at the Unit 2 RHR sump Train A screen rerouted through an area outside of the sump screen envelope.
• Three electrical interferences for the new Unit 2 CS sump Train A screen relocated/rerouted through an area outside of the sump screen envelope .
• The RHR strainers will be reduced in height by the removal of two disks from each stack to ensure full submergence for an increased number of postulated break scenarios as described in the Response to 3.j .1.

ES-113

Enclosure 5 Suppleme"ntal Response to NRC Generic Letter 2004-02 (Non-Proprietary)

k. Sump Structural Analysis The objective of the sump structural analysis section is to verify the structural adequacy of the sump strainer including seismic loads and loads due to differential pressure, missiles, and jet forces.

Provide the information requested in GL2004-02 Requested Information Item 2(d)(vii). GL 2004-02 Requested Information Item 2(d)(vii) Verification that the strength of the trash racks is adequate to protect the debris screens from missiles and other large debris. The submittal should also provide verification that the trash racks and sump screens are capable of withstanding the loads imposed by expanding jets, missiles, the accumulation of debris, and pressure differentials caused by post-LOCA blockage under flow conditions.

1. Summarize the design inputs, design codes, loads, and load combinations utilized for the sump strainer structural analysis.

Response to 3.k.1: Design Codes (1) American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section Ill, Subsection NC and ND, 1989 Edition. (2) American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section Ill, Appendix I, 1989 Edition, Table 1-6.0 for Modulus of Elasticity, Table 1-5.0 for thermal expansion, and Table 1-7.2 for allowable stress (S). ES-114

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Material Properties The Material properties come from the ASME Code and are tabulated in Table 3.k.1-1 below. Table 3.k.1-1 Material Properties Material I Property @Room @ Maximum Water Temperature Temperature (70°F) (250°F) SA-240 Type SS304 (Strainer): E, Elastic modulus, psi 28.3x1 QB 27.45 x1QB Coefficient of thermal expansion, in/in/°F 8.6x1 Q-B 8.995 x1Q-B Poisson's ratio Q.3 Q.3 SA-479 Type SS410 (Tie Rod): E, Elastic modulus, psi 28.3x1 QB 27.45x1QB Coefficient of thermal expansion, in/in/°F 5.9 x1 Q-B 6.1 x1 Q-B Load Combinations Table 3.k.1-2 shows the load combinations specified for the VEGP passive suction strainer design. Table 3.k.1-2 Load Combinations for VEGP Strainer Design Load Combination Strainer Assembly Design W+ Po+ OBE1 LevelB WD + Pd + OBE2 + TEmax + Per LevelD WD + Pd +SSE2 + Per Support Structure Design W +Po+ OBE1 Level B WD + Pd + OBE2 + TEmax LevelD WD +Pd +SSE2 Nomenclature: W = Weight (Dry strainer Assembly Weight) WD = Weight+ Debris Weight + Hydrodynamic Mass (LOCA Event with Strainer in Water) Per = Crush Pressure (During Suction Strainer Operation in Water Post LOCA) Pd = Design Pressure (LOCA Event) +Water Head (Strainer Open System) Po = Design Pressure (Strainer Open System) ES-115

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) OBE1 = Operating Basis Earthquake (Inertia Load in Air) OBE2 = Operating Basis Earthquake (Inertia Load with Strainer in Water - Include Debris Weight+ Hydrodynamic Mass) TE max = Thermal Expansion (Accident Condition) SSE1 = Safe Shutdown Earthquake (Inertia Load with Strainer in Air) SSE2 = Safe Shutdown Earthquake (Inertia Load with Strainer in Water - Include Debris Weight + Hydrodynamic Mass) The seismic loads are based on the lateral and vertical accelerations of the response spectrum according to the first mode of frequency of the strainer assembly in water. The natural frequency checks in the original analysis show that the system is in the rigid range. The design pressure, Po or Pd, has no impact on the system because the strainer is an open system. The hydrodynamic mass and debris weight are distributed evenly and are added to the strainer finite element model by adjusting the density of the material. A combined load table for the strainer component evaluation is summarized in Table 3.k.1-3. For the, design load case, the dry strainer weight (or 1G) is combined with the QBE vertical acceleration for a combined loading of 1.375G vertically. In addition, QBE horizontal acceleration of 0.27G is applied in both X and Y lateral directions. For the Level B load case, the strainer weight in water including debris and hydrodynamic mass (or 1G) is combined with the QBE vertical acceleration for a combined loading of 1.375G vertically. In addition, QBE horizontal acceleration of 0.27G is applied in both X and Y lateral directions as well as crush pressure and thermal loading. For the Level D load case, the strainer weight in water including debris and hydrodynamic mass (or 1G) is combined with SSE vertical acceleration for a combined loading of 1.6G vertically. In addition, SSE horizontal acceleration of 0.4125G is applied in both X and Y lateral directions as well as crush pressure. Table 3.k.1-3 Load Table for the VEGP Strainer Design Strainer Load Combination Inertia Z* Inertia X Inertia Y Per Temp*** Assembly (G) (G) (G) (psi) (oF) Design W+ Po+ OBE1 1.375 0.27 0.27 LevelB WO + Pd + OBE2 + TEmax + Per 1.375 0.27 0.27 4.46** 180 LevelD WO+ Pd +SSE2 +Per 1.6 0.4125 0.4125 4.46**

• Axis orientation: Z Vertical, X and Y Lateral

   ** Equivalent to 10.1 ft of head loss
   *** Stress free temperature is assumed to be 70 °F, t::..T =(250- 70) °F =180 °F E5-116 I

~

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Modal Analysis Modal analyses were performed using the suction strainer finite element models. Modal results were obtained for the dry strainer and for the wet strainer with added debris weight and hydrodynamic mass during LOCA and post-LOCA events. The strainer structural mass and natural frequencies are calculated for the first four modes and are summarized in Table 3.k.1 ~4. Table 3.k.1-4 Replacement Strainer Weight and Frequency Typical RHR Strainer in Air W= 7,150 lbm Mode 1 37.311 Hz Mode 2 37.722 Hz Mode 3 39.240 Hz Mode4 85.327 Hz Typical RHR Strainer in Water WO = 10,655 lbm Mode 1 30.566 Hz Mode2 30.902 Hz Mode 3 32.146 Hz Mode4 69.901 Hz RHR Train B Strainer in Water WO= 11,256 lbm Mode 1 31.421 Hz Mode 2 32.037 Hz Mode 3 33.426 Hz Mode4 68.966 Hz Load Application Loads used in the stress analysis of the strainer models include the weight of the strainer assembly, hydrodynamic mass and debris mass, the crush pressure due to suction strainer operation, and the lateral and vertical inertial accelerations of. Response Spectrum (OBE & SSE) corresponding to the first mode frequency of strainer assembly in water. The crush pressure is applied on the top and bottom surfaces of the disk sets accounting for debris blockage. The weight of the strainer assembly model in water (WD) is the sum of the weight of the strainer assembly in air (W), the debris weight, and the hydrodynamic mass. The debris and hydrodynamic mass are uniformly distributed over the strainer assembly and support for mode shape and stress analysis. The crush pressure is applied on the plenum for Level D load case. The ASME code combination stress limits are summarized in Tables 3.k.1-5 and 3.k.1-6. ES-117

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Table 3.k.1-5 Stress Limits for Strainer Components (250 degrees F) Service Level Stress Category Stress Limit (ksi) Design Pm s 17.15 Pm+ Pb 1.5 s 25.725 Service Level B Pm 1.1 s 18.865 Pm+ Pb 1.65 s 28.3 Pm* s 16.35 Pm+ Pb+ Q* 3 Sm 69.9 Service Level D Pm 2.0 s 34.3 Pm+ Pb 2.4 s 41.16 S: 17,150 psi for SS304 Sm: 23,300 psi for SS410 Table 3.k.1-6 Weld Stress Limits (250 degrees F) Type Service Level Stress Stress Limit (ksi) Category Fillet ND-3929 & ND-5260* Shear 0.85x0.7xS 10,200 Plug ND-3929 & ND-5260* Shear 0.65x0.8xS I 8,918 S: 17, 150 psi for SS304

• No specific weld inspection requirements. VT-visual test inspection will be performed.

ES-118

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary)

2. Summarize the structural qualification results and design margins for the various components of the sump strainer structural assembly.

Response to 3.k.2: Table 3.k.2-1 Stress Ratio Summary for Strainer Components Based on ASME Subsection NC Component Service Level Stress Ratio* Perforated Plates Design - RHR model 18.55 Fingers Design - RHR model 21.10 Finger Frames Design - RHR model 40.70 Perforated Spacers Design - RHR model 17.23 Center Post Design - RHR model 39.27 Connecting Plates Design - RHR model 41.69 Support Base Design - RHR model 16.40 Base Frame Design - RHR model 16.84 I-Beams Design - RHR model 16.40 Perforated Plates Level B - RHR model 2.30 Fingers Level B- RHR model 3.65 Finger Frames Level B- RHR model 13.78 Perforated Spacers Level B - RH R model 10.49 Center Post Level B- RHR model 26.13 Connecting Plates Level B - RH R model 14.63 Support Base Level B- RHR model 9.58 Base Frame Level B - RHR model 9.58 I-Beams Level B - RHR model 18.04 Tie Rods Level B - RHR model 2.38 Perforated Plates Level D - RHR model 3.33 Fingers Level D - RH R model 5.30 Finger Frames Level D - RHR model 19.83 Perforated Spacers Level D - RHR model 12.44 Center Post Level D - RHR model 30.11 Connecting Plates Level D - RHR model 21.09 Support Base Level D - RHR model 14.96 Base Frame Level D - RHR model 14.96 I-Beams Level D - RHR model 14.36 Perforated Plates Level D - RHR Train B model 3.35 Fingers & Frames Level D - RHR Train B model 5.21 Perforated Spacers Level D- RHR Train B model 10.67 Center Post Level D- RHR Train B model 26.78 Connecting Plates Level D- RHR Train B model 20.47 Support Base Level D- RHR Train B model 7.16 Base Frame Level D - RHR Train B model 9.97 I-Beams Level D- RHR Train B model 9.40

• Stress Ratio= ASME Code Stress Limit I Calculated Max Stress ES-119

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Table 3.k.2-2 Stress Summary for Welds based on Service Level D Load Weld Location Weld Stress Allowable Stress (psi) (psi) Stress** Ratio* Perforated Plate to Finger 3,708 8,918 2.4 Perforated Plate to Finger 9,016 10,200 1.13

• Stress Ratio= ASME Code Stress Limit I Calculated Max Stress

          ** Conservative Level A Stress Limits, ASME Code Section Ill, Subsection ND-3923 at 250 °F The ASME Code combination stress limits are summarized in Tables 3.k.1-5 and 3.k.1-6.

Table 3.k.2-3 Typical RHR Strainer Stress Ratios for Service Level D Component Stress Max Stress Intensity Stress Limit Stress Ratio* Category (psi) (psi) Perforated Plates Pm less than 12,347 34,300 2.78 minimum Pm+ Pb 12,347 41,160 3.33 Fingers Pm less than 7,764 34,300 4.42 minimum Pm+ Pb 7,764 41, 160 5.30 Finger Frames Pm less than 2,076 34,300 16.52 minimum Pm+ Pb 2,076 41, 160 19.83 Perforated Spacers Pm less than 3,308 34,300 10.37 minimum Pm+ Pb 3,308 41, 160 12.44 Center Post Pm less than 1,367 34,300 25.09 minimum Pm+ Pb 1,367 41, 160 30.11 Connecting Plates Pm less than 1,952 34,300 17.57 minimum Pm+ Pb 1,952 41, 160 21.09 Support Base Pm less than 6,855 34,300 5.00 minimum Pm+ Pb 6,855 41,160 6.00 Base Frame Pm less than 6,855 34,300 5.00 minimum Pm+ Pb 6,855 41, 160 6.00 I-Beams Pm less than 5,684 34,300 6.03 minimum Pm+ Pb 5,684 41, 160 7.24

• Stress Ratio= ASME Code Stress Limit I Calculated Max Stress E5-120

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Table 3.k.2-4 RHR Train B Strainer Stress Ratios for Service Level D Component Stress Max Stress Intensity Stress Limit Stress Ratio* Category (psi) (psi) Perforated Plates Pm less than 12,289 34,300 2.79 minimum Pm+ Pb 12,289 41, 160 3.35 Fingers & Frames Pm less than 7,894 34,300 4.35 minimum Pm+ Pb 7,894 41, 160 5.21 Perforated Spacers Pm less than 3,856 34,300 8.90 minimum Pm+ Pb 3,856 41, 160 10.67 Center Post Pm less than 1,537 34,300 22.32 minimum Pm+ Pb 1,537 41,160 26.78 Connecting Plates Pm less than 2,011 34,300 17.06 minimum Pm+ Pb 2,011 41,160 20.47 Support Base Pm less than 5,750 34,300 5.97 minimum Pm+ Pb 5,750 41, 160 7.16 Base Frame Pm less than 4,130 34,300 8.31 minimum Pm+ Pb 4,130 41, 160 9.97 I-Beams Pm less than 4,380 34,300 7.83 minimum Pm+ Pb 4,380 41, 160 9.40

• Stress Ratio= ASME Code Stress Limit I Calculated Max Stress Weld Analysis Since the finite element model with the typical RHR strainer configuration has slightly higher overall stress results, the ANSYS analysis results in this load case were used to calculate the load transfer through the welds. For a given weld location, the elements and corresponding nodes at the weld were selected on one side of the node, and the AN SYS post-processor was used to calculate the forces transferred across the weld section. These forces were then used to calculate the stresses based on the weld section properties. If the welds consisted of more than one weld, then the group section properties were used.

Weld stresses were calculated for simultaneous application of loads for Service Level D. These calculated stress values were compared with the ASME Code shear stress limits. The minimum weld stress ratios for all the weld locations are summarized in Tables 3.k.2-1 and 3.k.2-2. For the welds between the fingers and perforated plate, robotic welding will be utilized to ensure a weld diameter of 3/16 inches. At the worst stress intensity finger location, a net shear Fx of 115.8 pounds force (lbf), a net shear Fy of 184 lbf, and a net tensile Fz of 184 lbf are obtained between two sides of the finger as seen in Figure 3.k.2-1. ES-121

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) AFz= 184 lbf . z AFy= 184 lbf M'x= 115 .8 lbf Figure 3.k.2-1 Worst Stress Intensity Finger Location Free Body Diagram Considering a line of welds, net forces are reacted by the circular areas of the plug weld, Aw. The unbalanced Fz causes a moment of 30 inch-pounds force (in-lbf) and is reacted by the section modulus of the weld, Sw, when the weld is treated as a line. The unbalanced Fx and Fy cause torsion and are reacted by the twisting property of the weld, Jw. Jw is large because the line of weld is approximately 8 inches long. The stresses caused by torsion are therefore negligible. Weld load treated as a line: 2 2 2 _ ( M Fz ) ( Fx) + (AFwy) f- Sw +Aw + Aw f Sa= Nt where Sa= 8,918 psi and t = 0.078 inches Five welds along each finger will satisfy the stress allowable of 8,918 psi. The welds are to be distributed evenly along the finger. Similarly, at the weld location between the finger frame and the perforated plate, a net shear Fx of 180 lbf., a net shear Fy of 325 lbf, and a net tensile Fz of 1,437 lbf are obtained between two sides of the frames. The moment from the unbalanced Fz is 566 in-lbf and is reacted by Sw of 1,200 in 2 for the square frame shape weld line. The fillet weld area, Aw, is 0.707 x 2t, where t equals 0.078 inches. In addition, an intermittent weld has a knock down factor of 0.66 for weld length of 3 inches and pitch distance of 5 inches. The weld calculation shows that 18 inches of weld length is recommended along each edge of the disk. The fillet welds should cover corners and at finger protrusion areas. Based on a width of 0.070 inches for the weld and stress allowable of 10,200 psi, the recommended intermittent welds should be 3 inches with 5-inch pitch. ES-122

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Interface Load The original Vogtle strainers were re-evaluated when the original 16 bolt anchor configuration was revised to a 32 bolt anchor configuration. The updated anchor bolt loads from the simplified GEH finite element model are documented in GEH letter no. JXDR7-2006-02, Rev. 1. The worst case anchor loads were for Level D load cases (Wd + SSE2). The largest tension load (z-direction) that an individual anchor bolt sees is 195 lbf. The largest lateral X-direction force is 237 lbf, and the largest Y-direction force is 279 lbf. The worst overall anchor bolt interaction ratio for each load case is provided below in Table 3.k.2-5. Table 3.k.2-5 Worst Case Interaction Ratios (l.R.) for Anchor Bolts Fx (lbf) Fy (lbf) Fz (lbf) l.R.* WD + SSE2 152.4 167.59 -195.4 0.423 (downward) WD + SSE2 137.91 278.26 -38.734 0.350 (upward)

• The Interaction Ratio (l.R.) is the inverse of a stress ratios in the tables above and equals the resultant force per anchor divided by the allowable force per anchor.

16-Disk ECCS Suction Strainer Summary (( ES-123

Enclosure 5 Supplemental Response to 'NRC Generic Letter 2004-02 (Non-Proprietary) Table 3.k.2-6: Service Level D Stress Summary for 16-Disk Strainer II

3. Summarize the evaluations performed for dynamic effects such as pipe whip, jet impingement, and missile impacts associated with high-energy line breaks (as applicable).

Response to 3.k.3: As shown in Figure 3.a.1-1, there is only one line outside of the steam generator compartments with analyzed breaks (i.e. welds inside the first isolation valve). However, this 2" line is at an elevation of 208 ft, which is 32 ft above the strainer at approximately 176 ft. Thus, pipe whip from this line impacting the strainer is not considered a credible scenario. The strainers are seismically qualified, robust structures designed with a crush pressure of approximately 10.4 psi, which is approximately the impingement pressure at 11.5 pipe diameters from a break (Reference 3). Considering the distance from the analyzed break, jet impingement loads are not a credible concern. Finally, the line in question is on the side of the pressurizer cubicle wall, where no unsecured items would be located. Therefore, missiles generated by this analyzed break are not a credible concern. There are no other high-energy lines in the area of the emergency sumps except for the RHR and HHSI lines that are used for accident mitigation and are not assumed to be the accident initiator. ES-124

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) The 12-inch RHR hot leg recirculation line is located more than 6 ft above the Train B CS strainer outside of the secondary shield wall. The RHR suction header has two isolation valves in series to isolate it from the RCS hot leg recirculation line. The valves are normally closed except when the RHR is operating. The inboard isolation valve is on the other side of the secondary shield wall inside the steam generator compartment preventing high-energy RCS discharge outside of the secondary shield wall. Therefore, there is no possibility of pipe whip impacts or jet loads associated with this pipe. In addition, this line is only pressurized during shutdown, refueling, and accident mitigation. During normal operation, this line is not pressurized. Thus, this line is not considered for evaluation as a postulated location for a high-energy line break accident. The 6-inch HHSI line is located approximately 69 inches above the Train B RHR strainer outside of the secondary shield wall. There is a check valve on the other side of the secondary shield wall inside the steam generator compartment preventing high-energy RCS discharge outside of the secondary shield wall. Therefore, there is no possibility of pipe whip impacts or jet loads associated with this pipe. In addition, this line is only pressurized during accident mitigation. During normal operation, this line is not pressurized. Thus, this line is not considered for evaluation as a postulated location for a high-energy line break accident. The strainers are located outside of the steam generator compartments and inside the outer containment wall. Therefore, the strainers are adequately protected from the hazardous effects of missiles.

4. If a backflushing strategy is credited, provide a summary statement regarding the sump strainer structural analysis considering reverse flow.

Response to 3.k.4: Backflushing of the sump strainers, or any other active approach, is not credited in the VEGP analysis. The RHR strainer suction lines have check valves to prevent reverse flow through the strainers. If containment pressure is high enough to actuate containment spray, the pressure is high enough to prevent reverse flow through the CS strainers. Therefore, no structural analysis considering reverse flow is required. ES-125

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non~Proprietary) I. Upstream Effects The objective of the upstream effects assessment is to evaluate the flowpaths upstream of the containment sump for holdup of inventory, which could reduce flow to and possibly starve the sump. Provide a summary of the upstream effects evaluation including the information requested in GL 2004-02 Requested Information Item 2(d)(iv). GL 2004-02 Requested Information Item 2(d)Ov) The basis for concluding that the water inventory required to ensure adequate EGGS or GSS recirculation would not be held up or diverted by debris blockage at choke points in containment recirculation sump return flowpaths.

1. Summarize the evaluation of the flowpaths from the postulated break locations and containment spray washdown to identify potential choke points in the flow field upstream of the sump.

Response to 3.1.1: The following areas I items were considered as part of the evaluation to determine potential choke points for flow upstream of the sump: Refueling Cavity Evaluations of containment, along with review of the CFO model, indicated no significant areas would become blocked with debris and hold up water during the sump recirculation phase. The area of the refueling cavity, which is the area around the reactor head that is flooded prior to fuel movement, is the only significant area in containment that can retain water during an event that requires containment spray. However, this area is drained by a large clear flow path that cannot be easily blocked with debris. See the Response to 3.1.4 for additional information. Inside Secondary Shield Wall A postulated LOCA inside the secondary shield wall in the lower elevations of the containment was considered limiting with respect to flow restrictions upstream of the sump. The flow path from this break area to the sump strainers is primarily through two labyrinth-like walkways through the shield wall. There are also smaller openings through the shield wall for piping, but these are much smaller than the walkways. The walkways provide a large clear flow path from inside the shield wall to the screen area. In addition, any restriction of the smaller through-wall piping openings would have minimal effect on the overall flow path to the strainers, since water would simply flow through the open walkways. ES-126

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Containment Spray Washdown Containment spray washdown has a clear path to the containment sump area. Large sections of the floor on each level in containment are covered with grating that allows the water to pass. A complete evaluation of containment, along with a review of the CFO model, indicated no significant areas would become blocked with debris and hold up water during the sump recirculation phase.

2. Summarize measures taken to mitigate potential choke points.

Response to 3.1.2: Per the Response to 3.1.1, no measures were necessary to mitigate potential choke points.

3. Summarize the evaluation of water holdup at installed curbs and/or debris interceptors.

Response to 3.1.3: There are no curbs or debris interceptors that provide water volume holdup in the VEGP containments.

4. Describe how potential blockage of reactor cavity and refueling cavity drains has been evaluated, including likelihood of blockage and amount of expected holdup.

Response to 3.1.4: The refueling cavity is drained by two 12-inch pipes. During refueling, these drains are secured by installing flanges. These flanges are removed prior to entry into Mode 4 and above. The VEGP limiting break with respect to upstream flow blockage occurs under the operating deck and inside the secondary shield wall. This break would result in a torturous path for large debris to travel above the operating deck and land in the refueling cavity. Large debris would have to travel through the RCP access ports or the Steam Generator and Pressurizer cubicles. The RCS access ports are covered with grating, and there are significant structural elements that would prevent any large pieces of debris from entering upper containment. The same is true for the path through the Steam Generator and Pressurizer cubicles. Each cubicle contains several levels of grating and significant structural elements that would make large pieces of debris entering upper containment via these paths highly unlikely. Therefore, the clogging of the refueling cavity drains is minimized. ES-127

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 {Non-Proprietary) The drains into the area under the reactor (e.g., reactor cavity) could become blocked. For breaks outside of the reactor cavity, there is no detrimental impact of this blockage, as it would inhibit loss of water from the active EGGS sump to an inactive area beneath the vessel. The flooding analysis assumes this area floods during the event. For breaks inside the reactor cavity, there is also no detrimental impact of this blockage, since the majority of the flow from the break would travel to the EGGS sump through the hot leg and cold leg penetrations.

m. Downstream Effects - Components and Systems The objective of the downstream effects, components and systems section is to evaluate the effect of debris carried downstream of the containment sump screen on the function of the EGGS and GSS in terms of potential wear of components and blockage of flow streams.

Provide the information requested in GL 2004-02 Requested Information Item 2(d)(v) and 2(d)(vi) regarding blockage, plugging, and wear at restrictions and close tolerance locations in the EGGS and GSS downstream of the sump. GL 2004-02 Requested Information Item 2(d)(v) The basis for concluding that inadequate core or containment cooling would'not result due to debris blockage at flow restrictions in the ECCS and CSS flowpaths downstream of the sump screen (e.g., a HPSI throttle valve, pump bearings and seals, fuel assembly inlet debris screen, or containment spray nozzles). The discussion should consider the adequacy of the sump screen's mesh spacing and state the basis for concluding that adverse gaps or breaches are not present on the screen surface. GL 2004-02 Requested Information Item 2(d)(vi) , Verification that the close-tolerance subcomponents in pumps, valves and other ECCS and CSS components are not susceptible to plugging or excessive wear due to extended post-accident operation with debris-laden fluids.

1. If NRG-approved methods were used (e.g., WGAP-16406-P-A with accompanying NRG SE), briefly summarize the application of the methods. Indicate where the approved methods were not used or where exceptions were taken, and summarize the evaluation of those areas.

Response to 3.m.1: The following methodology was employed in the ex-vessel downstream effects* evaluations. The evaluations did not use any unapproved methods or take any exceptions to NRG-approved methods. ES-128

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Maximum Debris Ingestion Determination Debris blockage of flow restrictions in the ECCS and CSS flowpaths downstream of the sump screen was addressed within the downstream effects evaluations for ECCS valves and equipment. Each unit has two sets of screens - RHR and CS emergency sump screens. The adequacy of the sump screens' mesh spacing or strainer hole size [nominal hole diameter of 0.09375 inches (3/32 inches)] is conservatively addressed by assuming that the maximum amount of particulate (coatings and latent debris) transported to the strainers passes through the strainers. Additionally, the evaluation used a quantity of fiber debris that passes through the strainers (100 g/FA), which is greater than the maximum total reactor vessel fiber load amount shown for a hot-leg break in Table 3.n.1-6. The ex-vessel downstream effects evaluations were based on this maximum amount of ingested debris (see Initial Debris Concentrations below). The Unit 1 strainers were inspected after installation and found to conform to design specifications. No adverse gaps or breaches were found on the screen surface. The Unit 2 strainers were inspected upon installation, and deficiencies in the fabrication of the Unit 2 CS sump screens were discovered; specifically, there were 124 holes greater than the nominal specified sump screen hole-diameter of 0.09375 inches (3/32 inches). No holes greater than a 0.25-inch diameter were found in the Unit 2 strainers; therefore, ingestion of debris will not cause plugging of downstream CS components because the smallest component diameter is 0.375 in. Initial Debris Concentrations Initial debris concentrations were developed using the assumptions and methodology described in Chapter 5 of WCAP-16406-P-A. Additionally, for conservatism, the maximum amount of particulate (coatings and latent debris) transported to the strainer were assumed to pass through the strainer. The total maximum initial debris concentration was determined to be 919.17 ppm, with fiber debris contributing 11.61 ppm, and particulate and coating debris contributing 907.56 ppm. Flowpaths and Alignment Review Both trains of the RHR system, SI system, component cooling system (CCS), and CSS were reviewed to ensure that all of the flowpaths and components impacted by the debris passing through the sump screens were considered. Documents used for this effort included piping and instrumentation diagrams (P&IDs), vendor manuals, equipment specifications, and other documents as applicable. ES-129

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Component Blockage and Wear Evaluations Methodology All component evaluations were performed based on WCAP-16406-P-A. Components addressed in the evaluations include pumps, heat exchangers, orifices, spray nozzles, instrumentation tubing, system piping, and valves required for the post-LOCA recirculation mode of operation of the ECCS and CSS. The evaluations included the following steps:

                                                          ]a,c

2. Provide a summary and conclusions of downstream evaluations.

Response to 3.m.2: Summary and Conclusions of Downstream Evaluations The following is the summary of results and conclusions of the downstream effects evaluations: ECCS/CSS Pumps For pumps, the effects of debris ingestion through the sump* screen on three aspects of operability (hydraulic performance, mechanical-shaft seal assembly performance, ES-130

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary} and mechanical performance) were evaluated. The hydraulic and mechanical performances of the ECCS and CSS pumps were determined to be unaffected by the recirculating sump debris. The mechanical shaft seal assembly performance evaluation resulted in one action item with the suggested replacement of the RHR pumps' carbon/graphite backup seal bushings with a more wear-resistant material, such as bronze. However, because VEGP has an engineered safety feature (ESF) atmospheric filtration system in its auxiliary building, this action is not required per WCAP-16406-P-A. ECCS/CSS Heat Exchangers, Orifices, Spray Nozzles, and System Piping Heat exchangers, orifices, spray nozzles, and system piping were evaluated for the effects of erosive wear for an initial concentration of 919.17 ppm over the mission time of 30 days. The erosive wear on these components was determined to be insufficient to affect system performance. The smallest clearance found for VEGP heat exchangers, orifices, spray nozzles, and system piping in the ECCS recirculation flow path is 0.188 inches, for the SI pressure breakdown orifices. Therefore, no blockage of the ECCS flow path is expected with the current ECCS sump screen hole size of 0.09375 inches. The smallest clearance found in the CSS recirculation flow path is 0.375 inches, for the CS nozzles. No blockage of the CSS flow path is expected because the maximum debris size able to bypass the CSS sump screens is 0.25 in. System piping was evaluated for plugging based on system flow and material settling velocities .. For all piping, the minimum flow velocity was found to be greater than 0.42 ft/s, the minimum velocity required to prevent debris sedimentation. All system piping passed the acceptable criteria for plugging due to sedimentation. ECCS/CSS Valves Valves were evaluated for plugging impact in the downstream effects evaluations. Valves that were determined to be "Not Critical" did not warrant further evaluation, but those valves identified as "Evaluation Required" received a more detailed evaluation. It was determined that all valves passed the acceptance criteria for the plugging evaluation. Valves were evaluated for debris sedimentation. Valves identified as "No Evaluation" did not require additional analysis, but valves identified as "Evaluation Required" were analyzed further. The line velocity for all valves analyzed was found to be greater than [ ]a,c thus, debris sedimentation was not an issue. Valves were evaluated for wear impact. Valves determined to be "Not Critical" did not warrant further evaluation, but valves identified as "Evaluate" were analyzed further. It was found that four throttle valves did not pass the acceptance criteria of [

                                                                  ]a,c Because of this, ES-131

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) they were evaluated for additional opening. This additional opening is necessary for the valves to demonstrate acceptable wear. The turns open value (from a fully closed position) was calculated and rounded to the nearest half turn. With the new

  . calculated turns open value, the valves passed the acceptance criteria of [
                                                              ]a,c ECCS/CSS Instrumentation Lines Instrumentation tubing was evaluated for debris settling. It was found that the largest settling velocity of debris source material found inside a PWR containment is

[ ]a,c as provided by WCAP-16406-P-A. Therefore, as long as the recirculation flow velocity though the ECCS and CSS is greater than [

                                  ]a,c failure of instrumentation due to debris settlement will not occur. The evaluation showed that all flow velocities are greater than [
    ]a,c

3. Provide a summary of design or operational changes made because of downstream evaluations.

Response to 3.m.3: As noted in the Response to 3.m.2, an adjustment to the minimum opening of four throttle valves was required in order to resolve erosion concerns. The minimum turns open value required to demonstrate acceptable valve wear was implemented into plant procedures. The results of the VEGP downstream effects evaluations demonstrate that the evaluated components are acceptable for the expected mission time.

n. Downstream Effects - Fuel and Vessel The objective of the downstream effects, fuel and vessel section is to evaluate the effects that debris carried downstream of the containment sump screens and into the reactor vessel has on core cooling.

1. Show that the in-vessel effects evaluation is consistent with, or bounded by, the industry generic guidance (WCAP-16793-NP), as modified by NRC staff comments on that document. Briefly summarize the application of the methods. Indicate where the WCAP methods were not used or where exceptions were taken, and summarize the evaluation of those areas.

Response to 3.n.1: In-vessel downstream effects for VEGP were evaluated per the methodology in WCAP-16793-NP-A (Reference 22) and the associated NRC SE (Reference 94) using assumed values for in-vessel debris accumulation limits. The evaluation included the following: E5-132

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary)

1. Peak cladding temperature (PCT) due to deposition of debris on fuel rods 0/VCAP-16793-NP-A).

2. Deposition thickness (OT) due to collection of debris on fuel rods 0/VCAP-16793-N P-A).

3. Amount of fiber accumulation at the reactor core inlet and inside the reactor vessel (limits assumed for the purpose of exercising the risk-informed methodology).

These analyses concluded that post-accident long~term core cooling (L TCC) will not be challenged by deposition of debris on the fuel rods, accumulation of debris at the core inlet, or accumulation of debris in the heated region of the core for all postulated LOCAs inside containment. A brief summary of the relevant testing and analyses is provided below as it was used to inform the WCAP evaluations. VEGP Fiber Penetration Testing VEGP conducted fiber penetration testing in 2015. The purpose of the testing was to collect time-dependent fiber penetration data of a prototypical VEGP strainer under various conditions (e.g., approach velocity, water chemistry) and strainer configurations (e.g., number of strainer disks). The test results were used to derive a model that can be used to quantify fiber penetration for the RHR and CS strainers at plant conditions. Twelve penetration tests were conducted, nine of which are useful to inform the resolution of GL 2004-02. Within those nine tests, the approach velocity, the water chemistry, and the number of strainer disks were varied to investigate their effects on fiber penetration. Test Loop Design The test loop consisted of a metal test tank, which housed a test strainer at its downstream end. Water was circulated by a pump through the test strainer, a fiber filtering system, and various piping components. The test tank had a flume geometry, as shown in Figure 3.n.1-1. Debris was introduced in the high-agitation region. This region was equipped with two mechanical mixers to create adequate mixing and prevent the debris from settling. Mixing inside the low-agitation region was created by directing a portion of the returning flow through the perforated bottom plate of the region. This mixing motion kept fiber in suspension without disturbing the fiber bed on the strainer. The strainer region was designed such that the spacing between the test strainer and tank walls imitated the gaps between adjacent strainer stacks, or between the nearest object inside the containment and the strainer. The spacing between the strainer and tank walls was also designed to minimize settling of debris. ES-133

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) FLOW 0 LOW AGITATION HIGH AGITATION REGION REGION ( I

                    /         --------~
                                                 ----~-----1------~

STRAINER __/ l Mechanical Mixers _J Figure 3.n.1-1 General Arrangement of Test Tank The effectiveness of the agitation regions is displayed in Table 3.n.1-1, which documents the quantity of fiber that did not transport to the strainer and was collected from the high or low agitation regions after the conclusion of each test. Table 3.n.1-1: Summary of Useful Fiber Transport Tests Gross Fiber Non-Transported Net Fiber  % of Fiber Test # 1 Added (g) Fiber (g) Added (g) Transport 1 17350.2 236.7 17113.5 98.6% 2 17350.2 397.9 16952.3 97.7% 3 11403.1 203.9 11199.2 98.2% 4 11401.4 205.2 11196.3 98.2% 5 14375.5 190.6 14184.9 98.7% 6 14466.6 171.9 14294.7 98.8% 7 17281.3 578.1 16703.2 96.7% 8 17281.3 278.3 17003.1 98.4% 10 14375.6 0.0 14375.6 100.0% 1 The test numbers shown correspond to the number assigned to each test in the VEGP fiber penetration test report. Note that Tests 9, 11, and 12 are not shown because they were the three tests that were not applicable to this resolution. Test Strainer The test strainer was a prototypical strainer stack; the only difference was the number of disks installed on the test strainer. While the VEGP RHR and CS strainer stacks consist of 16 and 14 disks, respectively, the number of disks on the test strainer was varied among 12, 15, and 18 disks for different tests. The testing flow rate and debris load for each test were determined based on the area ratio of the test strainer to the prototypical strainer assembly, which was varied according to the ES-134

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) number of disks on the test strainer. As a result, the debris load per unit strainer area and water approach velocity were preserved among tests with varying number of strainer disks. This approach allowed the effect of varying the number of strainer disks on penetration quantities to be investigated separately. Though the number of disks for the tested strainer did not exactly match the plant strainers, the tested configurations bounded the plant configurations. As discussed later in this response, the results from those tests with different number of disks were interpolated to derive the penetration model for the plant strainer configurations. Debris Types and Preparation Nukon was the only debris type used in testing. This is appropriate because the only other type of fibrous debris in containment, fire barrier material, transports in negligible amounts of fine fiber. All Nukon used in testing was introduced as fines. Nukon fines were prepared according to the NEI protocol (Reference 46). Nukon sheets, with an overall thickness of 2 inches, were baked single-sided into approximately half the thickness. The heat-treated sheets were then cut into cubes and weighed out according to batch size. Batches were then pressure-washed with test water following the NEI protocol (Reference 46). Debris Introduction Debris was introduced in eight separate batches of increasing batch size for each test. The first two batches corresponded to a theoretical uniform bed thickness of 1/16 inch. The third through seventh batches corresponded to a theoretical uniform bed thickness of 1/8 inch. The final batch corresponded to a theoretical uniform bed thickness of 1/4 inch. The total debris load was equivalent to a theoretical uniform bed thickness of 1 inch. This debris load was chosen because it was sufficient to circumscribe the test strainer. Subsequent debris addition after the development of a circumscribed debris bed would not result in an appreciable amount of penetration. Note that the debris introduction rate was controlled to maintain a prototypical debris concentration in the test tank. Debris Capture Fiber can penetrate through the strainer by two different mechanisms: prompt penetration and shedding. Prompt penetration occurs when fiber reaching the strainer travels through the strainer immediately. Shedding occurs when fiber that already accumulated on the strainer migrates through the bed and ultimately travels through the strainer. Both mechanisms were considered during testing. Fibers that passed through the strainer were collected by the fiber filtering system downstream of the test strainer. The filtering system had 5-micron filter bags installed in filter housings. All of the flow downstream of the strainer travelled through the filter bags before returning to the test tank. The capture efficiency of the filter bags was verified to be above 95 percent. The filtering system allowed the ES-135

     -------------------------------------------~

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) installation of two filter bags in parallel lines such that a filter bag could be left online at all times, even during periods in which filter bags were swapped. Before and after each test, all of the filter bags required for the test were uniquely marked and dried, and their weights were recorded. The weight gain of the filter bags was used to quantify fiber penetration. After testing, the debris-laden filter bags were rinsed with DI water to remove residual chemicals before weighing. For each bag, the drying and weighing process was repeated until two consecutive bag weights (taken at least 1 hour apart) were within 0.05 g of each other. A clean filter bag was placed online before a debris batch was introduced to the test tank and was left on line for a minimum of five pool turnovers (PTOs) to capture the prompt fiber penetration. For Batches 1 and 3, two additional filter bags were used to capture the fiber penetration due to shedding. Before further debris addition, a visual confirmation was required to verify that all introduced debris had transported to the strainer. This approach allowed the testing to capture time-dependent fiber penetration data, which was used to develop a model for the rate of fiber penetration as a function of fiber quantity on the strainer. Test Parameters The test water used for fiber penetration testing had a chemical composition prototypical to VEGP. The plant conditions selected for testin'g were those of minimum and maximum boron concentrations and the corresponding buffer (trisodium phosphate, TSP) concentrations. The low boron concentration was taken from an SBLOCA event, and the high boron concentration was taken from an LBLOCA with the CSS active. For the low boron concentration, a corresponding high plant TSP concentration was used, and the high boron concentration was coupled with a corresponding low plant TSP concentration. The chemical concentrations used in testing are shown in Table 3.n.1-2. Test water was prepared by adding chemicals to DI water until the prescribed concentrations were achieved. Table 3.n.1-2: Summary of Penetration Test Water Chemistry Chemical High Level Low Level Boron (ppm) 2,522 2,169 TSP (ppm) 2,181 3,147 Several different strainer approach velocities, ranging between 0.0043 ft/s and 0.0130 ft/s, were determined from plant operating conditions and were used for the VEGP fiber penetration testing. As shown in Table 3.f.3-1, the 16-disk RHR strainers have a surface area of 677.6 ft2, and the 14-disk CS strainers have a surface area of 590 ft2

• The design flow rates of the RHR and CS pumps are 3,700 gpm and 2,600 gpm, respectively. The average approach velocities of the ES-136

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) RHR and CS strainers are therefore 0.0122 ft/s and 0.0098 ft/s, respectively. Both of these velocities were enveloped by the range tested. In addition to water chemistry and approach velocity, the number of disks was also varied. A test matrix was designed to quantify fiber penetration for different combinations of test conditions. The test matrix is displayed in Table 3.n.1-3. Only the nine tests that are applicable to plant design conditions are shown. Table 3.n.1-3: Large Scale Penetration Test Matrix Approach No. of Boron I TSP Test# Velocity (ft/s) Disks Concentration (ppm) 1 0.0130 18 2,522/2,181 2 0.0043 18 2,522/2,181 3 0.0130 12 2,522/2,181 4 0.0043 12 2,522/2,181 5 0.00314 15 2,522/2,181 6 0.0087 15 2,169 / 3,147 7 0.0043 18 2,169 / 3,147 8 0.0087 18 2,522 / 2,181 10 0.0087 15 2,522 / 2,181 Strainer Penetration Model Development Data gathered from VEGP fiber penetration tests were used to develop a model for quantifying the RHR and CS strainer fiber penetration under prototypical plant conditions. Given the different characteristics and flow rates of the RHR and CS strainers, separate formulas were derived for the two strainers. The models were developed per the following steps:

• General governing equations were developed to describe both the prompt fiber penetration and shedding through the strainer as a function of time and fiber quantity on the strainer. The summation of the developed equations can be used to describe total fiber penetration. The equations contain coefficients whose values were determined separately for each test based on the test results.
• The results for each test were fit to the governing equations using various optimization techniques to refine the coefficient values. This produced a unique set of equations and thus a unique penetration model for each test. Figure 3.n.1-2 compares the curve fit with the test data for Test 8. As the figure shows, it is the summation of the prompt and shedding penetration curves that was fit to the test data. Since three parameters were varied during testing (approach velocity, water chemistry, and number of disks), the fitting coefficients are functions of these three parameters.

ES-137

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) 1800 1600 ... _ 1400 ~~ 1200 { _/ bO

                              ~

VI VI ~ 1000 r -c Q) 4J -

...ro
                                                                                                  ~

4J 800 - Prompt Bypassed Mass f-Q) c Q)

                                                                       -   Shed Mass a_
                          ~                                                                        I 600
                                                                       -   Total Penetrated Mass r

___.! Actual Penetration [g) 400 . A ... { 200 0 0 10000 20000 30000 40000 50000 60000 Time [s] Figure 3.n.1-2: Test 8 Penetration Model Fit

• The test models from the previous step were then used to develop separate models for the RHR and CS strainers , respectively, by interpolating from the model coefficients of each test. During this process , different weighting factors were applied to each test model based on the similarity of the test conditions to the conditions of the actual RHR and CS strainer configurations (as shown in Table 3.n.1-4). Since the RHR and CS strainer conditions are different, each test was weighted separately for the RHR and CS model according to the similarity of its conditions to those of the RHR and CS strainers . Note that the RHR strainer configuration modeled below is based on the strainer modification described in the Response to 3.j .1. The effects of water chemistry on the penetration model were accounted for differently. Instead of interpolating between tests , the low-boron/high-TSP tests were used to represent the water chemistry of interest and were weighted more heavily when calculating the model parameters. This is conservative , firstly , because the large-scale fiber penetration test report showed that tests using low-boron and high-TSP concentrations resulted in higher fiber penetration under otherwise identical conditions , as shown in Figure 3.n .1-3.

Secondly, the low-boron water chemistry conditions used in testing were E5-138

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) determined with the intent of finding the minimum sump boron concentration by using an unlikely combination of conservative inputs. Table 3.n.1-4: RHR and CS Strainer Model Parameters Strainer Stack Disk# Velocity (ft/s) Boron (ppm) RHR 16 0.0122 2,169 cs 14 0.0098 2,169 The resulting RHR fiber penetration model was applied to prototypical plant conditions to calculate the total fiber penetration as a function of time , which is shown as the dark solid line in Figure 3.n.1-3. The curve fit models for the actual test cases were also applied to the same plant conditions , and the results are shown in Figure 3.n.1-3 as lines of different colors. It should be noted that the bracketed values shown in the legend of the figure are the parameters used for each test, not the model conditions, which are common for each curve . As shown in the figure , the model developed for the RHR strainer provides a higher total penetrated fiber quantity than all of the test conditions . This is expected because the RHR strainer model was developed with high approach velocity and low boron concentration (referred to as "low chem" in the figure legend) , which are the most conservative values for those parameters, and none of the tests were run with that combination . As shown in the figure , high approach velocity and low boron concentration , or "low chem" , increase fiber penetration under otherwise identical conditions . 16000 14000

               ~
               ~
                                                                    -   RHR [16 Disk, 0.0122 ft/s, Low Chem]
                                                                    - - Test 8 (18 Disk, 0.00876 ft/s. High Chem]

12000 {;// Ei Ji ii: "O 10000 fl/ / Test 1 [18 Disk, 0.01302 ft/s, High Chem] Test 7 [18 Disk, 0.00435 ft/s, Low Chem] Test 6 [15 Disk, 0.00879 ft/s, Low Chem] 2

~

Q) cQ) Q_ 8000 6000 r; - Test 10 [15 Disk, 0.00885 ft/s, High Chem] Test 3 [12 Disk, 0.01323 ft/s, High Chem]

            ' ~-

'iii 0 - - Test 2 [18 Disk, 0.00433 ft/s, High Chem] I-r 4000 - - Test 5 [15 Disk, 0.00317 ft/s, High Chem]

                                                                    -   Test 4 (12 Disk, 0.00443 ft/s, High Chem]

2000 0 0 20000 40000 60000 80000 100000 120000 Time[s] Figure 3.n.1-3: Comparison of RHR Model with Test Cases at Plant Scale It should be noted that the 95% confidence interval uncertainty in the RHR model output is 137.7 g. This quantity is not included in the model output displayed in E5-139

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Figure 3.n .1-3, but it is added to the calculated in-vessel fiber accumulation quantity found in the in-vessel hand calculation discussed later in this section . Figure 3.n.1-4 shows the prompt fiber penetration fraction as a function of fiber quantity on the strainer derived using the RHR strainer model. As expected , the prompt penetration fraction decreases as a fiber debris bed forms on the strainer. Because shedding penetration is a function of both fiber quantity on the strainer and time , it cannot be similarly shown . 45% 40% 35% 30% +-

          '*:ti"'a. 25%

co> a.E 20%

            ~

0.. 15% 10% 5% 0% 0 2000 4000 6000 8000 10000 12000 14000 16000 18000 Fiber on Strainer [g] Figure 3.n.1-4: Prompt Fiber Penetration Fraction for RHR Strainer Model In-Vessel Effects Evaluations Peak Cladding Temperature and Deposition Thickness The LOCA deposition model (LOCADM) , which is contained as part of WCAP-16793-NP-A (Revision 22) , was used to determine the scale thickness due to deposition of debris that passes through the strainer on the fuel rod surfaces and the resulting peak cladding temperature. The calculated scale thickness was then combined with the thickness of existing fuel cladding oxidation and crud build-up to determine the total deposition thickness . The calculated total deposition thickness and peak cladding temperature were compared with the acceptance criteria provided in WCAP-16793-NP. Note that the VEGP evaluation also considered the applicable E5-140

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) ,. requirements and recommendations from the following PWROG letters: OG-07-419, OG-07-534, OG-08-64, and OG-10-253. Two different cases were considered in this evaluation per the WCAP: minimum containment sump pool volume (Case 1) and maximum containment sump pool volume (Case 2). Table 3.n.1-5 below summarizes the maximum PCT and DT for these two cases. Table 3.n.1-5: Summary of PCT and OT for Cases 1 and 2 PCT (°F) DT (mils) Case Acceptance Acceptance Results Results Criteria Criteria Case 1: Minimum Initial 412 22.7 Sump Pool Volume

                                                < 800                    < 50 Case 2: Maximum Initial 412                     22.3 Sump Pool Volume For either case, the PCT is much lower than the acceptance criterion of 800 degrees F, and the DT is well within the acceptance criterion of 50 mils.

Therefore, deposition of post-LOCA chemical precipitate on the fuel rods will not block the LTCC flow through the core, nor will it create unacceptable local hot spots on the fuel cladding surfaces. The list below summarizes the key inputs and conservatisms used in the LOCADM analysis:

1. When calculating the "bump-up" factor to account for the fiber that passes through the strainer, a bounding value of 100 g/FA was used. As shown in Table 3.n.1-6, this quantity bounds the actual fiber loads for VEGP.

2. The surface area of aluminum coatings was conservatively calculated with operating margin.

3. The maximum sump pH, rather than the actual sump pH profile, was used for the entire 30-day mission time. This is conservative because higher sump pH values result in greater DT.

4. A combination of inputs was used to conservatively determine the PCT and/or DT in LOCADM.

a. Spray was assumed to start immediately and continue for 30 days after the beginning of the LOCA.

b. A conservatively high value for sump temperature was used to set the density of the sump water in order to minimize its mass for the given volume, thereby resulting in higher chemical concentrations within the sump.

c. A conservatively high value for reactor vessel coolant temperature was used to set the density of the reactor coolant in order to minimize its mass for the ES-141

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-f>roprietary) given volume, thereby resulting in higher chemical concentrations within the reactor vessel. The fiber limit at the reactor core inlet given in WCAP-16793-NP-A (15 g/FA) was not used. Instead, accumulation of fiber at the reactor core inlet and inside the reactor vessel was evaluated using the methodology described as discussed in the following section titled "Accumulation of Fiber inside Reactor Vessel." The NRC Safety Evaluation of WCAP-16793-NP provided analysis and recommendations on the use of Westinghouse's WCAP-16793-NP, Revision 2 methodology and identified 14 limitations and conditions that must be addressed. VEGP's responses to these limitations and conditions are summarized below.

1. Assure the plant fuel type, inlet filter configuration, and ECCS flow rate are bounded by those used in the FA testing outlined in Appendix G of the WCAP. If the 15 g/FA acceptance criterion is used, determine the available driving head for a hot leg (HL) break and compare it to the debris head loss measured during the j, FA testing. Compare the fiber bypass amounts with the acceptance criterion given in the WCAP.

VEGP Response: This limitation/condition is associated with the 15 g/FA limit established in WCAP-16793-NP-A, which does not apply to VEGP. In-vessel fiber accumulation is not calculated using WCAP-16793-NP-A.

2. Each licensee's GL 2004-02 submittal to the NRC should state the available driving head for an HL break, ECCS flow rates, LOCADM results, type of fuel and inlet filter, and amount of fiber bypass.

VEGP Response: This limitation/condition is associated with the 15 g/FA limit established in WCAP-16793-NP-A, which does not apply to VEGP. In-vessel fiber accumulation is not calculated using WCAP-16793-NP-A.

3. If a licensee credits alternate flow paths in the reactor vessel in their LTCC evaluations, justification is required through testing or analysis.

VEGP Response: This limitation/condition is associated with the 15 g/FA limit established in WCAP-16793-NP-A, which does not apply to VEGP. In-vessel fiber accumulation is not calculated using WCAP-16793-NP-A. ES-142

Enclosure 5 Suppremental Response to NRC Generic Letter 2004-02 (Non-Proprietary)

4. The numerical analyses discussed in Sections 3.2 and 3.3 of the WCAP should not be relied upon to demonstrate adequate LTCC.

VEGP Response: VEGP does not use any of the conclusions drawn based on the fuel blockage modeling discussed in Sections 3.2 and 3.3 of the WCAP report. In-vessel fiber accumulation is not calculated using WCAP-16793-NP.

5. The SE requires that a plant must maintain its debris load within the limits defined by the testing (e.g., 15 g/FA), and any debris amounts greater than those justified by generic testing in the WCAP must be justified on a plant-specific basis.

VEGP Response: This limitation/condition is associated with the 15 g/FA limit established in WCAP-16793-NP-A, which does not apply to VEGP. In-vessel fiber accumulation is not calculated using WCAP-16793-NP-A.

6. The debris acceptance criterion can only be applied to fuel types and inlet filter configurations evaluated in the WCAP FA testing.

VEGP Response: This limitation/condition is associated with the 15 g/FA limit established in WCAP-16793-NP-A, which does not apply to VEGP. In-vessel fiber accumulation is not calculated using WCAP-16793-NP-A.

7. Each licensee's GL 2004-02 submittal to the NRC should compare the PCT from LOCADM with the acceptance criterion of 800 degrees F.

VEGP Response: As shown in Table 3.n.1-5 above, the calculated VEGP PCT is well within the acceptance criterion of 800 degrees F.

8. When utilizing LOCADM to determine PCT and OT, the aluminum release rate must be doubled to predict aluminum concentrations in the sump pool in the initial days following a LOCA more accurately.

VEGP Response: The methodology outlined in PWROG Letter OG-08-64 was followed to double the aluminum release rate in the LOCADM analysis.

9. If refinements specific to the plant are made to the LOCADM to reduce conservatisms, the licensee should demonstrate that the results still adequately bound chemical product generation.

VEGP Response: The VEGP LOCADM runs do not employ any conservatism-reducing refinements specific to the plant. Therefore, no additional justification is required. ES-143

Enclosure 5 Supplemental Response to NRC Generic Letter"2004-02 (Non-Proprietary)

10. The recommended value for scale thermal conductivity of 0.11 BTU/(h-ft-°F) should be used for LTCC evaluations.

VEGP Response: As stated in Appendix E of WCAP-16793-NP, the recommended thermal conductivity of 0.11 BTU/(h-ft-°F) can be converted to 0.2 W/m-K, which is used in the LTCC calculation.

11. The licensee's submittals should include the means used to determine the amount of debris that bypasses the EGGS sump strainer and the fiber loading at the fuel inlet expected for the HL and cold leg (CL) break scenarios. Licensees should provide the debris loads, calculated on a fuel assembly basis, for both the HL and CL break cases in their GL 2004-02 responses.

VEGP Response: This limitation/condition is associated with the 15 g/FA limit established in WCAP-16793-NP-A, which does not apply to VEGP. In-vessel fiber accumulation is not calculated using WCAP-16793-NP-A.

12. Plants that can qualify a higher fiber load based on the absence of chemical deposits should ensure that tests for their conditions determine limiting head losses using particulate and fiber loads that maximize the head loss with no chemical precipitates included in the tests. In this case, licensees must also evaluate the other considerations discussed in the first limitation and condition.

VEGP Response: This limitation/condition is associated with the 15 g/FA limit established in WCAP-16793-NP-A, which does not apply to VEGP. In-vessel fiber accumulation is not calculated using WCAP-16793-NP-A.

13. The size distribution of the debris used in the FA testing must represent the size distribution of fibrous debris expected to pass through the EGGS sump strainer at the plant.

VEGP Response: This limitation/condition is associated with the 15 g/FA limit established in WCAP-16793-NP-A, which does not apply to VEGP. In-vessel fiber accumulation is not calculated using WCAP-16793-NP-A

14. Each licensee's GL 2004-02 submittal to the NRG should not utilize the "Margin Calculator" as it has not been reviewed by the NRG.

VEGP Response: The VEGP evaluation does not use the "Margin Calculator. In summary, the evaluation showed that the peak cladding temperature and deposition thickness due to accumulation of debris on the fuel rods met the acceptance criteria and did not cause any failures. ES-144

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Accumulation of Fiber inside Reactor Vessel During the post-LOCA sump recirculation phase, debris that passes through the strainer could accumulate at the reactor core inlet or inside the reactor vessel, thereby potentially challenging LTCC. This effect is evaluated for both hot leg break (HLB) and cold leg break (CLB) scenarios using the assumed acceptance criteria. The evaluation used time-dependent fiber penetration fractions obtained from VEGP testing based on plant-specific inputs, as described earlier in this response. The penetration fraction varies with the amount of fiber on the strainer and the amount of time passed since the onset of recirculation. The evaluation was performed in the NARWHAL CFP calculation as well as a bounding hand calculation. The NARWHAL model used assumed values and acceptance criteria to evaluate every break in a self-consistent and time-dependent manner. The model showed no failures for any of the postulated breaks due to accumulation of fiber at the core inlet or inside the reactor vessels. Impact on the results due to variabilities in the inputs was evaluated by sensitivity analyses. The hand calculation served as a bounding evaluation in which the worst case combination of input parameters (e.g., pool volume, transport fiber load, number of RHR and CS trains in operation, RHR and CS pump flow rates, sump recirculation and hot leg switchover times, and CS duration) were used. The uncertainty of the fiber penetration model was added to the calculated fiber quantities for conservatism. The results of the hand calculation are summarized in the table below and are compared with the acceptance criteria. The resulting fiber quantities for both the HLB and CLB are bounded by the assumed acceptance criteria. This conclusion is consistent with the results of the NARWHAL CFP calculation that showed no failures due to accumulation of fibers at the core inlet or inside the reactor vessel. Table 3.n.1-6: Bounding Fiber Loads for HLB and CLB Scenarios a,c In summary, no failures were recorded for any of the postulated breaks due to accumulation of debris at the core inlet or inside the reactor vessel. ES-145

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary)

o. Chemical Effects The objective of the chemical effects section is to evaluate the effect that chemical precipitates have on head loss and core cooling.

1. Provide a summary of evaluation results that show that chemical precipitates formed in the post-LOCA containment environment, either by themselves or combined with debris, do not deposit at the sump screen to the extent that an unacceptable head loss results, or deposit downstream of the sump screen to the extent that long-term core cooling is unacceptably impeded. Content guidance for chemical effects is provided in Enclosure 3 to a letter from the NRC to NEI dated March 2008 (Reference 106).

Response to 3.o.1: The chemical effects strategy for VEGP includes:

• Quantification of chemical precipitates using the WCAP-16530-NP-A methodology with refinements for phosphate passivation of aluminum surfaces.
• Introduction of those pre-prepared precipitates in prototypical array testing.
• Application of an aluminum solubility correlation and integrated autoclave chemical test results to determine formation temperature/timing.
• Time-based determination of acceptable head losses.
• Extrapolation of the resulting head losses to 30 days.

As discussed in the Response to 3.a.1, VEGP has evaluated breaks at all Class 1 weld locations on the primary RCS piping, upstream of the first isolation valve. The amount of chemical precipitates was quantified individually for each of these breaks using the amount of LOCA generated debris for that respective break location. Other plant-specific inputs such as pH, temperature, aluminum amount, and spray times were selected to maximize the generated amount of precipitates. These amounts were scaled by the ratio of the test strainer area to the plant strainer surface area and are compared with the chemical debris quantities used in the prototypical array tests to determine the resulting head loss across the strainers. Before the tests were conducted, the SAS and calcium phosphate were prepared according to the WCAP-16530-NP-A "recipes" and were verified to meet the settling criteria within 24 hours of the test. During the test, a fiber and particulate debris bed was established on the strainer surfaces, the stabilization criteria was satisfied, and the pre-prepared precipitates were added to the test tank in batches. See the Response to 3.f.4 for further details on the head loss measured after introduction of chemical precipitates. See the Response to 3.f.10 for further details on how the chemical precipitate head loss was utilized in the NARWHAL CFP calculation. See the in-vessel effects evaluations in the Response to 3.n.1 for the evaluation of chemical precipitate deposition on the fuel rod surfaces. ES-146 I

                                                                                  ---- - - - - _ _ _ _ _I

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary)

2. Content guidance for chemical effects is provided in Enclosure 3 to a letter from the NRC to NEI dated March 2008 (Reference 106).

Response to 3.o.2: The NRC identified evaluation steps in "NRC Staff Review Guidance Regarding Generic Letter 2004-02 Closure in the Area of Plant-Specific Chemical Effect Evaluations" in March of 2008 (Reference 106). VEG P's responses to the GL Supplement Content evaluation step for each debris characteristic are i I I summarized below.

1. Sufficient 'Clean' Strainer Area: Those licensees performing a simplified chemical effects analysis should justify the use of this simplified approach by providing the amount of debris determined to reach the strainer, the amount of bare strainer area and how it was determined, and any additional information that is needed to show why a more detailed chemical effects analysis is not needed.

VEGP Response: As discussed in the Response to 3.a.1, VEGP has evaluated breaks at all ISi weld locatio_ns on the primary RCS piping, upstream of the first isolation valve. Many of the breaks analyzed resulted in fiber loads sufficient to fully cover the sump strainer screens. Therefore, VEGP is not crediting clean strainer area to perform a simplified chemical effects analysis. See the Figure 1 flow chart in "NRC Staff Review Guidance Regarding Generic Letter 2004-02 Closure in the Area of Plant-Specific Chemical Effect Evaluations" from March 2008 (Reference 106).

2. Debris Bed Formation: Licensees should discuss why the debris from the break location selected for plant-specific head loss testing with chemical precipitate yields the maximum head loss. For example, plant X has break location 1 that would produce maximum head loss without donsideration of chemical effects.

However, break location 2, with chemical eff~cts considered, produces greater head loss than break location 1. Therefore, the debris for head loss testing with chemical effects should be based on break location 2. VEGP Response: Three head loss tests were completed for VEGP: thin bed, full load, and confirmatory full load. The full load produced the highest head loss at each stage of the test, as shown in the table below. Therefore, the full load test was utilized to develop the contributions from conventional debris, calcium phosphate, aluminum precipitates, and the 30-day extrapolation. Table 3.o.2-1 lists the chemical head loss contributions. ES-147

Enclosure 5 I Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Table 3.o.2-1: Chemical Head Loss Values Confirmatory Thin Bed Full Load Test Point Full Load (ft-H20) (ft-H20) (ft-H20)

onventional Debris 0.625 5.46 . 3.50 Full Calcium Phosphate Load 1.65 6.57 5.75 Full Aluminum Precipitate Load 2.60 11.81 8.99 130-day Extrapolation 3.15 15.70 11.12 See the Response to 3.f.10 for additional chemical head loss information.

3. Plant-Specific Materials and Buffers: Licensees should provide their assumptions (and basis for the assumptions) used to determine chemical effects loading: pH I. range, temperature profile, duration of containment spray, and materials expected to contribute to chemical effects.

VEGP Response: The VEGP chemical model requires a number of plant-specific inputs. Each input is chosen to maximize the calculated quantity and minimize the solubility (aluminum only) of the chemical precipitates. VEGP uses TSP to buffer the post-LOCA containment sump pool to a final pH between 7.12 and 7.78. In order to maximize chemical release, TSP is conservatively assumed to dissolve immediately, and the maximum pH of 7.8 was used for the containment sump pool for the entire 30-day event and for the containment spray while recirculating from the containment sump pool. A maximum pH of 5.72 was used for the containment spray during the post-LOCA RWST injection mode. To minimize aluminum solubility, the minimum containment sump pool pH of 7.0 was used, and precipitation was forced at 24 hours whether the solubility limit was reached or not. Different pH values for release and solubility were combined in a non-physical way, bounding the effects

'*'     of all potential pH profile variations. The pH values are summarized in Table 3.o.3-1:

E5-148

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Table 3.o.3-1: VEGP pH Values Sump and Recirculation Spray pH Used 7.8 To Determine Chemical Release Rates Maximum VEGP Long-Term Sump pH 7.78 Minimum VEGP Long-Term Sump pH 7.12 Sump pH Used To Determine Aluminum 7.0 Solubility Injection Spray pH Used To Determine 5.72 Chemical Release Rates The maximum temperature profiles of containment and the sump pool used for the analysis are from the double-ended pump suction LOCA with minimum safeguards case. The total amount of unsubmerged aluminum exposed to containment sprays was assumed to be 926.6 ft2. The total amount of submerged aluminum exposed to the containment sump fluid at VEGP Units 1 and 2 was assumed to be 348.4 ft2. The total amount of concrete assumed to be exposed and submerged in th.e containment sump pool is 10,000 ft2. The quantity of chemical precipitates is negligibly impacted by this large assumed surface area of exposed concrete. Therefore, exposed concrete is not a significant impact to chemical product generation in the VEGP post-LOCA containment sump pool and is not tracked for this purpose. The NARWHAL software (see Enclosure 3, Section 13.1 for general description of the software) analysis accounts for the change in water volume with respect to time and uses assumptions that minimize the water volume in containment. The water volume is dependent on break size, break location, and whether the containment sprays actuate, among other factors. The break size affects the volume in that the accumulators do not inject for breaks less than 2 inches. The RCS holdup volume is dependent on the break location/elevation. Finally, the containment spray activation affects the volume of water in transit. It is acknowledged that water volume has competing effects with respect to chemical release versus solubility; therefore, water volume was included in the sensitivities evaluated in Enclosure 3. A spray duration of 24 hours is used in the analysis. Although there is some operational response flexibility in the spray duration, this is a reasonably conservative assumption because sprays are required to operate for at least 2 hours if they are initiated (assuming there are no containment radiation monitor alarms). Phosphate inhibition minimizes the effect of chemical release from extended spray durations. Enclosure 3, Section 14.3 ES-149

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) includes sensitivity studies with both 2 hour and 30-day containment spray durations. Variable Debris Amounts Of the debris types described in WCAP-16530-NP-A, both E-Glass (Nukon) and lnteram fire barrier are present in the VEGP containments. The quantities of these materials are specific to each break analyzed. Figure 3.a.3-1 through Figure 3.a.3-8 show the ranges of Nukon and lnteram debris versus break size for DEGBs and partial breaks. The mass of latent fiber included as E-Glass for all breaks is 30 lbm. lnteram is a potential source of aluminum from its metal foil surface and is a source of leachable silicon (Reference 73). The lnteram fire barrier at VEGP uses stainless steel foil and is, therefore, not a source of aluminum. Additionally, as discussed in the Response to 3.o.2.7.i, silicon release is not tracked because aluminum is assumed to precipitate only as SAS. Therefore, lnteram does not impact the VEGP chemical model.

4. Approach to Determine Chemical Source Term (Decision Point): Licensees should identify the vendor who performed plant-specific chemical effects testing.

VEGP Response: VEGP is using the separate chemical effects approach to determine the chemical source term. Alien Science and Technology Corporation performed the testing in their test lab in Warrenville, IL.

5. Separate Effects Decision (Decision Point): Within this part of the process flow chart, two different methods of assessing the plant-specific chemical effects have been proposed. The WCAP-16530-NP-A study (Box 7 WCAP Base Model) uses predominantly single-variable test measurements. This provides baseline information for one material acting independently with one pH-adjusting chemical at an elevated temperature. Thus, one type of insulation is tested at each individual pH, or one metal alloy is tested at one pH. These separate effects are used to formulate a calculational model, which linearly sums all of the individual effects. A second method for determining plant-specific chemical effects that may rely on single-effects bench testing is currently being developed by one of the strainer vendors (Box 6, AECL).

VEGP Response: VEGP is primarily using the WCAP-16530-NP-A chemical effects base model to determine the chemical source term. Refinements to this model for aluminum solubility and phosphate inhibition of aluminum release from metallic aluminum are discussed in the Response to 3.o.2.8 and Response to 3.o.2.9.i. E5-150

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary)

6. AECL Model:

i. Since the NRG is not currently aware of the complete details of the testing approach, the NRG staff expects licensees using it to provide a detailed discussion of the chemical effects evaluation process along with head loss test results.

VEGP Response: This question is not applicable because VEGP is not using the AECL model. See the Figure 1 flow chart in "NRG Staff Review Guidance Regarding Generic Letter 2004-02 Closure in the Area of Plant-Specific Chemical Effect Evaluations" from March 2008 (Reference 106). ii. Licensees should provide the chemical identities and amounts of predicted plant-specific precipitates. VEGP Response: This question is not applicable because VEGP is not using the AECL model. See the Figure 1 flow chart in "NRG Staff Review Guidance Regarding Generic Letter 2004-02 Closure in the Area of Plant-Specific Chemical Effect Evaluations" from March 2008 (Reference 106).

7. WCAP Base Model:

i. Licensees proceeding from block 7 to diamond 10 in the Figure 1 flow chart

[in Enclosure 3 to a letter from the NRG to NEI dated March 2008 (Reference 106)] should justify any deviations from the WCAP base model spreadsheet (i.e., any plant specific refinements) and describe how any exceptions to the base model spreadsheet affected the amount of chemical precipitate predicted. VEGP Response: The VEGP chemical model includes quantification of chemical precipitates using the WCAP-16530-NP-A (Reference 73) methodology with refinements for phosphate passivation of aluminum surfaces and the application of an aluminum solubility correlation to determine formation temperature/timing. Silicon inhibition of aluminum release is not credited. Refinements to this model for aluminum solubility and phosphate inhibition of aluminum release from metallic aluminum are discussed in the Response to 3.o.2.9.i. The chemical precipitates assumed by the VEGP chemical model are calcium phosphate (Ca3(PQ4)2) and SAS (NaAISbOa). Although the WCAP-16530-NP-A model typically includes aluminum oxyhydroxide ES-151

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) (AIOOH), the VEGP model assumes that all precipitated aluminum forms SAS, independent of the quantity of silicon available. Per the WCAP-16530-NP-A SE, both aluminum precipitates are acceptable surrogates for aluminum precipitate in head loss testing. As described in Enclosure 3, Section 8.0, the chemical quantity methodology is integrated into the NARWHAL software. The quantity of chemical precipitate is determined for each break as a function of thermal hydraulic inputs and debris generation quantities. The temperature profiles, pH profiles, aluminum metal surface area, and concrete surface area are constant for each break and were selected to maximize aluminum and calcium release and minimize aluminum solubility. The sump fluid volume, spray duration, and debris quantities are break-dependent variables in the NARWHAL calculations. There are two parts to the determination of the chemical precipitate quantity: the elemental chemical release from substrates in containment and chemical product formation. Elemental Chemical Release The two classifications of substrates for which chemical release is analyzed are debris and exposed surfaces. Fiber debris and lnteram debris contribute to chemical release, which is quantified using the WCAP-16530-NP-A release equations. Note that the quantity of each of these debris types is break-specific; therefore, the quantity of elemental chemical release will vary for each break analyzed. Also, note that the lnteram only releases silicon, which does not contribute to the chemical precipitates being tracked for VEGP (see chemical product formation discussion below). This is because SAS is the only aluminum precipitate that is being tracked in the VEGP NARWHAL model, and NARWHAL conservatively assumes an infinite source of silicon when SAS is the only aluminum precipitate tracked. The amount of elemental chemical release from a given debris source is limited by the quantity. Table 3.o.2.7-1 shows the chemical limits of fiberglass used in the chemical release model. Table 3.o.2.7-1: Chemical Mass Limits E-Glass Aluminum Mass Available per Material Mass 1.95% Calcium Mass Available per Material Mass 2.16% The exposed surfaces include aluminum metal and concrete surfaces that either are submerged in the containment pool or are exposed to containment E5-152

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) sprays. The same surface areas were analyzed for each break. Table 3.o.2.7-2 shows these areas. Note that the chemical release from exposed concrete was evaluated using the WCAP-16530-NP-A release equations, and the chemical release from aluminum was evaluated using the University of New Mexico (UNM) release equations. Table 3.o.2.7-2: Other Source Release Values Surface Area Substrate (ft2) Aluminum Metal: Submerged 348.4 Aluminum Metal: Unsubmerged 926.6 Exposed Concrete: Submerged 10,000 Exposed Concrete: Unsubmerged 0 Chemical Product Formation The chemical precipitates analyzed for the VEGP NARWHAL CFP calculation are SAS and calcium phosphate. The calcium phosphate is conservatively assumed to precipitate immediately. The SAS precipitates when the concentration of aluminum in the pool exceeds the aluminum solubility limit as calculated with the Argonne National Laboratory (ANL) solubility equation. Note that if precipitation of SAS is not predicted before 24 hours, then precipitation is forced at that time. Also, note that aluminum does not remain dissolved in the pool after precipitation occurs. Forcing precipitation at 24 hours and not taking credit for aluminum remaining dissolved in the pool are conservative factors in the chemical product formation model. ii. Licensees should list the type (e.g., AIOOH) and amount of predicted plant-specific precipitates. VEGP Response: Chemical precipitate quantities were calculated in the NARWHAL CFP calculation and in a bounding hand calculation. The NARWHAL calculation performs comprehensive evaluation of GSl-191 phenomena in a self-consistent and time-dependent manner. It should be noted that the chemical debris quantities used for quantifying head loss were directly calculated in NARWHAL, not from the hand calculation. The NARWHAL calculation uses the plant-scale precipitate loads from the 2009 head loss testing as the maximum debris limit acceptance criteria (see Response to 3.f.5). The results from the hand calculation are provided below as bounding numbers. The bounding precipitate surrogate masses that would be generated at VEGP are 40.2 kg (88.6 lbm) SAS and 63.2 kg (139.3 lbm) calcium phosphate. ES-153

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) These precipitate masses represent the bounding quantity of aluminum and calcium that could precipitate in the VEGP Unit 1 and Unit 2 containment sump pool. Both AIOOH and SAS chemical surrogates are considered equivalent generators of head loss across a debris bed. Therefore, AIOOH and SAS surrogates may be substituted for each other stoichiometrically relative to aluminum. The maximum temperature where aluminum precipitation could occur in the containment sump pool was calculated to be 136.8 degrees F. Furthermore, the use of the aluminum solubility model for a TSP buffered solution is supported by the integrated autoclave experiments, as shown in Table 3.o.2.9-2. However, as the duration of these experiments was only 24 hours, aluminum precipitation is assumed to occur at 24 hours for calculating strainer head loss unless the solubility model predicts precipitation at an earlier time. Calcium phosphate is assumed to precipitate at all temperatures. These results are bounding for both Unit 1 and Unit 2.

8. WCAP Refinements: State whether refinements to WCAP-16530-NP-A were utilized in the chemical effects analysis.

VEGP Response: The chemical effects strategy for VEGP includes quantification of chemical precipitates using the WCAP-16530-NP-A (Reference 73) methodology with refinements for phosphate passivation of aluminum surfaces and the application of an aluminum solubility correlation to determine formation temperature/timing. Silicon inhibition of aluminum release was not credited. Refinements to the model for aluminum solubility and phosphate inhibition of aluminum release from metallic aluminum are discussed in the Response to 3.o.2.9.i.

9. Solubility of Phosphates. Silicates and Al Alloys:

i. Licensees should clearly identify any refinements (plant-specific inputs) to the base WCAP-16530-NP-A model and justify why the plant-specific refinement is valid.

VEGP Response: The VEGP chemical model includes quantification of chemical precipitates using the WCAP-16530-NP-A (Reference 73) methodology with refinements for phosphate passivation of aluminum surfaces and the application of an aluminum solubility correlation to determine formation temperature/timing. Silicon inhibition of aluminum release was not credited. ES-154

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 {Non-Proprietary) Phosphate Inhibition of Aluminum Surfaces The release of aluminum from metallic aluminum material into the TSP sump solution was modeled using the equations developed by Howe et al. Equations 3.o.2.9-1 through 3.o.2.9-3 (UNM aluminum release equations) calculate the release rate for aluminum from sprayed and submerged aluminum metal in containment. (Equation 3.o.2.9:-1) 6.2181-4.6454 1000 +1.7716(pH)-1.95SO(pH)TK Rmax = 10 TK 1000 (Equation 3.o.2.9-2) (Equation 3.o.2.9-3) Nomenclature: RAI,m = release rate of aluminum from aluminum metal, mg/(m 2 min) Rmax = non-passivated aluminum release rate, mg/(m 2 min) P = phosphate passivation term, unitless TK =temperature, K TKP =temperature utilized in the phosphate passivation term, K pH = pH at 25 degrees C tTsP = time elapsed with phosphate present in solution, min The above equations were developed in testing that was performed at temperatures from 55 degrees C to 85 degrees C (131 degrees F to 185 degrees F, 328.15 K to 358.15 K) and at pH values from 6.84 to 7.84. The following two constraints were used to extend the applicability of these equations:

1. The passivation term, P, is an exponential decay function that approaches zero as trsP increases. This term models the decrease in aluminum surface area available for release as the passivation layer forms. Since testing was not performed below a pH of 6.84, it is not known if this term is applicable at very low TSP concentrations. Therefore, phosphate is not considered "present in solution" unless the pH is above 6.84, and trsP is held at 0 minutes. When the pH rises above 6.84, as TSP dissolves into solution, the trsP "clock" starts.

In practice, the VEGP analysis assumes that TSP is present in the containment sump pool at the start of the LOCA by assuming that the initial sump pH is at its maximum value of 7.8 (see Response to 3.o.2.3). This assumption conservatively increases the aluminum release rate by non-physically combining the highest sump pH with the high initial sump E5-155

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) temperatures. Similarly, the containment sprays are assumed to be at a pH of 7.8 (TSP is present) immediately upon the switch to recirculation.

2. The phosphate passivation function is not assumed to extrapolate beyond the test temperature range. Therefore, T KP equals T K unless above or below the temperatures described here. For temperatures below 328.15K, T KP equals 328.15K. For temperatures above 358.15 K, T KP equals 358.15 K. The passivation equation predicts faster passivation both above and below this range, which is not justifiable without additional testing.

A comparison of the VEGP chemical model with the WCAP-16530-NP-A model is shown in Figure 3.o.2.9-10 and Figure 3.o.2.9-11. Additionally, the aluminum release rate equations with the above constraints were verified for use by VEGP by modeling several integrated autoclave tests. Test 40-01 used VEGP specific materials in testing to replicate the post-LOCA containment debris amounts reported for VEGP Units 1 and 2. Given that the relevant VEGP specific design inputs are a maximum sump pH of 7.78 and a post-LOCA debris type of E-Glass (i.e., no Calcium Silicate, Silica, or Mineral Wool insulation), tests 39-01, 42 01, and 44-01 also use test inputs similar to that of VEGP [

                  ]a,c These tests were run at a range of temperatures from [
                          ]a,c which bounds the VEGP maximum post-LOCA containment sump pool and containment temperature profiles. Therefore, tests 39-01 (in-bag and out-of-bag), 40-01, 42-01, and 44-01 were simulated using the WCAP-16530-NP-A methodology with the refined aluminum release equation. The critical parameters for the integrated autoclave tests 39-01, 40-01, 42-01, and 44-01 are summarized in Table 3.o.2.9-1 and Figure 3.o.2.9-1.

ES-156

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) a,c Figure 3.o.2.9-1: Integrated Autoclave Test Temperature Profiles Table 3.o.2.9-1: Critical Integrated Autoclave Test Parameters a,c E5-157

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Vogtle's maximum pH (7.78) and the pH of test 44-01 [ ]a,c are within the pH range for the UNM aluminum release equations of 6.84 through 7.84. Therefore, test 44-01 serves as the primary validation of the UNM aluminum release equations in an integrated chemical environment. The aluminum concentration results for test 40-01 and the simulation is provided in Figure 3.o.2.9-2. a,c Figure 3.o.2.9-2: Test 44-01 (pH = 7.52) Aluminum Concentrations Because the maximum pH of Test 44-01 [ ]a,c is below the VEGP maximum pH (7.78), integrated autoclave tests 39-01 [ ]a,c 40-01 [

            ]a,c and 42-01 [       ]a,c were also simulated to validate the UNM aluminum release equations at bounding pH values. However, these tests are above the pH range for the single effects tests (6.84 through 7.84) used to develop the UNM aluminum release equations. The aluminum concentration results for these tests and their simulations are provided in Figures 3.o.2.9-3 through 3.o.2.9-5.

ES-158

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) a,c Figure 3.o.2.9-3: Test 39-01 (pH = 8.00) Aluminum Concentrations a,c Figure 3.o.2.9-4: Test 40-01 (pH = 8.03) Aluminum Concentrations ES-159

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) a,c Figure 3.o.2.9-5: Test 42-01 (pH = 8.03) Aluminum Concentrations In addition to the aluminum results, the integrated autoclave test calcium release results were also compared with the simulation in Figures 3.o.2.9-6 through 3.o.2.9-9 to demonstrate the overall conservatism of the VEGP chemical model. ES-160

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) a,c Figure 3.o.2.9-6: Test 39-01 (pH = 8) Calcium Concentrations a,c I : Figure 3.o.2.9-7: Test 40-01 (pH = 8.03) Calcium Concentrations ES-161

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) a,c Figure 3.o.2.9-8: Test 42-01 (pH = 8.03) Calcium Concentrations a,c Figure 3.o.2.9-9: Test 44-01 (pH = 7.52) Calcium Concentrations E5-162

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) The aluminum release results indicate that the aluminum release rate as a function of time in a TSP buffered solution decreases to approximately zero within 24 hours, as is predicted by the aluminum release equations. For integrated autoclave tests 40-01, 42-01, and 44-01, the measured aluminum concentrations either follow or are bounded by the trends predicted by the UNM aluminum release equations, which justifies the UNM methodology with the extended pH (5 to 7.84) and temperature (104.8 degrees F to 266.5 degrees F) constraints as acceptable. Note that the aluminum concentration for Test 39-01 is under-predicted. The presence of zinc in solution can reduce the rate of aluminum corrosion. A reduction of approximately 2/3 of the released aluminum concentration was observed in the Howe et al. bench tests when zinc coupons were present. The Howe et al. equations do not credit zinc inhibition, which was shown by bench testing to result in an over-prediction of aluminum release under the maximum pH tested of 7.84. However, integrated autoclave tests 39-01, 40-01, and 42-01 are above the pH range for the single effects tests. Of these three integrated autoclave tests, Test 39-01 contained the least amount of galvanized steel I ,. and contained less than 38% (0.117 ft2/0.308 ft 2 ) of the galvanized steel ' ' surface area as Test 40-01, which used VEGP-specific material quantities. Furthermore, the aluminum release result for Test 40-01 is accurately predicted, and Test 42-01 is slightly over-predicted with 238% of the VEGP-specific galvanized steel quantity (0.732 ft2/0.308 ft 2 ). Although these results demonstrate that the aluminum release equations accurately predict aluminum concentrations at elevated pH when VEGP-specific or greater zinc quantities are present, the maximum acceptable pH is not extended above the value of 7.84 as set by Howe, et al. Additionally, as shown in Figures 3.o.2.9-6 through 3.o.2.9-9 calcium concentrations are significantly over-predicted by the WCAP-16530-NP-A model. As discussed in Section 3.o.2.7.i, calcium phosphate is conservatively assumed to form immediately as calcium is released. Finally, as discussed in Section 3.f.10, the full load calcium phosphate head loss is assumed as soon as calcium phosphate starts to accumulate on the strainer. Because the VEGP chemical model results in over-predicted quantities of aluminum and calcium precipitates at VEGP-specific conditions, the overall methodology is conservative for use to determine the precipitate loading for strainer head loss. Solubility of Aluminum The aluminum solubility limit is determined using Equation 3.o.2.9-4, developed by ANL. 26980

• 10(pH+LlpH)-14.4+0.0243T if T < 175 op c - , - (Equation 3.o.2.9-4)

Al.sol - { 26980 . lO(pH+LlpH)-10.41+0.00148T 1 if T > 175 op E5-163

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Nomenclature: L\pH =pH change due to radiolysis acids T =solution temperature, degrees F The aluminum solubility limit equation was used to determine the temperature and timing of aluminum precipitation and to determine the aluminum concentration in solution for use in the aluminum release equations for concrete and insulation. When precipitation is predicted by this equation, the full amount of aluminum released is assumed to precipitate as SAS. The aluminum solubility limit equation was not used to reduce the predicted quantity of precipitate by crediting the amount remaining in solution. Aluminum solubility testing developed by ANL was completed in boric acid/NaOH buffered solutions. As shown in Table 3.o.2.9-2, the method that Vogtle utilizes to predict aluminum precipitation temperature yields much higher temperatures than the filtration tests that did not detect precipitation. This demonstrates the applicability of the ANL equation in a boric acid/TSP buffered solution. Since the tests were performed for a 24-hour duration, the maximum amount of time allowed in the VEGP chemical model for aluminum precipitation to occur was capped at 24 hours. Table 3.o.2.9-2: Aluminum Precipitation Test Results a,c Comparison of VEGP Chemical Model with WCAP-16530-NP-A Enclosure 3 Section 14.4 includes a sensitivity study comparing the VEGP release model with phosphate inhibition credited (UNM Aluminum Metal Release Equation) to the WCAP-16530-NP-A chemical model without this refinement (WCAP-16530 Equation). Additionally, the bounding VEGP case for maximum precipitate generation (based on the hand calculation) was conducted using both the VEGP release model and the WCAP-16530-NP-A model without the refined equations for phosphate inhibition. Note that to address NRC concerns on the use of the WCAP-16530-NP-A for aluminum ES-164

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) metal release rates early in the event, the aluminum metal release rate was doubled for the initial 15 days. The maximum precipitation case uses the pH, temperature, aluminum metal, and concrete inputs described in the VEGP response to 3.o.2.3. The maximum sump pool mass of 6,675,858 lbm was used to maximize the dissolution of aluminum and calcium from concrete and insulation materials. The containment sprays were assumed to remain active for 30 days with a minimum RWST injection duration of 27.9 minutes to maximize aluminum release from sprayed aluminum surfaces. Lastly, the maximum generated quantity of Nukon insulation, 2,229.2 ft3, was used. Note that these inputs apply to this comparison completed in the hand calculation, which were selected to yield bounding results for all break scenarios as opposed to the break-specific and time dependent inputs used in the NARWHAL CFP calculation. Figure 3.o.2.9-10 shows a comparison of the aluminum concentration results for the two models over the full 30-day window, and Figure 3.o.2.9-11 shows the results for the first 24 hours. The UNM curve shows the results with aluminum passivation and the 2xWCAP-16530-NP-A curve shows the results without this refinement. 5 _,.... ---*- **-* *- *-*-. -**-* ---*-**- - . 4.5 4 ................. . ..... -~-.-. *-**** ......, l _3.5 E c.

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0.5 0 7200 14400 21600 28800 36000 43200 Time(min)

                                                     - -       UNM        - 2 x WCAP*16530*NP*A Figure 3.o.2.9-10: Maximum Aluminum Release Cases (30 days)

ES-165

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) 2 1.75 1.5 - *---- *-*:*"" .......- ...- i ..-- *--*-*-*-,----- ---.

        "Ec.                                                  - - ~ - - - - i - - - - -~' - - - - - - - -
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                                 - - UNM               - 2 x WCAP-16530-NP-A Figure 3.o.2.9-11: Maximum Aluminum Release Cases (24 hours)

The use of the aluminum passivation equations clearly decreases the release of aluminum predicted over 30 days. However, as would be expected, the initial release rate with passivation credited (UNM) follows the model without passivation credited (2xWCAP-16530-NP-A) closely over the initial two hours before diverging as the aluminum metal surface area available for release passivates. ii. For crediting inhibition of aluminum that is not submerged, licensees should provide the substantiation for the following: (1) the threshold concentration of silica or phosphate needed to passivate aluminum, (2) the time needed to reach a phosphate or silicate level in the pool that would result in aluminum passivation, and (3) the amount of containment spray time (following the achieved threshold of chemicals) before aluminum that is sprayed is assumed to be passivated. VEGP Response: See the Response to 3.o.2.9.i. ES-166

                                       .Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) iii. For any attempts to credit solubility (including performing integrated testing),

licensees should provide the technical basis that supports extrapolating solubility test data to plant-specific conditions. In addition, licensees should indicate why the overall chemical effects evaluation remains conservative when crediting solubility given that small amount of chemical precipitate can produce significant increases in head loss. VEGP Response: See the Response to 3.o.2.9.i. iv. Licensees should list the type (e.g., AIOOH) and amount of predicted plant-specific precipitates. VEGP Response: See the Response to 3.o.2.7.ii.

10. Precipitate Generation (Decision Point): State whether precipitates are formed by chemical injection into a flowing test loop or whether the precipitates are formed in a separate mixing tank.

VEGP Response: As discussed in the Response to 3.o.2.12, VEGP pre-mixed surrogate chemical precipitates in a separate mixing tank for chemical head loss testing. The direct chemical injection method was not used in head loss testing.

11. Chemical Injection into the Loop:

i. Licensees should provide the one-hour settled volume (e.g., 80 ml of 100 ml solution remained cloudy) for precipitate prepared with the same sequence as with the plant-specific, in-situ chemical injection.

VEGP Response: As discussed in the response to item 3.o.2.12, VEGP pre-mixed surrogate chemical precipitates in a separate mixing tank for chemical head loss testing. The direct chemical injection method was not used in head loss testing. See the Figure 1 flow chart in "NRC Staff Review Guidance Regarding Generic Letter 2004-02 Closure in the Area of Plant-Specific Chemical Effect Evaluations" from March 2008 (Reference 106). ES-167

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) ii. For plant-specific testing, the licensee should provide the amount of injected chemicals (e.g., aluminum), the percentage that precipitates, and the percentage that remains dissolved during testing. VEGP Response: See the Response to 3.o.11.i. iii. Licensees should indicate the amount of precipitate that was added to the test for the head loss of record (i.e., 100 percent, 140 percent of the amount calculated for the plant). VEGP Response: See the Response to 3.o.11.i.

12. Pre-Mix in Tank: Licensees should discuss any exceptions taken to the procedure recommended for surrogate precipitate formation in WCAP-16530-N P-A.

VEGP Response: The chemical head loss tests employed the pre-mixed chemical surrogate

• methodology. The WCAP-16530-NP-A precipitate formation methodology for SAS and calcium phosphate was followed with no exceptions. The 1-hour settling volume for each batch of chemical precipitates was determined at the time that the batch was produced and was required to be 6 ml or greater. The chemical precipitate settling was also required to be measured within 24 hours of the time the surrogate was to be used, and the 1-hour settled volume was required to be 6 ml (SAS and AIOOH) or greater and within 1 .5 ml of the freshly prepared surrogate (Reference 73). Chemical precipitates that failed the criteria of being 6 ml or greater (initial test or re-test) and within 1.5 ml of the freshly prepared surrogate criteria were not used in testing.

13. Technical Approach to Debris Transport (Decision Point): State whether near-field settlement is credited or not.

VEGP Response: VEGP chemical effects testing used agitation and turbulence in the test tank to ensure that essentially all debris analyzed to reach the strainer in the plant reached the strainer in. head loss testing.

                                         .    . VEGP did not credit any near field settlement in head loss testing.

ES-168

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary)

14. Integrated Head Loss Test with Near-Field Settlement Credit:

i. Licensees should provide the one-hour or two-hour precipitate settlement values measured within 24 hours of head loss testing.

VEGP Response: VEGP is not crediting near field settlement of chemical precipitate in chemical head loss testing. See the Figure 1 flow chart in "NRC Staff Review Guidance Regarding Generic Letter 2004-02 Closure in the Area of Plant-Specific Chemical Effect Evaluations" from March 2008 (Reference 106). ii. Integrated Head Loss Test with Near-Field Settlement Credit: Licensees should provide a best estimate of the amount of surrogate chemical debris that settles away from the strainer during the test. VEGP Response: See the Response to 3.o.2.14.i.

15. Head Loss Testing Without Near Field Settlement Credit:

i. Licensees should provide an estimate of the amount of debris and precipitate that remain~ on the tank/flume floor at the conclusion of the test and justify why the settlement is acceptable.

VEGP Response: Even though all debris had an opportunity to collect on the surfaces of the test strainer, a portion of the debris added to the test settled on the floor. Post-test photographs show that while nearly all of the Nukon had reached the strainers, approximately 10-15 percent of the (larger/heavier) dirt/dust surrogate and nearly all of the lnteram debris settled on the floor in front of the strainer array (see Attachment A). Additionally, a minor amount (approximately 10-20 lb.) of silicon carbide settled underneath the simulated containment floor, and less than 1.5 lbm of particulate debris was removed along with the water drained from the tank to ensure sufficient volume in the tank for chemical additions. Finally, a small amount of calcium phosphate (less than 0.25 L of the 480.72 L total) was spilled outside the tank such that it was unrecoverable. Because of the measures taken during the test, as described in Response to 3.f.12, to keep debris suspended and transportable to the test strainer, the amount of settled debris described above is considered acceptable. E5-169

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) ii. Licensees should provide the one-hour or two-hour precipitate settlement values measured and the timing of the measurement relative to the start of head loss testing (e.g., within 24 hours). VEGP Response: See the Response to 3.o.2.12.

16. Test Termination Criteria: Licensees should provide the test termination criteria.

VEGP Response: The head loss was considered stable when the differential pressure across the debris bed changed by less than or equal to 1 percent over a 1-hr period. In addition, the rate of head loss increase was required to be significantly decreasing, or the head loss was required to be consistently steady at termination of the test. ES-170 __I

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary)

17. Data Analysis:

i. Licensees should provide a copy of the pressure drop curve(s) as a function of time for the testing of record .

VEGP Response: The pressure drop curves for the full load test are provided as Figures 3.o.2.17-1 through 3.o.2.17-4 below. Day 1

• Head Loss & Approach Velocity Profile

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             ~~::'.'.:=----1~=~~-t.---__i__ __;_i_-,-__J__                             __J__-,-_L_ _ _ _.,._l_ 0.000 2:04:00 PM       6:52:00 PM    11:40:00 PM Time (H :MM:SS AM /PM)

I- Head Loss -*-* Flow Adjustment - - - *Drain Down ---- Tank Fill - Velocity I Figure 3.o.2.17-1: Full Load Test Differential Pressure and Velocity vs. Time - Day 1 ES-171

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Day 2

• Head Loss & Approach Velocity Profile 7.0 ,---------------~--i~~.~-------~ 0.016

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• Head Loss & Approach Velocity Profile 10.0 0.018

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• Flow Adjustment - Velocity I Figure 3.o.2.17-3: Full Load Test Differential Pressure and Velocity vs . Time - Day 3 E5-172

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Day 4 - Head Loss & Approach Velocity Profile 16.0 ~----~-----~~~---~--~~-~~~ 0 . 018 14.0 120 l!l 10.0 "3:

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2.0 0.002 0.0 +----~-'----~-------r-~--~~--~r-'---'--~-+ 0.000 11:50:00 PM 3:26:00AM 7:02:00 AM 10:38:00 AM 2:14:00 PM 5:50:00 PM 9:26:00 PM Time (H:MM:SS AM/PM) I- Head Loss - * - -Flow Adjustment - Velocity I Figure 3.o.2.17-4: Full Load Test Differential Pressure and Velocity vs. Time - Day 4 ii. Licensees shou ld explain any extrapolation methods used for data analysis . VEGP Response: See the Response to 3.f.10.

18. Integral Generation (Alien) : Licensees should explain why the test parameters (e.g. , temperature , pH) provide for a conservative chemical effects test VEGP Response :

VEGP is using the separate chemical effects approach to determine the chemical source term . Th is section is not applicable to the VEGP chemical effects analysis. E5-173

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) *

19. Tank Scaling I Bed Formation:

i. Explain how scaling factors for the test facilities are representative or conservative relative to plant-specific values.

VEGP Response: VEGP is using the separate chemical effects approach to determine the chemical source term. This section is not applicable to the VEGP chemical effects analysis. ii. Explain how bed formation is representative of that expected for the size of materials and debris that is formed in the plant specific evaluation. VEGP Response: VEGP is using the separate chemical effects approach to determine the chemical source term. This section is not applicable to the VEGP chemical effects analysis.

20. Tank Transport: Explain how the transport of chemicals and debris in the testing facility is representative or conservative with regard to the expected flow and transport in the plant-specific conditions.

VEGP Response: VEGP is using the separate chemical effects approach to determine the chemical source term. This section is not applicable to the VEGP chemical effects analysis.

21. 30-Day Integrated Head Loss Test: Licensees should provide the plant-specific test conditions and the basis for why these test conditions and test results provide for a conservative chemical effects evaluation.

VEGP Response: VEGP is using the separate chemical effects approach to determine the chemical source term. This section is not applicable to the VEGP chemical effects analysis. ES-174 J

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary)

22. Data Analysis Bump Up Factor: Licensees should provide the details and the technical basis that show why the bump-up factor from the particular debris bed in the test is appropriate for application to other debris beds.

VEGP Response: VEGP is using the separate chemical effects approach to determine the chemical source term. This section is not applicable to the VEGP chemical effects analysis.

p. Licensing Basis The objective of the licensing basis section is to provide information regarding any changes to the plant licensing basis due to the sump evaluation or plant modifications.

1. Provide the information requested in GL 2004-02 Requested Information Item 2(e) regarding changes to the plant-licensing basis. The effective date for changes to the licensing basis should be specified. This date should correspond to that specified in the 10 CFR 50.59 evaluation for the change to the licensing basis.

GL 2004-02 Requested Information Item 2(e) A general description of and planned schedule for any changes to the plant licensing bases resulting from any analysis or plant modifications made to ensure compliance with the regulatory requirements listed in the Applicable Regulatory Requirements section of this GL. Any licensing actions or exemption requests needed to support changes to the plant licensing basis should be included. Response to 3.p.1: VEGP is following the "STP Piloted Risk-Informed Approach for GSl-191," (References 44 and 45). The proposed change replacing the current deterministic methodology with a risk-informed methodology requires changes to the descriptions of how VEGP meets 10 CFR 50.46(a)(1 ), GDC 35, GDC 38, and GDC 41. Those changes require exemptions to certain requirements of 10 CFR 50.46(a)(1 ), GDC 35, GDC 38, and GDC 41, and the requests for the exemptions are provided in the future LAR. VEGP's risk-informed approach to assess the effects of LOCA debris replaces the existing deterministic approach described in the VEGP licensing basis. This, in turn, requires an amendment to the VEGP Unit 1 and Unit 2 operating licenses to incorporate the revised methodology per the requirements of 10 CFR 50.59. This proposed amendment to the operating license is included in the future LAR. ES-175

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) 4.0 NRC Request for Additional Information: VEGP received two requests for additional information from the NRC: NL-08-1497 and NL-08-1829 (References 101 and 87, respectively). The first RAI, issued September 17, 2008, addressed critical issues with test protocols used in chemical effects testing performed at the VUEZ facility by ALION Science and Technology. SNC responded to the VUEZ-specific RAls with the issuance of NL-08-1583 (Reference 102), November 7, 2008. SNC determined the need to consider an alternate approach to demonstrating adequate performance of the emergency containment sump strainers after careful consideration of the NRC's concerns with the VUEZ test protocol. Therefore, the VUEZ-specific RAls are no longer applicable and do not require a response, as recorded in NL-08-1583. The second RAI, NL-08-1829, was issued December 2, 2008, containing 29 requests based on the review of all four VEGP Supplemental Responses to GL 2004-02: NL-07-1777, NL-08-0670, NL-08-1155, and NL-08-1228 (References 95, 96, 98, and 100, respectively). The final SNC responses to RAls 1 through 29 are referenced in the table below. Additionally, SNC provided its intended path forward for the resolution of GSl-191 letter to the NRC, NL-13-0953(Reference105). By letter dated November 14, 2013, the NRC sent SNC an RAI. The RAI and SNC's response are included below as provided in SNC letter to the NRC, NL-13-2544 (Reference 110). SNC has since revised the ERGs as discussed in the response below. NRC RAI SN C's May 16, 2013, letter did not identify that SNC had implemented, or identified for future implementation, any mitigative measures to deal with the potential for in-vessel blockage. SNC stated that it was evaluating Westinghouse recommendations for mitigative measures for in-vessel blockage and that appropriate procedure changes and operator training would be completed, as deemed necessary, following the evaluation. Please provide the mitigative measures chosen for the VEGP to deal with in-vessel blockage, should it occur. SNC Response to NRC RAI SNC will make changes to the Vogtle Units 1 and 2 Emergency Response Guidelines (ERGs). Specifically, after transferring to cold leg recirculation, Vogtle will monitor core exit temperatures, injection flow, and reactor vessel level indicating system (RVLIS) indications to identify any abnormal indications. Should abnormal indications exist, realignment to hot leg recirculation and back flowing through the reactor vessel will be necessary. ES-176 L

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) RAI Complete RAls as Provided in NL-08-1829 (Reference 87) Response No. 1 Please provide a description of the jacketing/banding systems 17.0D ZOI is used used to encapsulate Nukon insulation at Vogtle (e.g., on piping, for all Nukon. steam generators, reactor coolant pumps) and during WCAP-16710-P jacketing/banding system qualification testing. The information is no longer should include the jacket materials used in the testing, geometries credited. and sizes of the targets and jet nozzle, and materials used for Because we are jackets installed in the plant. Please provide information that no longer compares the mechanical configuration and sizes of the test crediting the targets and jets, and the potential targets and two-phase jets in the jacketing in the plant. Pleases evaluate how any differences in jet/target sizing ZOI, a response and jet impingement angle affect the ability of the insulation system to this portion of to resist damage from jet impingement. In doing so, please provide the question is a justification for applying debris generation test data obtained for not provided. the Nukon jacketing systems employed at the Wolf Creek and Callaway plants to the jacketing systems used at Vogtle and See the demonstrating that the Vogtle jacketing systems are as resistant to Response to destruction as the jacketing systems tested. In responding to this 3.c.1. question, please address the potential varied jacketing systems for various components of the reactor coolant system, which are within the LOCA ZOls (e.g., piping, coolant pumps, and steam generators). 2 Please specify the ZOI radius used to calculate the quantity of See the lnteram fire barrier debris that could be generated. Please provide Response to the characteristics of the lnteram fire barrier material including the 3.b.1, 3.c.1, and type of lnteram installed and its anticipated debris characteristics. 3.f.4. Please provide information on how the material was prepared for inclusion in head loss testing or provide information on the surrogate material used and its properties. Please provide assumptions made regarding the physical properties of LOCA damaged lnteram fire barrier material and the bases for how any surrogate material used in testing conservatively model these properties. 3 Please provide the following additional information needed to See the support the assumption of 15% erosion of fibrous debris pieces in Response to the containment pool: 3.e.1.

a. The similarity of the flow conditions (velocity and turbulence),

chemical conditions, and fibrous material present in the erosion Note that a tests to the analogous conditions applicable to the expected smaller fraction plant conditions, and for erosion of

b. The durations of the erosion tests and how the test results were fibrous debris extrapolated to the sump performance mission time. pieces in the containment pool is used based on testing.

E5-177

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) RAI Complete RAls as Provided in NL-08-1829 (Reference 87) Response No. 4 On pages E1-19 and E1 -20 of the supplemental response dated No longer February 28, 2008, it is indicated that, based on the fact that less applicable than 25% of the strainer perimeter area is in excess of the curb lift because the curb velocity metric, 25% of small debris pieces are assumed to is not being surmount the curb/plenum on which the strainer modules rest. credited as a However, based on the diagram of containment provided on page debris interceptor. E1-9 of the same supplemental response, the staff expects that it Any debris that is likely that flow and debris will preferentially approach the sump transported I from openings in the shield wall.. As a result, the fraction of debris during f* approaching the sump in the higher velocity flow channel could recirculation was I significantly exceed 25%. In light of the considerations such as assumed to reach this, please provide a technical justification for the assumption that the strainers. I only 25% of small pieces of fibrous debris can surmount the curb/plenum and reach the sump strainers. See the Response to 3.e.4 and 3.e.6. 5 The supplemental response states that the head loss test results See the were scaled to the full-sized strainer system based on Response to temperature, velocity, and bed thickness differences. Without 3.f.10. additional information on the methodology used to make these extrapolations, it is not possible to determine whether they were performed conservatively or prototypically. It appears that the head loss test result of 6.84 ft was extrapolated to 8.126 ft. Please provide the details of all extrapolations performed for the head loss test data. Please include the raw test data and conditions, and the final head loss value and the conditions to which it was extrapolated. Please include any differences in temperature, velocity, bed thickness between the head loss testing and anticipated plant conditions. ES-178 J

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) RAI Complete RAls as Provided in NL-08-1829 (Reference 87) Response No. 6 The supplemental response stated that the submergence value for See the the SBLOCA was not calculated. The stated [reason] for this was Response to 3.f.2 that an SBLOCA would create less debris and therefore result in a and 3.f.3. less challenging head loss. This is true for some portions of the evaluation. However, the vortexing evaluation is often most limiting when there is no debris on the strainer. In addition, if the strainer is not fully submerged, the acceptance criteria for maximum head loss may be limited by the strainer height (50% of the strainer height per RG 1.82, Rev. 3) instead of the pump NPSH margin. This would be a reduction to about 25% of the tested strainer head loss. Also, if the strainer is not submerged, air ingestion would have to be evaluated more rigorously. Un-submerged strainer area cannot be credited to accumulate debris, so other areas of the strainer would have to absorb the debris that cannot be collected on the uncovered portion of the strainer. Due to break location, the SBLOCA level may not include some RCS inventory and also may not include all or part of the accumulator volume. Please provide the minimum submergence for an SBLOCA. If the strainer is not fully submerged for this event, please provide appropriate evaluations for air ingestion and strainer head loss, including acceptability based on the guidance in RG 1.82 (or other appropriate methodology). 7 Related to the RAI above on the response of the plant to cases See the where the strainers may not be fully submerged, have various Response to scenarios such as an SBLOCA with the failure of one train of EGGS 3.g.6 and no CS actuation been considered? For this case all debris would transport to a single EGGS strainer that may not be fully submerged. A thin bed with the bulk of the particulate debris could form on the operating strainer surface. Please provide an evaluation that demonstrates that adequate NPSH margin will be provided to the EGGS pumps (reference Regulatory Guide 1.82, Revision 3, Section 1.3.4.4). 8 It was implied that the debris was added to the testing prior to Debris was added starting the recirculation pump. Please provide justification that after starting the this test sequence provides prototypical or conservative test reci rcu latio n conditions. pump. See the Response to 3.f.4. ES-179

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) RAI Complete RAls as Provided in NL-08-1829 (Reference 87) Response No. 9 Scaling was based on the circumscribed area for the test and plant Scaling was strainers. Normally scaling is based on the screen area. For the based on screen module testing the scaling factors based on circumscribed and area for test and screen area appear to be the same. The scaling factors for the plant strainers. sector testing could not be determined by the staff. Please provide information that justifies the use of the circumscribed area of the See the test and plant strainers for scaling of the sector tests. Response to 3.f.4 and 3.f.10 10 Without information on how debris was prepared and introduced See the into the thin bed tests it cannot be concluded that the thin bed Response to 3.f.4 testing was valid. It is possible that a properly conducted thin bed and 3.f.6. test would result in higher head loss than the full load test that was stated to be the limiting head loss condition for Vogtle. Please provide information that justifies that the sector testing conducted to determine the strainer's ability to deal with a thin bed was conducted under conditions that would conservatively model the debris bed. Please reference the staff Head Loss and Vortexing Guidance for thin bed testing considerations (ADAMS Accession No. ML080230038). 11 The supplemental response stated that air ingestion was evaluated See the at the top of the module. The results of the vortexing evaluation Response to were not provided. Please provide the results of the air 3.f.3. ingestion/vortexting evaluation including the plant conditions assumed. 12 No documentation of fiber size distribution used for testing See the compared to the fiber size distribution predicted to arrive at the Response to strainer was provided. The supplemental response stated that 3.f.4. only fine and small pieces of fiber would be created by the break. The size distribution of the fibrous debris used in testing was not provided. In general shredded fiber does not imply that all fine fibers are created. For thin bed testing, only fine fibrous debris should be added to the test flume until all fibrous fines predicted to be created are added. Please provide information regarding the size distribution of fibrous debris used in various tests and how these size distributions compare to the transport evaluation predictions. E5-180

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) RAI Complete RAls as Provided in NL-08-1829 (Reference 87) Response No. 13 The documentation of fibrous concentration during addition and Tank test was methods of addition to the flume were unclear. Documentation utilized which did should be provided showing that the concentration of debris during not allow settling addition was controlled so that non-prototypical agglomeration of of debris. the debris would not occur. Please provide information that justifies that the debris introduction methods used during testing See the did not result in non-prototypical settling or agglomeration of Response to debris. Also, please include the amounts of debris added during 3.f.4. each addition, the actual size distribution of the debris, and the debris types. 14 Documentation of the amount of debris that settled in the agitated See the and non-agitated areas of the test tanks was not provided. Please Response to 3.f.1 provide the amount of debris that settled in the agitated and non- and 3.f.4. agitated areas of the test tank for each test. 15 There is no discussion of extrapolation of head loss test results to See the ECCS mission times, nor discussion of test termination criteria and Response to subsequent extrapolation. Please provide information that shows 3.f.10. that the head loss testing was run to a maximum value, or that an extrapolation was performed to obtain the head loss at the end of the strainer mission time. Please provide sufficient time dependent test data so that the termination criteria and any extrapolations conducted can be verified. Please provide a graph of the head loss over time for the limiting module and sector tests. Please specify the sector test that created the limiting head loss. 16 The flashing evaluation did not describe the margin to flashing See the through the strainer. The supplemental response stated that Response to overpressure is credited, but the amount of overpressure required 3.f.14. was not provided, nor was the available margin. The total head loss (without chemicals) is about 8 ft with a submergence of about 3 inches (LBLOCA). Please provide the minimum margin to flashing across the strainer throughout the strainer mission time. Please provide the assumptions used to determine this value. ES-181

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) RAI Complete RAls as Provided in NL-08-1829 (Reference 87) Response No. 17 The supplemental response stated that the vortex testing was See the conducted at a submergence that may have been slightly greater Response to (non-conservative) than that expected under LBLOCA conditions 3.f.3. (3.4 in vs. 3.675 +/- 0.5 in). Another section of the supplemental response stated that the testing was conducted with a representative or conservatively lowered water level, but no other details were provided. No vortex testing appears to have been conducted for SBLOCA conditions. No details on the test flow I I rates for vortex evaluations were provided. Vortex testing should be conducted with the minimum potential submergence and the maximum potential scaled flow rate through the strainer. Please provide an evaluation of vortex formation for the minimum level at which the strainer is required to operate (likely an SBLOCA condition). Please verify that the flow rates used in the vortex testing were conservative including the potential for higher flow rates in some sections of the strainer (generally those hydraulically closer to the pump suction). Please verify that the level that was tested for the LBLOCA case was in fact conservatively low. Please provide the submergence value for LBLOCA testing. Alternatively, please provide an updated evaluation considering all of these considerations. 18 The clean strainer head loss (CSHL) calculation methodology was See the not provided. It was not clear how the CSHL was divided between Response to strainer module head loss and piping head loss. Please provide 3.f.9. the methodology used to determine CSHL. Please provide information indicating that each section of the strainer, plenum, or piping was included in the calculation, the head loss value for each section, and the method used to determine the head loss for each strainer section. Please include any assumptions made for each portion of the calculation. 19 The supplemental response stated that for the sector tests debris Sector test is no was maintained in suspension using stirring. No information was longer utilized. provided to show that the stirring did not drive non-prototypical debris onto the bed nor prevent debris from collecting naturally on See the the strainer during these tests. For the module tests, from the Response to provided diagrams it appeared that the stirrer was far enough from 3.f.4. the strainer to prevent non-prototypical bed formation. Please provide information that justifies that the debris beds were not disrupted by the stirring and that the stirring did not result in non-prototypical debris accumulation on the strainer (accumulation of larger sizes of debris than would be expected). ES-182 J

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) RAI Complete RAls as Provided in NL-08-1829 (Reference 87) Response No. 20 During module tests stirring was used outside the area of the See the strainer. The supplemental response stated that the flow in the Response to test flume conservatively represented the plant flow patterns in the 3.f.4. area of the strainer to ensure that non-prototypical settling would not occur. No details were provided on how the plant and flume flow rates were modeled to assure that flow and turbulence would be prototypical. In general flow patterns in the plant are affected by, for example, upstream conditions, drainage into the sump pool, flow rates from various locations, upstream obstructions, and obstructions near the strainer. The boundary conditions in the models for determining typical plant flow patterns should be prototypical or conservative. Please provide information that justifies that the flow rates and patterns in the test flume for the module tests were prototypical or conservative with respect to plant conditions. 21 Please provide the basis for the statement in the supplemental RHR and CS response that the debris head loss for the RHR strainers bounds strainer head the head loss for the containment spray strainers. losses are evaluated independently. See the Response to 3.f.10. 22 Because of the large volume of debris and the relatively low See the submergence of the strainer it is possible for debris to collect on Response to top of the strainer and provide a pathway for air ingestion. This 3.f.3. was not discussed in the supplemental response. Air ingestion could result from a damming effect, or, if head loss exceeds submergence and holes form in the debris bed, these holes could allow air to be ingested through the debris bed. Please provide an evaluation of the potential for debris to collect on top of the strainer and provide a pathway for air ingestion into the strainer. 23 It was unclear how varying the debris loading affected the results See the in all of the head loss testing. Please provide the debris amounts Response to added to each test, the resulting theoretical bed thicknesses, and 3.f.4, 3.f.6, 3.f.7, the maximum head loss determined for each test. and 3.f.10. E5-183 I ___________J

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) RAI Complete RAls as Provided in NL-08-1829 (Reference 87) Response No. 24 Please provide the containment sump/pool level both soon after See the the realignment to containment spray recirculation as well as at the Response to 3.f.2 time post-accident when all assumed water contributions and and 3.g.1. diversions/hold-ups have completely taken effect (except for subsequent sump/pool water thermal contraction). See the The differences in the assumptions and results for these two cases Response to should be clearly explained, as should the times when the short- 3.g.2 and 3.g.9. term and long-term results are applicable. The strainer submergence should be provided for both cases. 25 The second portion of item 3.k of the revised content guide for the See the GL2004-02 supplemental responses requests that the licensee Response to "summarize the structural qualification results and design margins 3.k.2. for the various components of the sump strainer structural assembly." Please provide the actual and allowable stresses and show the design margins for the 16 bolt locations of the strainer base frame (in addition to the reaction forces already provided in Table 3.k.2-8 of the supplemental response). 26 Item 3.k.3 of the revised content guide for the GL2004-02 See the supplemental responses requests that the licensee "summarize Response to the evaluations performed for dynamic effects such as pipe whip, 3.k.3. jet impingement, and missile impacts associated with high-energy line breaks (as applicable)." In addition to the information provided in your September 2005 and February 2008 responses, please submit a detailed summary along with any additional supporting information regarding your assessment that the strainers are not subject to the aforementioned dynamic effects. 27 Please provide additional basis for concluding that the refueling See the cavity drains would not become blocked with debris. Please Response to 3.1.4. identify the potential types and characteristics of debris that could reach these drains. In particular, could large pieces of debris be blown into the upper containment by pipe breaks occurring in the lower containment, and subsequently fall into the cavity? In the case that partial/total blockage of the drains might occur, what would be the impact to minimum sump water level and ECCS and CS pump NPSH? Are there any potential flow restrictions in the two 12-inch refueling cavity drain lines (e.g., valves, meshing or gratings), and if so, how are these potential restrictions addressed so as ensure that these lines are not blocked during a LOCA? ES-184 __J

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) RAI Complete RAls as Provided in NL-08-1829 (Reference 87) Response No. 28 The NRC staff considers in-vessel downstream effects to not be See the fully addressed at Vogtle Units 1 and 2, as well as at other PWRs. Response to The supplemental response refers to draft WCAP-16793-NP, 3.n.1.

         "Evaluation of Long-Term Cooling Considering Particulate, Fibrous, and Chemical Debris in the Recirculating Fluid." TheNRC staff has not issued a final safety evaluation (SE) for WCAP-16793-NP. The licensee may demonstrate that in-vessel downstream effects issues are resolved for Vogtle Units 1 and 2 by showing that the Vogtle plant conditions are bounded by the final WCAP-16793-NP and the corresponding final NRC staff SE, and by addressing the conditions and limitations in the final SE. The licensee may alternatively resolve this item by demonstrating, without reference to WCAP-16793-NP or the staff SE, that in-vessel downstream effects have been addressed at Vogtle Units 1 and 2. In any event, the licensee should report how it has addressed the in-vessel downstream effects issue within 90 days of issuance of the final NRC staff SE on WCAP-16793-NP. The NRC staff is developing a Regulatory Issue Summary to inform the industry of the staffs expectations and plans regarding resolution of this remaining aspect of GSl-191.

29 The NRC Staff understands that SNC has changed its test See the approach to evaluate chemical effects. Please submit the revised Response to chemical effects test results and analyses to the NRC when they 3.o.1. become available. 5.0

References:

1. NRC Generic Letter (GL) 2004-02, "Potential Impact of Debris Blockage on Emergency Recirculation during Design Basis Accidents for Pressurized-Water Reactors," dated September 13, 2004

2. NEI Guidance Report NEI 04-07, Revision 0, "Pressurized Water Reactor Sump Performance Evaluation Methodology 'Volume 1 - Pressurized Water Reactor Sump Performance Evaluation Methodology'," December 2004

3. NEI Guidance Report NEI 04-07-, Revision 0, "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'," December 2004

4. Not Used

5. NRC Bulletin 2003-01, "Requests for Additional Information, Bulletin 2003-01, Potential Impact of Debris Blockage on Emergency Sump Recirculation at Pressurized Water Reactors" for VEGP Electric Generating Plant, Units 1 and 2, Docket Nos. 50-424 and 50-425

6. 60 Day Response to NRC Bulletin 2003-01 SNC-to-NRC NL-03-1514 dated 8/07/2003 ES-185

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Combined SNC response for Joseph M. Farley Nuclear Plant (FNP) and Vogtle Electric Generating Plant (VEGP) as required by NRC Bulletin 2003-01, "Potential Impact of Debris Blockage on Emergency Sump Recirculation at Pressurized Water Reactors" (ML032240030)

7. Response to a Request for Additional Information on NRC Bulletin 2003-01 NRC-to-SNC (NL-04-2013) dated 10/29/2004 Combined SNC response for Joseph M. Farley Nuclear Plant (FNP) and Vogtle Electric Generating Plant (VEGP) as required by NRC Bulletin 2003-01, "Potential Impact of Debris Blockage on Emergency Sump Recirculation at Pressurized Water Reactors"

8. Revised Response to a Request for Additional Information on NRC Bulletin 2003-01 NRC-to-SNC (NL-05-1207) dated 7/22/2005 Combined SNC response for Joseph M. Farley Nuclear Plant (FNP) and Vogtle Electric Generating Plant (VEGP) as required by NRC Bulletin 2003-01, "Potential Impact of Debris Blockage on Emergency Sump Recirculation at Pressurized Water Reactors - Revision 1"

9. NRC-to-SNC (NL-05-1633) dated 8/26/2005 Vogtle Electric Generating Plant, Units 1 and 2 - Response to NRC Bulletin 2003-01, "Potential Impact of Debris Blockage on Emergency Sump Recirculation at Pressurized Water Reactors" (TAC Nos. MB9625 and MB9626)

10. 90 day response to GL 2004-02

• SNC-to-NRC NL-05-0290 dated 2/25/2005 (ML052430746)

Joseph M. Farley Nuclear Plant, Vogtle Electric Generating Plant, Response to NRC Generic Letter 2004-02, "Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors"

11. Response (VEGP and FNP) to GL 2004-02 SNC-to-NRC NL-05-1264 dated 8/31/2005 Combined SNC response for Joseph M. Farley Nuclear Plant (FNP) and Vogtle Electric Generating Plant (VEGP) as required by NRC Generic Letter 2004-02, "Potential Impact of Debris Blockage on Emergency Recirculation during Design Basis Accidents at Pressurized Water Reactors"

12. NRC Request for Additional Information NRC-to-SNC (NL-06-0279) dated 2/9/2006 Vogtle Electric Generating Plant, Units 1 And 2, Request For Additional Information Re: Response To Generic Letter 2004-02, "Potential Impact Of Debris Blockage On Emergency Recirculation During Design-Basis Accidents At Pressurized Water Reactors" (TAC Nos. MC4727 and MC4728)

13. NRC-to-SNC (NL-06-0753) dated 3/28/2006 (ML060870274)

Alternative Approach for Responding to the Nuclear Regulatory Commission Request for Additional Information Letter Re: Generic Letter 2004-02,

14. VEGP 1st extension request to complete CAs (Unit 1 downstream effects) for GL 2004-02 SNC-to-NRC (NL-06-1275) dated 6/22/06 (ML061730462)

Vogtle Electric Generating Plant - Units 1 and 2 Request for Extension for Completing Corrective Actions for Generic Letter 2004-02, "Potential Impact of ES-186

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors"

15. SNC-to-NRC (NL-06-1483) dated 7/28/2006 Response to NRC RAI (6/30/06 phone call )on SNC Request for Extension for Completing Corrective Actions for Generic Letter 2004-02, "Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors"

16. NRC-to-SNC (NL-06-2055) dated 9/7/2006 (ML062500269)

Vogtle Electric Generating Plant, Unit 1, Approval of Generic Letter 2004-02 Extension Request (SNC request dated 6/22/2006)

17. NRC-to-NEI (NL-06-2686) dated 11/14/2006 Nuclear Regulatory Communication Request for Additional Information to Pressurized Water Reactor Licensees Regarding Reponses to Generic Letter 2004-02

18. NRC-to-All Licenses (NL-07-0090) dated 1/4/2007 Alternative Approach for Responding to the NRC request for Additional Information Letter Regarding GL 2004-02

19. SNC-to-NRC (NL-07-1969) dated 12/7/2007 Vogtle Electric Generating Plant Units 1 and 2 Generic Letter 2004-02 Response Extension Request for completion of Chemical Effects testing and analysis, Downstream Effects analysis for Components - Systems, and Fuel -

Vessel

20. NRC-to-SNC (NL-07-2367) dated 12/19/2007 Vogtle Electric Generating Plant, Units 1 and 2 -Generic Letter 2004-02 "Potential Impact of Debris Blockage on Emergency Recirculation during Design Basis Accidents at Pressurized-Water Reactors," Extension Request Approval (to May 31, 2008)

21. WCAP-16406-P-A Revision 1.0, "Evaluation of Downstream Sump Debris Effects in Support of GSl-191" March 2008

22. WCAP-16793-NP-A Revision 2.0, "Evaluation of Long-Term Cooling Considering Particulate, Fibrous and Chemical Debris in the Recirculating Fluid," July 2013

23. WCAP-16568-P Revision 0.0, "Jet Impingement Testing to Determine the Zone of Influence (ZOI) for OBA-Qualified I Acceptable Coatings"

24. Not used

25. Regulatory Guide 1.82, Revision 3, "Water Sources for Long Term Recirculation Cooling Following a Loss of Coolant Accident," November 2003

26. NUREG/CR-0800, Revision 1, "U.S. Nuclear Regulatory Commission Standard Review Plan," Section 3.6.2, "Determination of Rupture Locations and Dynamic Effects Associated with the Postulated Rupture of Piping," July 1981

27. NUREG/CR-2791, "Methodology for Evaluation of Insulation Debris Effects, Containment Emergency Sump Performance Unresolved Safety Issue A-43,"

Issued September 1982

28. NUREG/CR-3616, Transport and Screen Blockage Characteristics of Reflective Metallic Insulation Materials," January 1984

29. NUREG/CR-6224, "Parametric Study of the Potential for BWR ECCS Strainer Blockage Due to LOCA Generated Debris, Final Report," Issued October 1995 ES-187

                                        . Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary)

30. NUREG/CR-6369, "Drywell Debris Transport Study, Final Report," Volume 1, Issued September 1999

31. NUREG/CR-6369, "Drywell Debris Transport Study: Experimental Work, Final Report," Volume 2, Issued September 1999

32. NUREG/CR-6369, "Drywell Debris Transport Study: Computational Work, Final Report," Volume 3, Issued September 1999

33. NUREG/CR-6762, Volume 1, "GSl-191 Technical Assessment: Parametric Evaluations for Pressurized Water Reactor Recirculation Sump Performance,"

Issued August 2002

34. NUREG/CR-6762, Volume 2, "GSl-191 Technical Assessment: Summary and Analysis of U.S. Pressurized Water Reactor Industry Survey Responses and Responses to GL 97-04," Issued August 2002

35. NUREG/CR-6762, Volume 3, "GSl-191 Technical Assessment: Development of Debris Generation Quantities in Support of the Parametric Evaluation," Issued August 2002

36. NUREG/CR-6762, Volume 4, "GSl-191 Technical Assessment: Development of Debris Transport Fractions in Support of the Parametric Evaluation," Issued August 2002

37. NUREG/CR-6772, "GSl-191: Separate Effects Characterization of Debris Transport in Water," Issued August 2002

38. NUREG/CR-6773, "GSl-191: Integrated Debris-Transport Tests in Water Using Simulated Containment Floor Geometries," Issued December 2002

39. NUREG/CR-6808, "Knowledge Base for the Effect of Debris on Pressurized Water Reactor Emergency Core Cooling Sump Performance," Issued February 2003

40. NUREG/CR-6916, "Hydraulic Transport of Coating Debris, A Subtask of GSl-191,"

Issued December 2006

41. NEI Document 02-01, Revision 1, "Condition Assessment Guidelines: Debris Sources Inside PWR Containments"

42. Westinghouse Technical Bulletin, TB-06-15, "Unqualified Service Level 1 Coatings on Equipment in Containment," Dated September 28, 2006

43. C.D.I. Report 96-06, Revision A, "Air Jet Impact Testing of Fibrous and Reflective Metallic Insulation," included in Volume 3 of General Electric Document NED0-32686-A, "Utility Resolution Guide for ECCS Suction Strainer Blockage"

44. STPNOC Letter NOC-AE-13003043 to NRC, "Supplement 1 to Revised STP Pilot Submittal and Requests for Exemptions and Licensing Amendment for a Risk-Informed Approach to Resolving Generic Safety Issue (GSl)-191,"

November 13, 2013 (ML13323A183)

45. STPNOC Letter NOC-AE-15003241 to NRC, "Supplement 2 to STP Pilot Submittal and Requests for Exemptions and License Amendment for a Risk-Informed Approach to Address Generic Safety Issue (GSl)-191 and Respond to Generic Letter (GL) 2004-02 (TAC NOS. MF2400-MF2409)," August 20, 2015 (ML15246A126)

46. NEI Document (ML120481057), Revision 1, "ZOI Fibrous Debris Preparation:

Processing, Storage and Handling," January 2012

47. NRC Bulletin 2003-01, "Potential Impact of Debris Blockage on Emergency Sump Recirculation at Pressurized Water Reactors," dated June 9, 2003 ES-188

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary)

48. NRCB 93-02, NRC Bulletin 93-02, "Debris Plugging of Emergency Core Cooling Suction Strainers," May 11, 1993

49. NRCB 96-03, NRC Bulletin 96-03, "Potential Plugging of Emergency Core Cooling Suction Strainers by Debris in Boiling-Water Reactors," May 6, 1996

50. NUREG/CR-1829 Volume I, "Estimating LOCA Frequencies Through the Elicitation Process," 2008

51. NUREG/CR-2982, Revision 1, "Buoyancy, Transport, and Head Loss of Fibrous Reactor Insulation," July 1983

52. NUREG/CR-5640, "Overview and Comparison of US Commercial Nuclear Power Plants," September 1990

53. NUREG/CR-6367, "Experimental Study of Head Loss and Filtration for LOCA Debris," February 1996

54. NUREG/CR-6808, "Knowledge Base for the Effect of Debris on Pressurized Water Reactor Emergency Core Cooling Sump Performance," November 2005

55. NUREG/CR-6874, GSl-191: Experimental Studies of Loss-of-Coolant-Accident-Generated Debris Accumulation and Head Loss with Emphasis on the Effects of Calcium Silicate Insulation, May 2005

56. NUREG/CR-6877, Characterization and Head-Loss Testing of Latent Debris from Pressurized-Water-Reactor Containment Buildings, July 2005

57. NUREG/CR-6917, Experimental Measurements of Pressure Drop across Sump Screen Debris Beds in Support of Generic Safety Issue 191, February 2007

58. NUREG/CR-6988, Final Report- Evaluation of Chemical Effects Phenomena in Post-LOCA Coolant, March 2009: Revision 0

59. NUREG/CR-7172, "Knowledge Base Report on Emergency Core Cooling Sump Performance in Operating Light Water Reactors," January 2014

60. NUREG/CR-0869, USI A-43 Regulatory Analysis, Revision 1: October 1985

61. NUREG/CR-0897, Technical Findings Related to Unresolved Safety Issue A-43, Revision 1: October 1985

62. Not used

63. NUREG/CR-1862, Development of a Pressure Drop Calculation Method for Debris-Covered Sump Screens in Support of Generic Safety Issue 191, February 2007

64. NUREG/CR-1918, "Phenomena Identification and Ranking Table Evaluation of Chemical Effects Associated with Generic Safety Issue 191," February 2009

65. PWROG, OG-07-419, Transmittal of LOCADM Software in Support of WCAP-16793-NP, "Evaluation of Long-Term Cooling Associated with Sump Debris Effects" (PA-SEE-0312), September 2007

66. PWROG, OG-07-534, Transmittal of Additional Guidance for Modeling Post-LOCA Core Deposition with LOCADM Document forWCAP-16793-NP (PA-SEE-0312),

December 2007

67. PWROG, OG-08-64, Transmittal of LTR-SEE-1-08-30, "Additional Guidance for LOCADM for Modification to Aluminum Release" for Westinghouse Topical Report WCAP-16793-NP, "Evaluation of Long-Term Cooling Considering Particulate, Fibrous and Chemical Debris in the Recirculating Fluid" (PA-SEE-0312),

February 2008 ES-189

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary)

68. PWROG, OG-10-253, PWROG Response to Request for Additional Information Regarding PWROG Topical Report WCAP-16793-NP, Revision 1, "Evaluation of Long-Term Cooling Considering Particulate, Fibrous and Chemical Debris in the Recirculating Fluid" (PA-SEE-0312), August 2010

69. Regulatory Guide 1.174, Revision 2, "An Approach for Using Probabilistic Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," May 2011

70. SRM-SECY-12-0093, "Closure Options for Generic Safety Issue - 191, Assessment of Debris Accumulation on PWR Sump Performance,"

December 14, 2012

71. SECY-83-472, Information Report from W.J. Dircks to the Commissioners, "Emergency Core Cooling System Analysis Methods," November 17, 1983

72. WOG-06-113, "Submittal ofWCAP-16530-NP, 'Evaluation of Post-Accident Chemical Effects in Containment Sump Fluids to Support GSl-191' for Formal Review," 3/27/2006

73. WCAP-16530-NP-A, "Evaluation of Post-Accident Chemical Effects in Containment Sump Fluids to Support GSl-191," March 2008

74. WCAP-16613-P, Vogtle Electric Generating Plant Measurement Uncertainty Recapture Power Uprate Program Engineering Report, Revision 2, June 2007

75. WCAP-16785-NP, Revision 0, "Evaluation of Additional Inputs to the WCAP-16530-NP Chemical Model," May 2007

76. Not Used

77. Not Used

78. BWR Owners Group, "Utility Resolution Guide for ECCS Suction Strainer Blockage," Volume 3, October 1998

79. PA-SEE-1090(ML14153A013), PWROG Presentation, "GSl-191 Comprehensive Analysis and Test Program Update," NRC Public Meeting: April 2014

80. Letter from William Ruland (NRC) to Anthony Pietrangelo (NEI) (ML080230112),

Revised Guidance for Review of Final Licensee Responses to Generic Letter 2004-02, "Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors," March 28, 2008

81. ML081550043, C. B. Bahn et al., "Technical Letter Report on Evaluation of Long-term Aluminum Solubility in Borated Water Following a LOCA," February 2008

82. NRC Letter to SNC (ML092370630), "Summary of August 13, 2009, Public Conference Call with Southern Nuclear Operating Company, Inc. (SNC), on the Request for Additional Information Pertaining to Generic Letter 2004-02 (TAC NOS. MC4727 and MC2728)," August 31, 2009

83. ML102280594, "Evaluation of Chemical Effects Phenomena Identification and Ranking Table Results," March 2011

84. ML121520429, Nuclear Regulatory Commission, Official Transcript of.

Proceedings, Advisory Committee on Reactor Safeguards Thermal Hydraulic Phenomena Subcommittee Open Session, May 9, 2012

85. SNC Letter NL-04-2321 to NRC, "Joseph M. Farley Nuclear Plant Response to a Request for Additional Information on NRC Bulletin 2003-01 'Potential Impact of Debris Blockage on Emergency Sump Recirculation at Pressurized Water Reactors'," November 30, 2004 ES-190

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary)

86. SNC Letter NL-07-2168 to NRC (ML080150161), "Vogtle Electric Generating Plant License Amendment Request to Revise Technical Specifications (TS) 3.3.2,

        'ESFAS Instrumentation,' and TS 3.5.4, 'Refueling Water Storage Tank (RWST)',"

January 2008

87. NRC Letter NL-08-1829 to SNC, "Vogtle Electric Generating Plant, Units 1 and 2 -

Generic Letter 2004-02, 'Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized Water Reactors,' Request for Additional Information (TAC NOS. MC4727, MC4728)," December 2, 2008 (ML083100142) I ' 88. Pressurized Water Reactor Owners Group (PWROG) Letter OG-07-408, Revision 0, "PWROG Responses to NRC Second Set of Requests for Clarification and Supplemental Information Regarding WCAP-16530," September 2007

89. WOG-06-107, "PWR Owners Group Letter to NRC Regarding Error Corrections to WCAP-16530-NP (PA-SEE-0275)," March 21, 2006

90. SECY-10-0113, "Closure Options for Generic Safety Issue - 191, Assessment of .

Debris Accumulation on PWR Sump Performance," December 23, 2010

91. NEI letter to NRC, "GSl-191 - Current Status and Recommended Actions for Closure," May 4, 2012(ML12142A316)

92. NEI letter to NRC, "GSl-191 - Revised Schedule for Licensee Submittal of Resolution Path," November 15, 2012 (ML12325A072)

93. "NRC Review of Generic Safety lssue-191 Nuclear Energy Institute revised Schedule for Licensee Submittal of Resolution Path," (ML12326A497),

November 21, 2012

94. "Final Safety Evaluation for Pressurized Water Reactor Owners Group Topical Report WCAP-16793-NP, Revision 2, 'Evaluation of Long-Term Cooling Considering Particulate Fibrous and Chemical Debris in the Recirculating Fluid' (TAC NO. ME1234)," April 8, 2013(ML13084A152 and ML13084A154)

95. SNC Letter NL-07-1777 to NRC (ML080640601), "Vogtle Electric Generating Plant Supplemental Response to NRC Generic Letter 2004-02," February 28, 2008

96. SNC Letter NL-08-0670 to NRC (ML081640617), "Vogtle Electric Generating Plant Supplemental Response to NRC Generic Letter 2004-02," May 21, 2008

97. SNC Letter NL-08-0818 to NRC (ML081430616), "Vogtle Electric Generating Plant Generic Letter 2004-02 Response Extension Request," May 22, 2008

98. SNC Letter NL-08-1155 to NRC (ML082170513), "Vogtle Electric Generating Plant Supplemental Response to NRC Generic Letter 2004-02," July 31, 2008

99. SNC Letter NL-08-1195 to NRC (ML082170306), "Vogtle Electric Generating Plant Generic Letter 2004-02 Response Extension Request," July 31, 2008 100. SNC Letter NL-08-1228 to NRC (ML082380890), "Vogtle Electric Generating Plant Supplemental Response to NRC Generic Letter 2004-02," August 22, 2008 101. NRC Letter NL-08-1497 to SNC (ML082560233), "Vogtle Electric Generating Plant, Units 1 and 2 - Request for Additional Information Regarding Generic Letter 2004-02, 'Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors'," September 17, 2008 102. SNC Letter NL-08-1583 to NRC (ML083150262), "Vogtle Electric Generating Plant Generic Letter 2004-02 Extension Request for 'Completion of Chemical Effects and Closeout of GL 2004-02'," November 7, 2008 ES-191

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) 103. SNC Letter NL-09-0159 to NRC (ML090420235), "Vogtle Electric Generating Plant i I Generic Letter Supplemental Response," February 10, 2009 104. SNC Letter NL-09-1839 to NRC (ML093240098), "Vogtle Electric Generating Plant Generic Letter 2004-02 Closeout Status," November 19, 2009 I' 105. SNC Letter NL-13-0953 to NRC (ML13137A130), "Vogtle Electric Generating Plant Generic Letter 2004-02 Closeout Status," May 16, 2013 106. NRC Evaluation Guide (ML080380214), "NRC Staff Review Guidance Regarding Generic Letter 2004-02 Closure in the Area of Plant-Specific Chemical Effect Evaluations," March 2008

• 107. Regulatory Guide 1.82, Revision 4, "Water Sources for Long Term Recirculation Cooling Following a Loss of Coolant Accident," March 2012 108. NRC Letter (ML073110278) "Revised Content Guide for Generic Letter 2004-02 Supplemental Responses," November 2007 109. STPNOC letter NOC-AE-16003401 to the NRC, "Supplement 3 to Revised STP Pilot Submittal and Requests for Exemptions and License Amendment for a Risk-Informed Approach to Address Generic Safety Issue (GSl)-191 and Respond to Generic Letter (GL) 2004-02," October 20, 2016 110. SNC Letter NL-13-2544(ML13351A409) "Vogtle Electric Generating Plant, Units 1 and 2 Response to Request for Additional Information Regarding Closure of Option 2 to Address In-Vessel Mitigative Measures for Potential In-Vessel Blockage," December 13, 2013 111. NRC Evaluation Guide (ML080230038), "NRC Staff Review Guidance Regarding Generic Letter 2004-02 Closure in the Area of Strainer Head Loss and Vortexing,"

March 2008 ES-192

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Vogtle Electric Generating Plant Supplemental Response to NRC Generic Letter 2004-02 Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Attachment 1 General Electric Hitachi (GEH) Proprietary Information with Affidavit E5:A1-1

Global Nuclear Fuel - Americas AFFIDAVIT I, Peter M. Yandow, state as follows: (1) I am the Vice President, NPP/Services Licensing, Regulatory Affairs, GE-Hitachi Nuclear Energy Americas, LLC (GEH), and have been delegated the function of reviewing the information described in paragraph (2) which is sought to be withheld, and have been authorized to apply for its withholding. (2) The information sought to be withheld is contained in Enclosure 1 of GEH's letter, 0006-7789-010, Jack Noonan (GEH) to Jim A. Wade (Southern Nuclear Company), entitled "GEH Proprietary Information in SNC Supplemental Response to NRC Generic Letter 2004-02," March 29, 2017. GEH proprietary information in Enclosure 1, which is entitled "Excerpt of SNC Supplemental Response to NRC Generic Letter 2004 GEH Proprietary Information - Class II (Internal), is identified by a dotted underline inside double square brackets. ((J.h.t~--~-~nt~n£~.j~Jm ..~~J!!DP.l~Y-~J] In each case, the superscript notation {3} refers to Paragraph (3) of this affidavit, which provides the basis for the proprietary determination. (3) In making this application for withholding of proprietary information of which it is the owner or licensee, GEH relies upon the exemption from disclosure set forth in the Freedom of Information Act ("FOIA"), 5 USC Sec. 552(b)(4), and the Trade Secrets Act, 18 USC Sec. 1905, and NRC regulations 10 CFR 9.17(a)(4), and 2.390(a)(4) for "trade secrets" (Exemption 4). The material for which exemption from disclosure is here sought also qualify under the narrower definition of "trade secret", within the meanings assigned to those terms for purposes of FOIA Exemption 4 in, respectively, Critical Mass Energy Project v. Nuclear Regulatory Commission, 975F2d871 (DC Cir. 1992), and Public Citizen Health Research Group v. FDA, 704F2d1280 (DC Cir. 1983). (4) Some examples of categories of information which fit into the definition of proprietary information are:

a. Information that discloses a process, method, or apparatus, including supporting data and analyses, where prevention of its use by GEH's competitors without license from GEH constitutes a competitive economic advantage over other companies;

b. Information which, if used by a competitor, would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing of a similar product;

c. Information which reveals aspects of past, present, or future GEH customer-funded development plans and programs, resulting in potential products to GEH;

d. Information which discloses patentable subject matter for which it may be desirable to obtain patent protection.

0006-7789-010 Affidavit Page 1 of 3 l

The information sought to be withheld is considered to be proprietary for the reasons set forth in paragraphs (4 )a. and (4)b. above. (5) To address 10 CPR 2.390 (b) (4), the information sought to be withheld is being submitted to NRC in confidence. The information is of a sort customarily held in confidence by GEH, and is in fact so held. The information sought to be withheld has, to the best of my knowledge and belief, consistently been held in confidence by GEH, no public disclosure has been made, and it is not available in public sources. All disclosures to third parties including any required transmittals to NRC, have been made, or must be made, pursuant to regulatory provisions or proprietary agreements which provide for maintenance of the information in confidence. Its initial designation as proprietary information, and the subsequent steps taken to prevent its unauthorized disclosure, are as set forth in paragraphs (6) and (7) following. (6) Initial approval of proprietary treatment of a document is made by the manager of the originating component, the person most likely to be acquainted with the value and sensitivity of the information in relation to industry knowledge, or subject to the terms under which it was licensed to GEH. (7) The procedure for approval of external release of such a document typically requires review by the staff manager, project manager, principal scientist or other equivalent authority, by the manager of the cognizant marketing function (or his delegate), and by the Legal Operation, for technical content, competitive effect, and determination of the accuracy of the proprietary designation. Disclosures outside GEH are limited to regulatory bodies, customers, and potential customers, and their agents, suppliers, and licensees, and others with a legitimate need for the information, and then only in accordance with appropriate regulatory provisions or proprietary agreements. (8) The information identified in paragraph (2) is classified as proprietary because it contains detailed results of analytical model and methods of emergency core cooling system and containment spray strainers in Boiling Water Reactors and Pressurized Water Reactors. The development and approval of these models and methods were achieved at a significant cost to GEH. The development of the evaluation process along with the interpretation and application of the analytical results is derived from the extensive experience database that constitutes a major GEH asset. (9) Public disclosure of the information sought to be withheld is likely to cause substantial harm to GEH's competitive position and foreclose or reduce the availability of profit-making opportunities. The information is part of GEH's comprehensive safety and technology base, and its commercial value extends beyond the original development cost. The value of the technology base goes beyond the extensive physical database and analytical methodology and includes development of the expertise to determine and apply 0006-7789-010 Affidavit Page 2 of 3

the appropriate evaluation process. In addition, the technology base includes the value derived from providing analyses done with NRC-approved methods. The research, development, engineering, analytical, and NRC review costs comprise a substantial investment of time and money by GEH. The precise value of the expertise to devise an evaluation process and apply the correct analytical methodology is difficult to quantify, but it clearly is substantial. GEH's competitive advantage will be lost if its competitors are able to use the results of the GEH experience to normalize or verify their own process or if they are able to claim an equivalent understanding by demonstrating that they can arrive at the same or similar conclusions. The value of this information to GEH would be lost if the information were disclosed to the public. Making such information available to competitors without their having been required to undertake a similar expenditure of resources would unfairly provide competitors with a windfall, and deprive GEH of the opportunity to exercise its competitive advantage to seek an adequate return on its large investment in developing and obtaining these very valuable analytical tools. I declare under penalty of perjury that the foregoing is true and correct. Executed on this 29th day of March 2017. Peter M. Yandow Vice President, NPP/Services Licensing GE-Hitachi Nuclear Energy Americas, LLC 3901 Castle Hayne Road, MIC A-65 Wilmington, NC 28401 Peter.Yandow@ge.com 0006-7789-010 Affidavit Page 3 of 3

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Vogtle Electric Generating Plant Supplemental Response to NRC Generic Letter 2004-02 Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Attachment 2 Westinghouse Electric Corporation (WEC) Proprietary Information with Affidavit E5:A2-1}}

Text

{{#Wiki_filter:A Southern Nuclear J. J. Hutto Regulatory Affairs Director

• 40 Inverness Center Parkway Post Office Box 1295 Birmingham, AL 35242 205 992 5872 tel 205 992 7601 fax jjhutto@southemco.com APR 2 1 2017 Enclosure 2 to this letter contains Proprietary Information to be withheld from public disclosure under 10 CFR 2.390. When separated from Enclosure 2, this transmittal document and the other Enclosures are decontrolled.

Docket Nos.: 50-424 NL-16-2002 50-425 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Vogtle Electric Generating Plant - Units 1 & 2 Supplemental Response to NRC Generic Letter 2004-02 Ladies and Gentlemen: The purpose of this report is to provide, for Nuclear Regulatory Commission (NRC) staff review and approval, the Southern Nuclear Operating Company (SNC) supplemental response for Vogtle Electric Generating Plant Units 1 & 2 (VEGP) to Generic Letter (GL) 2004-02, dated September 13, 2004, "Potential Impact of Debris Blockage on Emergency Recirculation during Design Basis Accidents at Pressurized-Water Reactors." This supplemental response supersedes previous responses and uses a risk informed approach to evaluate effects of debris. This report is organized as described below:

• Enclosure 1 provides a high-level summary of Enclosures 2 through 4, and is organized with the same layout as draft Regulatory Guide (RG) 1.229 Section C.
• Enclosure 2 provides a detailed description of the plant-specific conditions and models related to generic safety issue (GSl)-191 (including proprietary information). This enclosure is organized in accordance with the content guideline for GL 2004-02 responses. This enclosure also includes a response to each of the previous requests for additional information (RAls) that VEGP had received on earlier GL 2004-02 submittals.

Accordingly, the responses provided in this enclosure supersedes those provided in previous SNC responses. Additionally, this enclosure contains attachments with affidavits for withholding of proprietary information.

• Enclosure 3 provides a description of the risk quantification using the NARWHAL computer code and the VEGP probabilistic risk assessment (PRA) model. This enclosure is organized with the same layout as draft RG 1.229 Appendix A The enclosure explains how all the individual parts are combined to quantify risk. It also provides discussion on uncertainty quantification.

U.S. Nuclear Regulatory Commission NL-16-2002 Page 2

• Enclosure 4 provides a summary of defense-in-depth and safety margin. This enclosure shows that the health and safety of the public are not adversely affected by debris-related failures of the strainers, pumps, downstream components, or core.
• Enclosure 5 is a duplicate of Enclosure 2 with the proprietary information redacted.

The determination of in-vessel debris limits is necessary to support the final VEGP risk-informed resolution to GL 2004-02 (including a corresponding license amendment request). Please note that the methodology SNC intends to use to determine in-vessel debris limits is currently under NRC review. By letter dated February 14, 2017, the NRC stated that it would not be appropriate for the staff to accept for review a requested licensing action (RLA) that relied upon an unapproved methodology. However, in this letter, the NRC stated their support for an SNC technical report that does not rely on an unapproved methodology. The intent of this technical report is that it will be used to prepare an NRC staff evaluation to support a subsequent RLA submittal after the in-vessel debris limits methodology is approved. Accordingly, the purpose of this report is to receive NRC review and approval of the SNC supplemental GL 2004-02 response which uses a risk-informed methodology to evaluate debris effects, excepting in-vessel fiber limits, which will be provided with the RLA. To support a timely RLA (including a license amendment request pursuant to 10 CFR 50.90) that will resolve this safety issue, SNC requests NRC approval of this report by April 21, 2018. If you have any questions, please contact Ken McElroy at 205.992.7369. Mr. J. J. Hutto states he is Regulatory Affairs Director of Southern Nuclear Operating Company, is authorized to execute this oath on behalf of Southern Nuclear Operating Company and, to the best of his knowledge and belief, the facts set forth in this letter are true. Respectfully submitted, 9<fbc-:> J. J. Hutto Regulatory Affairs Director JJH/RMJ Sworn to and subscri ed before me this _d}_ day of A~i '2017. i(/V\. --=-----1..~~---=-------=~~----J'-- My commission expires: /!J - I?-;)__() ( 1

U. S. Nuclear Regulatory Commission NL-16-2002 Page 3

Enclosures:

1. Introduction and Overall Summary

2. Supplemental Response to NRC Generic Letter 2004-02 (Proprietary)

3. Risk Quantification

4. Defense-in-Depth and Safety Margin

5. Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) cc: Southern Nuclear Operating Company Mr. S. E. Kuczynski, Chairman, President & CEO Mr. D. G. Bost, Executive Vice President & Chief Nuclear Officer Mr. R. D. Gayheart, Fleet Operations General Manager Mr. M. D. Meier, Vice President - Regulatory Affairs Mr. B. K. Taber, Vice President - Vogtle 1 & 2 Mr. B. J. Adams, Vice President- Engineering Mr. D. D. Sutton, Regulatory Affairs Manager - Vogtle 1 & 2 RType: CVC?OOO U.S. Nuclear Regulatory Commission Ms. C. Haney, Regional Administrator Mr. M. D. Orenak, NRR Project Manager - Vogtle 1 & 2 Mr. M. F. Endress, Senior Resident Inspector - Vogtle 1 & 2 State of Georgia Mr. R. E. Dunn, Director- Environmental Protection Division

CAW-17-456'5 AFFIDAVIT COMMONWEALTH OF PENNSYLVANIA: SS COUNTY OF BUTLER: . I, James A: Gresham, am authorized to execute this Affidavit on behalf of Westinghouse Electric

• Co1Upany LLC ("Westingliouse~') ahd declare that the averments of fact set forth in this. Affida.vit are true*

and correct to the best of my knowledge, information, atid belief. EXecuted on: +/tr fe !

                                                      .. ~Jaines A. Gresham, Manager v Regulafory Compliance

3 CAW-17-4565 (1) I am Manager, Regulatory Compliance, Westinghouse Electric Company LLC ("Westinghouse"), and as such, I have been specifically delegated the function of reviewing the proprietary information sought to be withheld from public disclosure in connection with nuclear power plant licensing and rule making proceedings, and am authorized to apply for its withholding on behalf of Westinghouse. (2) I am making this Affidavit in conformance with the provisions of 10 CFR Section 2.390 of the Nuclear Regulatory Commission's ("Commission's") regulations and in coajunction with the Westinghouse Application for Withholding Proprietary Information from Public Disclosure accompanying this Affidavit. (3) I have personal knowledge of the criteria and procedures utilized by Westinghouse in designating information as a trade secret, privileged or as confidential commercial or financial information. (4) Pursuant to the provisions of paragraph (b)(4) of Section 2.390 of the Commission's regulations, the following is furnished for consideration by the Commission in determining whether the information sought to be withheld from public disclosure should be withheld. (i) The information sought to be withheld from public disclosure is owned and has been held in confidence by Westinghouse. (ii) The information is ofa type customarily held in confidence by Westinghouse and not customarily disclosed to the public. Westinghouse has a rational basis for determining the types of information customarily held in confidence by it and, in that connection, utilizes a system to determine when and whether to hold certain types of information in confidence. The application of that system and the substance of that system constitute Westinghouse policy and provide the rational basis required. Under that system, information is held in confidence if it falls in one or more of several types, the release of which might result in the loss of an existing or potential competitive advantage, as follows: (a) The information reveals the distinguishing aspects of a process (or component, structure, tool, method, etc.) where prevention of its use by any of

4 CAW-17-4565 Westinghouse's competitors without license from Westinghouse constitutes a competitive economic advantage over other companies. (b) It consists of supporting data, including test data, relative to a process (or component, structure, tool, method, etc.), the application of which data secures a competitive economic advantage, e.g., by optimization or improved marketability. (c) Its use by a competitor would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing a similar product. (d) It reveals cost or price information, production capacities, budget levels, or commercial strategies of Westinghouse, its customers or suppliers. (e) It reveals aspects of past, present, or future Westinghouse or customer funded development plans and programs of potential commercial value to Westinghouse. (f) It contains patentable ideas, for which patent protection may be desirable. (iii) There are sound policy reasons behind the Westinghouse system which include the following: (a) The use of such information by Westinghouse gives Westinghouse a competitive advantage over its competitors. It is, therefore, withheld from disclosure to protect the Westinghouse competitive position. (b) It is information that is marketable in many ways. The extent to which such information is available to competitors diminishes the Westinghouse ability to sell products and services involving the use of the information. (c) Use by our competitor would put Westinghouse at a competitive disadvantage by reducing his expenditure of resources at our expense.

5 CAW-17-4565 (d) Each component of proprietary information pertinent to a particular competitive advantage is potentially as valuable as the total competitive advantage. If competitors acquire components of proprietary information, any one component may be the key to the entire puzzle, thereby depriving Westinghouse of a competitive advantage. (e) Unrestricted disclosure would jeopardize the position of prominence of Westinghouse in the world market, and thereby give a market advantage to the competition of those countries. (f) The Westinghouse capacity to invest corporate assets in research and development depends upon the success in obtaining and maintaining a competitive advantage. (iv) The information is being transmitted to the Commission in confidence and, under the provisions of 10 CFR Section 2.390, is to be received in confidence by the Commission. (v) The information sought to be protected is not available in public sources or available information has not been previously employed in the same original manner or method to the best of our knowledge and belief. (vi) The proprietary information sought to be withheld in this submittal is that which is appropriately marked in "Vogtle Electric Generating Plant Supplemental Response to NRC Generic Letter 2004-02, Enclosure 2, 'Supplemental Response to NRC Generic Letter 2004-02' "(Proprietary), for submittal to the Commission, being transmitted by Letter GP-19572. The proprietary information as submitted by Westinghouse is that associated with resolution of and response to NRC Generic Letter 2004-02 and may be used only for that purpose. (a) This information is part of that which will enable Westinghouse to provide commercial support for resolution of and response to NRC Generic Letter 2004-02.

6 CAW-17-4565 (b) Further this information has substantial commercial value as follows: (i) Westinghouse plans to sell the use of similar information to its customers for the purpose of providing support for resolution of and response to NRC Generic Letter 2004-02. (ii) Westinghouse can sell support and defense of industry guidelines and acceptance criteria for plant-specific applications. (iii) The information requested to be withheld reveals the distinguishing aspects ofa methodology which was developed by Westinghouse. Public disclosure of this proprietary information is likely to cause substantial harm to the competitive position of Westinghouse because it would enhance the ability of competitors to provide similar technical evaluation justifications and licensing defense services for commercial power reactors without commensurate expenses. Also, public disclosure of the information would enable others to use the information to meet NRC requirements for licensing documentation without purchasing the right to use the information. The development of the technology described in part by the information is the result of applying the results of many years of experience in an intensive Westinghouse effort and the expenditure ofa considerable sum of money. In order for competitors of Westinghouse to duplicate this information, similar technical programs would have to be performed and a significant manpower effort, having the requisite talent and experience, would have to be expended. Further the deponent sayeth not.

Proprietary Information Notice Transmitted herewith are proprietary and non-proprietary versions of a document, furnished to the NRC associated with resolution of and response to NRC Generic Letter 2004-02 and may be used only for that purpose. In order to conform to the requirements of 10 CFR 2.390 of the Commission's regulations concerning the protection of proprietary information so submitted to the NRC, the information which is proprietary in the proprietary versions is contained within brackets, and where the proprietary information has been deleted in the non-proprietary versions, only the brackets remain (the information that was contained within the brackets in the proprietary versions having been deleted). The justification for claiming the information so designated as proprietary is indicated in both versions by means of lower case letters (a) through (f) located as a superscript immediately following the brackets enclosing each item of information being identified as proprietary or in the margin opposite such information. These lower case letters refer to the types of information Westinghouse customarily holds in confidence identified in Sections (4)(ii)(a) through (4)(ii)(f) of the Affidavit accompanying this transmittal pursuant to 10 CFR 2.390(b)(l). Copyright Notice The reports transmitted herewith each bear a Westinghouse copyright notice. The NRC is permitted to make the number of copies of the information contained in these reports which are necessary for its internal use in connection with generic and plant-specific reviews and approvals as well as the issuance, denial, amendment, transfer, renewal, modification, suspension, revocation, or violation of a license, permit, order, or regulation subject to the requirements of 10 CFR 2.390 regarding restrictions on public disclosure to the extent such information has been identified as proprietary by Westinghouse, copyright protection notwithstanding. With respect to the non-proprietary versions of these reports, the NRC is permitted to make the number of copies beyond those necessary for its internal use which are necessary in order to have one copy available for public viewing in the appropriate docket files in the public document room in Washington, DC and in local public document rooms as may be required by NRC regulations if the number of copies submitted is insufficient for this purpose. Copies made by the NRC must include the copyright notice in all instances and the proprietary notice ifthe original was identified as proprietary.

AL I 0 N 5(1f?lCE A~D TECHNOlOejiY ALION Science & Technology AFFIDAVIT We, Andy Roudenko, Project Manager and Martin Rozboril, Jr. Assistant Vice President Division Manager (AVPDM) state as follows: (1) We, Andy Roudenko, Project Manager, and Martin Rozboril, Jr. AVPDM, Nuclear Services, ALION Science & Technology ("Alion") and have been delegated the function of reviewing the information described in paragraph (2) which is sought to be withheld, and have been authorized to apply for its withholding. (2) The information sought to be withheld is contained in all revisions of ALION Science & Technology report "Erosion Testing of Small Pieces of Low Density Fiberglass Debris-Test Report," ALION-REP-ALION-1006-04, with the latest revision to date, Rev. I, dated November 17, 20011. Information from this report was used to support analysis of post-LOCA debris transport in work designed to address GSI-191, Assessment of Debris Accumulation on PWR Sump Performance, issues at Southern Nuclear Operating Company, Vogtle Units 1 and 2. Specifically, the following Sections and Figures are to be withheld, on that basis that these unique attribute of the testing approach, test results and conclusions:

• Background
• Figure 1.1. I
• Figure 2.1.2
• Figure 2.1.3
• Figure 2.1.5
• Figure 2.1.6
• Figure 2.1.9
• Test Results, including Figures and Tables
• Data Analysis, including Figures and Tables
• Conclusions
• Appendices (3) In making this application for withholding of proprietary information of which it is the owner, Alion relies upon the exemption from disclosure set forth in the Freedom of Information Act ("FOIA"), 5 USC Sec. 552(b)(4), and the Trade Secrets Act, 18 USC Sec. 1905, and NRC regulations 10 CFR 9.l 7(a)(4), and 2.390(a)(4) for "trade secrets" (Exemption 4). The material for which exemption from disclosure is here sought also qualify under the narrower definition of "trade secret", within the meanings assigned to those terms for purposes of FOIA Exemption 4 in, respectively, Critical Mass Energy Project v. Nuclear Regulatory Commission, 975F2d871 (DC Cir. 1992), and Public Citizen Health Research Group v. FDA, 704F2dl280 (DC Cir. 1983).

Page I of3 MAS Affidavit

                                        .ALION SCIEMC.E A.ND TECHNOLOGY (4) Some examples of categories of information which fit into the definition of proprietary information are:

a. Information that discloses a process, method, or apparatus, including supporting data and analyses, where prevention of its use by Alion's competitors without license from Alion constitutes a competitive economic advantage over other companies

b. Information which, if used by a competitor, would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing of a similar product;

c. Information which reveals aspects of past, present, or future Alion customer-funded development plans and programs, resulting in potential products to Alion;

d. Information which discloses patentable subject matter for which it may be desirable to obtain patent protection.

The information sought to be withheld is considered to be proprietary for the reasons set forth in paragraphs (4) a, and (4) b, above. (5) To address 10 CFR 2.390 (b) (4), the information sought to be withheld is being submitted to NRC in confidence. The information is of a sort customarily held in confidence by Alion, and is in fact so held. The information sought to be withheld has, to the best of my knowledge and belief, consistently been held in confidence by Alion, no public disclosure has been made, and it is not available in public sources. All disclosures to third parties including any required transmittals to NRC, have been made, or must be made, pursuant to regulatory provisions or proprietary agreements which provide for maintenance of the information in confidence. Its initial designation as proprietary information, and the subsequent steps taken to prevent its unauthorized disclosure, are as set forth in paragraphs (6) and (7) following. (6) Initial approval of proprietary treatment of a document is made by the manager of the originating component, the person most likely to be acquainted with the value and sensitivity of the information in relation to industry knowledge. Access to such documents within Alion is limited on a "need to know" basis. (7) The procedure for approval of external release of such a document typically requires review by the staff manager, project manager, principal scientist or other equivalent authority, by the manager of the cognizant marketing function (or their delegate), and by the Legal Operation, for technical content, competitive effect, and determination of the accuracy of the proprietary designation. Disclosures outside Alion are limited to regulatory bodies, customers, and potential customers, and their agents, suppliers, and licensees, and others with a legitimate need for the information, and then only in accordance with appropriate regulatory provisions or proprietary agreements. (8) The document identified in paragraph (2), above, is classified as proprietary because it contains "know-how" and "unique data" developed by Alion within our research and Page 2 of 3 MAS Affidavit

ft.LION SCIENCE AND TECHNOLOGY development programs. The development of this document, supporting methods and data constitutes a major Al ion asset in this current market. (9) Public disclosure of the information sought to be withheld is likely to cause substantial harm to Alion's competitive position and foreclose or reduce the availability of profit-making opportunities. The information is part of Alion's comprehensive BWR/PWR GSI-191 analysis base, and its commercial value extends beyond the original development cost. The value of the technology base goes beyond the extensive physical database and experimental methodology and includes development of the expertise to determine and apply the appropriate evaluation process. The research, development, engineering, analytical and experimental costs comprise a substantial investment of time and money by Alion. The precise value of the expertise to devise an evaluation process and apply the correct analytical methodology is difficult to quantify, but it clearly is substantial. Alion's competitive advantage will be lost if its competitors are able to use the results of the Alion experience to normalize or verify their own process or if they are able to claim an equivalent understanding by demonstrating that they can arrive at the same or similar conclusions. The value of this information to Alion would be lost if the information were disclosed to the public. Making such information available to competitors without their having been required to undertake a similar expenditure of resources would unfairly provide competitors with a windfall, and deprive Alion of the opportunity to exercise its competitive advantage to seek an adequate return on its large investment in developing these very valuable analytical tools. I declare under penalty of perjury that the foregoing affidavit and the matters stated therein are true and correct to the best of my knowledge, information, and belief. Executed on this 20th day of April 2017. Martin r Digitally signed by Andy Roudenko

                                   ~! DN: cn=Andy Roudenko, o=Alion Rozboril, Jr.

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                                 .*~\Science and Technology, ou=Nudear
                                 ,1 ':services Division,
                                      ~~ail=aro~denko@alionscience.com, 2017.04.20

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                       , *,/          Date:2017.04.2012:09:50-07'00' 15:06:17 -06'00'

( .* Andy Roudenko Martin Rozboril, Jr. Project Manager Assistant Vice President ALION Science & Technology Division Manager, Nuclear Services ALION Science & Technology Page 3 of3 MAS Affidavit

Vogtle Electric Generating Plant - Units 1 & 2 Supplemental Response to NRC Generic Letter 2004-02 Enclosure 1 Introduction and Overall Summary

Enclosure 1 Introduction and Overall Summary Table of Contents 1.0 Introduction 2.0 Systematic Risk Assessment of Debris 3.0 Initiating Event Frequencies 4.0 Defense-in-Depth and Safety Margin 5.0 Uncertainty 6.0 Monitoring Program 7.0 Quality Assurance 8.0 Periodic Update of Risk-Informed Analysis 9.0 Reporting and Corrective Actions 10.0 License Application 11.0 References Attachments E1:A1 Resolution of the VEGP Internal Events PRA Peer Review Findings E1 :A2 Comparison of RG 1.200 Revision 1 and Revision 2 Supporting Requirements E1 :A3 Resolution of the VEGP Seismic PRA Peer Review Findings E1-1

Enclosure 1 Introduction and Overall Summary 1.0 Introduction In 2010, due to the ongoing challenges of resolving Generic Safety Issue (GSl)-191, the United States Nuclear Regulatory Commission (NRG) commissioners issued a staff requirements memorandum (SRM) directing the NRG staff to consider new and innovative resolution approaches (Reference 1). One of the approaches included in the SRM was the option of addressing GSl-191 using a risk-informed approach. In 2011, South Texas Project (STP) initiated a multi-year effort as a pilot plant to define and implement a risk-informed approach to address the concerns associated with GSl-191. In 2012, the NRG staff issued SRM-SECY-12-0093 (Reference 2) providing recommendations for closure options, and these options were accepted by the NRG commissioners. In 2013, Southern Nuclear Operating Company (SNC) selected Option 2b (full risk-informed resolution path) for closure of NRG Generic Letter (GL) 2004-02 (Reference 24) at Vogtle Electric Generating Plant (VEGP) Units 1 and 2 (Reference 3). The objective of GSl-191 is to ensure that post-accident debris blockage will not impede or prevent the operation of the emergency core cooling system (ECCS) or containment spray system (CSS) in recirculation mode at pressurized water reactors (PWRs) during loss of coolant accidents (LOCAs) or other high energy line break (HELB) accidents for which recirculation is required (Reference 4). SNC has provided multiple responses to the NRG supporting the resolution of GSl-191. An in-depth history of the VEGP correspondences issued by or submitted to the NRG on the subject of GSl-191 is provided in Sections 1.0 and 2.0 of Enclosure 2, documenting VEGP's compliance with regulatory requirements. This submittal provides a complete summary of the risk-informed GSl-191 evaluation performed for VEGP Units 1 and 2, superseding all previous GL 2004-02 responses. The VEGP_ GSl-191 submittal is organized as described below:

• Enclosure 1 provides a high-level summary of the other enclosures, and is organized with the same layout as draft Regulatory Guide (RG) 1.229 Section C (Reference 4).
• Enclosure 2 provides a detailed description of the plant-specific conditions and models related to GSl-191 (including some proprietary information). This enclosure is organized in accordance with the content guideline for GL 2004-02 responses (Reference 5). This enclosure also includes a response to each of the previous requests for additional information (RAls) that VEGP had received on earlier GL 2004-02 submittals. Additionally, this enclosure contains attachments with affidavits for withholding the proprietary information.
• Enclosure 3 provides a description of the risk quantification using NARWHAL and the VEGP probabilistic risk assessment (PRA) model. This enclosure is organized with the same layout as draft RG 1.229 Appendix A (Reference 4). The enclosure explains how all of the individual parts are combined to quantify risk. It also provides discussion on uncertainty quantification.

E1-2

Enclosure 1 Introduction and Overall Summary

• Enclosure 4 provides a summary of defense-in-depth and safety margin. This enclosure shows that the health and safety of the public are not adversely affected by debris-related failures of the strainers, pumps, downstream components, or core.
• Enclosure 5 is a duplicate of Enclosure 2 with the proprietary information redacted.

The overall evaluation for VEGP is based heavily on models that have been used in the past and accepted by the NRC for GSl-191 resolution. The results of this evaluation show with high confidence that the risk associated with GSl-191 is very low, as defined by RG 1.174 Region Ill (Reference 6). In addition, the analysis includes significant safety margin and does not affect any of the existing defense-in-depth measures that are in place to protect the public. 2.0 Systematic Risk Assessment of Debris As described in RG 1.174 (Reference 6), the systematic risk assessment should consider all hazards, initiating events, and plant operating modes. However, a screening process can be used to eliminate scenarios that are not relevant, not affected by debris, or have an insignificant contribution. The specific GSl-191 failure modes that were considered are:

1. Debris accumulation in an upstream flow path choke point (e.g., a refueling canal drain) exceeds blockage limits and reduces the available sump volume.

2. Strainer head loss exceeds the net positive suction head (NPSH) margin for the ECCS and CSS pumps when the strainer is fully submerged.

3. Strainer head loss exceeds half of the submerged strainer height when the strainer is partially submerged.

4. Strainer head loss exceeds the strainer structural margin.

5. Gas voids (i.e., water vapor due to flashing or air intrusion due to degasification or vortexing) downstream of the strainers exceed the acceptable void fraction limits of the ECCS and CSS pumps.

6. Debris penetration exceeds ex-vessel downstream effects limits for component wear or clogging.

7. Debris penetration exceeds in-vessel downstream effects limits for core blockage.

8. Buildup of oxides and other chemical precipitates on fuel cladding exceed heat transfer limits.

9. Boric acid concentration in the core exceeds the solubility limit resulting in boric acid precipitation.

Failure Modes 1, 6, and 8, as well as the vortexing portion of Failure Mode 5, have been addressed in a bounding manner for the range of possible breaks with no issues of concern (see Enclosure 2-Section 3.1 for upstream blockage, Section 3.f.3 for vortexing, Section 3.m for ex-vessel downstream effects, and Section 3.n.1 for the LOCA deposition model (LOCADM) portion of the analysis of in-vessel effects) and were therefore not explicitly modeled in NARWHAL (a software tool that analyzes the E1-3

Enclosure 1 Introduction and Overall Summary GSl-191 phenomenological models in a self-consistent, time-dependent manner). The remaining failure modes were explicitly modeled. Note that core blockage (Failure Mode 7) and boric acid precipitation (Failure Mode 9) were addressed by using assumed debris limits. Figure 1-1 shows the relationship between the various elements of the risk-informed GSl-191 analysis and documentation. Sump Volume (Enc 2) Various Bounding Analyses (Enc 2) Determine min/max post- Fiber Penetration

• Ex-vessel wear and blockage accident sump volume Testing
• Fuel cladding debris deposition (Enc 2) (2014)
• Vortexing for bounding conditions Debris Transport Debris (Enc 2)

Generation Quantify debris Head Loss Testing (Enc 2) transported to (Enc 2) (2009) Quantify debris strainer for each Risk Assessment generated by break (Flow-3D) (Enc 3) breaks at all Calculate delta core weld locations Chemical Debris damage frequency on primary (Enc 2) Evaluation of and delta large early piping Quantify chemical each break release frequency scenario against (Enc 2 (BADGER) precipitate debris all GSl-191 (Enc 3 -~ Break passes ~ (CAFTA) for each break Yes failure criteria (NARWHAL) CAD Model Develop defense in (Enc 2) No depth and mitigative CAD model of strategies (Enc 4) Vogtle Break fails containment (inventor) Prepare LAR with exemptions and Tech 1 4 ' - - - - - - ' Prepare Licensing DOC lo<'----* Develop GL 2004-02 Spec changes (Future (DOC SNCS45368) Submittal (Enc 2) Submittal) Figure 1 Flow chart illustrating analysis and documentation elements 2.1 Hazards, Initiating Events, and Plant Operating Modes The only scenarios that need to be considered for GSl-191 are scenarios that require recirculation through the ECCS and/or CSS strainers. If recirculation is not required, there is no potential for debris-related failures of the strainers, pumps, downstream components, or core. The hazards and initiating events relevant to GSl-191 at VEGP include:

1. Reactor coolant system (RCS) pipe breaks resulting in small, medium, and large LOCAs

2. Non-piping LOCAs

3. Secondary side breaks inside containment (SSBI) that result in a consequential LOCA upon failure to terminate safety injection or a stuck open power-operated relief valve (PORV)

4. Seismically-induced LOCAs

5. Water hammer-induced LOCAs E1-4 L

Enclosure 1 Introduction and Overall Summary These hazards and initiating events are discussed in more detail in Section 3.0. The quantitative risk assessment was performed for LOCAs and SSBls that occur during full power operation (i.e., Mode 1), which is assumed to be equivalent or bounding compared to the other operating modes. This is a reasonable assumption because the RCS pressure and temperature (key inputs affecting the ZOI size) would either be approximately the same or significantly lower for Modes 2 through 6. In addition, the flow rate required to cool the core (a key input affecting core blockage) would be significantly reduced for low power or shutdown modes. 2.2 Risk Attributable to Debris The risk attributable to debris was quantified in terms of the change in core damage frequency (b.CDF) and the change in large early release frequency (b.LERF) compared to a hypothetical plant condition without any debris. This was done using a conservative approach that results in mean b.CDF and b.LERF values that are skewed high (as opposed to a best-estimate approach that would result in a more accurate prediction of the mean b.CDF and b.LERF values, or a bounding approach that would significantly over-predict the mean b.CDF and b.LERF values). The risk quantification was performed using the NARWHAL software (to calculate the conditional failure probabilities (CFPs) associated with the effects of debris) and the VEGP internal events PRA model (with some modifications to represent the GSl-191 failure events accurately). The PRA model of record is referred to as the "base PRA model", and the modified PRA model is referred to as the "GSl-191 PRA model". The base PRA model has been peer reviewed against RG 1.200 (Reference 7) and is therefore appropriate to use for risk-informed applications. In order to support the detailed quantification of the GSl-191 risk impact, the base PRA model was modified to incorporate events for GSl-191 sump strainer and core blockage failures, along with the LOCA initiating *events and equipment configurations associated with each potential GSl-191 failure. The risk evaluation relies on many engineering calculations and tests that have been developed and conducted for VEGP over the last several years to address GSl-191 and GL 2004-02. These calculations and tests are described in detail in Enclosure 2. The GSl-191 risk quantification for VEGP shows that the overall risk associated with debris (GDF, LERF, b.CDF, and b.LERF) is very low as defined by Region Ill of RG 1.174 (Reference 6). Figure 1-2 and Figure 1-3 show the RG 1.174 risk guidelines. E1-5

Enclosure 1 Introduction and Overall Summary t LL 0 (.)

               <I 10*5 Region II 10* 5    - - - - - - -- --   - -"---'

Region Ill CDF --+ Figure 1 RG 1.174 Risk Acceptance Guidelines for CDF and ACDF t LL a: w

                 <1 10"6 Region II Region Ill LEAF-+-

Figure 1 RG 1.174 Risk Acceptance Guidelines for LERF and ALERF As shown in Table 1-3 the total baseline risk (from internal events , internal fire , internal flood , and seismic events) for the VEGP Unit 1 PRA model is 4.39x10-5 per reactor-year (ry-1 ) for CDF and 1.73x1 Q-6 ry-1 for LERF. The total baseline risk for the VEGP Unit 2 PRA model is 5.05x10-5 ry-1 for CDF and 1.90x1Q-6 ry- 1 for LERF . The change in risk calculated using the VEGP GSl-191 PRA model is shown in Table 1-1. Note that the internal events and therefore the GSl-191 PRA models are identical for Units 1 and 2. E1 -6

Enclosure 1 Introduction and Overall Summary Table 1 VEGP Total Risk Impact due to GSl-191 Failures

                                                               .dCDF         .dLERF Case                              {ry-1)         {ry-1)

Risk increase from GSl-191 failures for high-likelihood 2.32x10-8 3.10x10-11 LOCA confi!:1urations Bounding risk increase from GSl-191 failures for 1.41x10-9 4.09x10-12 unlikely LOCA confiQurations Risk increase from GSl-191 failures for seismically- 1.50x10-10 1.50x10-9 induced LOCAs Risk increase from GSl-191 failures for SSBls 1.39x10-9 8.25x10- 11 Total risk increase associated with GSl-191 2.75x10-s 2.68x10-10 Enclosures 2 and 3 provide a detailed description of the GSl-191 models and risk evaluation that were used to calculate these LlCDF and LlLERF values. 2.3 Technical Adequacy of Probabilistic Risk Assessment Results The systematic risk assessment of debris for the resolution of GSl-191 at VEGP uses the applicable VEGP PRA models. This section provides information on the technical adequacy of the VEGP Internal Events (including internal flooding) and Seismic PRA model results in support of the resolution of GSl-191. The guidance provided in RG 1.200 (Reference 7, Section 4.2) indicates that the following items be addressed in documentation submitted to the NRC to demonstrate the technical adequacy of the PRA:

1. Identification of permanent plant changes (such as design or operational practices) that have an impact on the PRA but have not been incorporated in the PRA.

2. The parts of the PRA used to produce the results are performed consistently with the PRA Standard as endorsed by RG 1.200.

3. A summary of the risk assessment methodology used to assess the risk of the application, including how the PRA model was modified to appropriately model the risk impact of the application.

4. Identifications of key assumptions and approximations in the PRA relevant to the results used in the decision making process.

5. A discussion of the resolution of peer review or self-assessment findings and observations that are applicable to the parts of the PRA required for the application.

6. Identification of parts of the PRA used in the analysis that were assessed to have capability categories less than that required for the application.

This section provides the information to address these items. E1-7

Enclosure 1 Introduction and Overall Summary The VEGP PRA models are highly detailed, including a wide variety of initiating events, modeled systems, operator actions, and common cause events. The quantification process used for the VEGP PRA models is based on the event tree I fault tree methodology, which is a well-known methodology in the industry. The VEGP PRA models are controlled in accordance with the SNC procedure for PRA generation, maintenance and updates. The procedure defines the process for implementing regularly scheduled and interim PRA model updates, for tracking issues identified as potentially affecting the PRA models (e.g., due to changes in the plant, errors or limitations identified in the model, industry operating experience, etc.), and for controlling the model and associated computer files. To ensure that the current PRA model remains an accurate reflection of the as-built, as-operated plants, the PRA maintenance procedure requires the following activities be routinely performed:

• Design changes and procedure changes are reviewed for their impact on the PRA model on an on-going basis.
• Reliability data, unavailability data, initiating events frequency data, human reliability data, and other such PRA inputs shall be reviewed approximately every two fuel cycles and updated as necessary to maintain the PRA consistent with the as-operated plant.

As demonstrated by the information presented in the following sections, the VEGP Units 1 and 2 Internal Events and Seismic PRA models are technically adequate for the systematic risk assessment of debris for the resolution of GSl-191. 2.3.1 Plant Changes Not Yet Incorporated into the PRA Model As part of PRA model configuration control, SNC maintains a PRA model maintenance database that tracks all issues that have been identified that could impact the VEGP PRA model. Per the SNC procedure for PRA maintenance, the significance of the pending items in the database is evaluated to determine the impact on model results. Each pending item is prioritized for future model updates per its significance to model results. Based on a review of the VEGP PRA maintenance log, there are no significant outstanding changes that would impact the GSl-191 risk assessment. A summary of plant changes implemented since the cutoff date of the VEGP Seismic PRA and a qualitative assessment of the likely impact of those changes is provided in Table 1-2. E1-8

Enclosure 1 Introduction and Overall Summary Table 1 Summary of Significant Plant Changes Description of Plant Change Impact on PRA Results Safety-related battery chargers are no An assessment of this change on the VEGP longer operated in a load-share internal events PRA model indicated no configuration. Instead, a single charger significant impact. will be in service and if it fails, the other charger will be placed in service by The battery chargers are modeled as operator action. seismically correlated. Thus, modeling the change would not affect the Seismic PRA results. Permanently installed and portable Credit for FLEX equipment is likely to FLEX equipment (other than low improve the PRA results, but the impact is leakage RCP seals) have not been difficult to quantify without detailed modeling. modeled in the PRA. The Westinghouse RCP shutdown seals have been installed at VEGP, and therefore are credited in the PRA model per the guidance in PWROG-14001-P (Reference 29) although the NRC has not yet issued a Safety Evaluation for this guidance. 2.3.2 Parts of the VEGP PRA Used Version 5 of the VEGP Units 1 and 2 Internal Events PRA model is used for the GSl-191 risk assessment. The internal events PRA model is an at-power model (i.e., it addresses Modes 1 and 2 of reactor operation). The model includes both CDF and LERF from internal events, including internal flooding. Version 2 of the VEGP Units 1 and 2 seismic PRA model is used for the assessment of GSl-191 risk from seismically-induced LOCAs. These versions are part of the current VEGP PRA model of record at the time of this analysis. The latest CDF and LERF results for internal events (including internal flooding), fire, seismic, and other external hazards for VEGP Units 1 and 2 are provided in Table 1 VEGP Units 1 and 2 Internal and External Events Summary. E1-9

Enclosure 1 Introduction and Overall Summary Table 1 VEGP Units 1 and 2 Internal and External Events Summary Event Type Unit 1 CDF Unit 1 LERF Unit 2 CDF Unit 2 LERF (per/year) (per/year) (per/year) (per/year) Internal Events 2.52x10-5 7.33x10- 9 2.52x10-5 7.33x10- 9 Fire 3.86x10-5 1.39x1 o-6 4.52x10-5 1.56x1 o-6 Seismic 2.8x1 o- 6 3.3x10- 7 2.8x10- 5 3.3x10-7 Other External Screened Screened out Screened out Screened out out Total 4.39x10- 5 1.73x1o-s 5.05x1Q- 5 1.90x1Q- 6 It is noted that for VEGP Units 1 and 2, the Total CDF for internal and external events is less than 1.0x1 o-4 /year and the Total LERF is less than 1x1 o-5/year, and therefore meets RG 1.174 total risk guidelines (Reference 6). 2.3.3 Summary of the Risk Assessment Methodology The GSl-191 risk assessment methodology used for VEGP Units 1 and 2 involves quantifying the VEGP Units 1 and 2 internal events and seismic PRA models to determine the increase in risk from debris (i.e., the "risk attributable to debris"). The risk increase is defined as the difference in risk calculated considering debris effects and the risk calculated assuming debris is not present to determine both the increase in CDF (LlCDF) and the increase in LERF (LlLERF). Enclosures 2 and 3 provide a detailed description of the GSl-191 models and risk evaluation that were used to calculate LlCDF and LlLERF. 2.3.4 Key Assumptions and Approximations in the PRA Modeling uncertainties are considered in both the internal events PRA and the seismic PRA. Assumptions are made during the PRA development to address a modeling uncertainty because there is not a single definitive approach. The GSl-191 risk assessment methodology also incorporates various assumptions and approximations pertaining to modeling uncertainties. These assumptions and modeling uncertainties are reviewed to determine the impact on the GSl-191 risk assessment, as described in Enclosure 3, Section 14.2.3. 2.3.5 Assessment of PRA Model Technical Adequacy Internal Events PRA Model Numerous assessments of technical capability have been made for the VEGP Units 1 and 2 internal events PRA model: E1-10 L_ ..

Enclosure 1 Introduction and Overall Summary

• An independent PRA peer review was conducted under the auspices of the Westinghouse Owners Group 0fVOG) in 2001, following the industry PRA peer review process described in NEI 00-02 (Reference 30). This peer review included an assessment of the PRA model maintenance and update process. All "Grade B" findings (there were no "Grade A" findings) were resolved in VEGP PRA model Revision 3.
• In 2005, the VEGP PRA model results were evaluated in the WOG PRA cross-comparison study performed in support of implementing the Mitigating Systems Performance Indicator (MSPI) process. After allowing for plant-specific features there were no MSPI cross-comparison outliers for the VEGP PRA.
• In 2006, MAAP (Modular Accident Analysis Program) evaluations performed for the VEGP PRA model were reviewed by an industry MAAP expert (from Fauske Associates, Inc., the company that developed the MAAP code) to check errors and reasonableness of the MAAP results. No significant issues were found from the review.
• In 2006, a gap analysis was performed for Revision 3 of the VEGP PRA model by an independent contractor against the 2005 addenda to the 2002 version of the ASME/ANS PRA Standard and the 2004 trial use version of RG 1.200. The major gaps related to documentation (especially system notebooks), the internal flooding PRA, and the treatment of uncertainty correlations were resolved in VEGP PRA model Revision 4 in 2009.
• In 2007, during the NRG review of severe accident mitigation alternative (SAMA) analysis for VEGP license renewal, the NRG issued RAls for the VEGP PRA "L2UP" model related to dominant minimal cutsets for GDF, LERF, and other Level 2 release categories, questions about PRA quality, and Level 2 methodology. SNC provided responses to the RAls and no additional RAls were received from the NRG.
• In 2008 and 2009, as a part of MSPI margin improvement, the VEGP PRA Level 1 model was reviewed by an independent contractor (Westinghouse) to identify any excessive conservatism in the PRA model. The review concluded that there were no significant issues or excessive conservatism in the VEGP PRA that needed to be revised or refined.
• In 2008, a gap analysis was performed by an independent contractor (ERIN) for the VEGP internal flooding PRA. Issues from the gap analyses were resolved before finalizing the internal flooding PRA update.
• In 2008, a gap analysis was performed by an independent contractor (Scientech) for the human reliability analysis (HRA) and dependency analyses for post-initiator human failure events. No significant issues were found.
• An industry peer review was performed in May 2009. The Pressurized Water Reactor Owners Group (PWROG) peer review was based on the 2007 addenda to 2002 version of the ASME/ANS PRA standard and RG 1.200 Revision 1. The VEGP PRA was found to meet Capability Category II (CC-II) or better for most of the supporting requirements (SRs) in the PRA standard. The outstanding issues primarily pertained to documentation. A total of 46 facts and observations (F&Os) were identified, 11 of which were categorized as "Findings" (which were related to documentation). Seven of the F&Os recognized areas of strength in the PRA.

E1-11

Enclosure 1 Introduction and Overall Summary

• In 2011 the VEGP PRA model was reviewed along with F&Os from the 2009 peer review to determine if model changes were necessary to be able to assess the risk impact of the proposed surveillance frequency change per NEI 04-10, "Risk Informed Method for Control of Surveillance Frequencies". Open F&Os related to the systems of interest or that could potentially impact the results of the assessment were dispositioned as having no impact, incorporated into the model, or addressed with sensitivity analyses. PRA modeling changes were identified and incorporated into the model. Several components were added to the VEGP PRA model during this task.
• In 2011 the VEGP internal events PRA model (including flooding) was reviewed (along with the fire PRA model) to determine the technical capability for use in supporting the license amendment request to implement NEI 06-09, "Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines". The review demonstrated and documented that the VEGP at-power internal events PRA model (including flooding) and the fire PRA model conform to the PRA standard at CC-II which satisfies the guidance of RG 1.200, Revision 2. In addition, the VEGP PRA model complies with all requirements for technical adequacy of the baseline PRA as defined in NEI 06-09.
• In 2013 a review and update of the*MAAP parameter file was performed by an independent contractor (Fauske) to check for errors and reasonableness of MAAP results. A significant upgrade in MAAP capabilities was initiated.
• In 2014 a review was performed by an independent contractor (Reliability and Safety Consulting Engineers, Inc.) for initiating events and data update.
• In 2014 a review by an independent contractor (Scientech) was performed for the HRA and dependency analysis for post-initiator human failure events. Human error '

probabilities (HEPs) were reviewed, scenario timing verified, and a new dependency analysis was implemented.

• In 2014 a review was performed by an independent contractor (Nuenergy) which focused on the model of record and plant interface. presents the finding-level F&Os from the 2009 VEGP internal events PRA peer review F&Os, based on the 2007 addenda to the PRA standard. The resolution for each finding is described and the manner of that resolution is referenced. All VEGP internal events PRA model peer review findings are resolved. presents the additional/revised requirements associated with the most recent PRA standard (issued in 2009) as amended by RG 1.200, Revision 2. This table also describes the VEGP PRA model and documentation changes that assure consistency with the latest endorsed versions of the PRA standard and RG 1.200.

Seismic PRA Model Version 2 of the VEGP Unit 1 and 2 seismic PRA reflects the as-built and as-operated plant as of August 31, 2015. The VEGP seismic PRA model has been assessed against RG 1.200, Revision 2. Specifically, the model was subject to a self-assessment and a E1-12

Enclosure 1 Introduction and Overall Summary peer review conducted by the PWROG in November 2014. A total of 73 F&Os were identified, 46 of which were categorized as findings and 27 as suggestions. The peer review team determined that the VEGP seismic PRA model is of good quality and integrates the seismic hazard, the seismic fragilities, and the systems-analysis aspects appropriately to quantify COF and LERF. The general conclusion of the peer review was that the VEGP seismic PRA is judged to be suitable for use for risk-informed applications. After the seismic PRA peer review, the peer review finding-level F&Os were appropriately dispositioned, and the seismic PRA model was updated to reflect these dispositions and further refine several fragility values. The seismic PRA peer review conclusions, the disposition of the finding-level F&Os, and the discussion below demonstrate that the VEGP seismic PRA is technically adequate for all aspects of this submittal. Attachment 3 provides a summary of VEGP seismic PRA peer review finding-level F&Os and their disposition, as well as sensitivity analyses performed to address issues identified in the findings. 2.3.6 Capability Categories for Parts of the PRA The 2007 version of the PRA standard used for the May 2009 peer review of the VEGP internal events PRA contains a total of 327 SRs in nine technical elements and one configuration control element. Eleven of the SRs represent deleted requirements (IE-AB, IE-A9, SC-A3, SY-A9, SY-89, HR-GB, IF-A2, IF-84, IF-02, IF-E2, and QU-02), and 20 were determined to be not applicable to the VEGP PRA. Thus, a total of 296 SRs was applicable to the VEGP internal events PRA. Among the 296 applicable SRs, 99% met Capability Category II or higher, as shown in Table 1-4. Ca ability Catego Met No. of SRs  % of total applicable SRs CC-1/11/111 (or SR Met 210 70.9% CC-11/111 24 8.1 % CC-I/II 14 4.7% CC-Ill 7 2.4% CC-II 38 12.8% CC-I 0 0% SR Not Met 3 1.0% Total 296 100% Three SRs were judged to be "Not Met". These were HR-G6, QU-03, and LE-G5. Supporting requirement HR-G6 was characterized as Not Met because the reasonableness check of the HRA was done for the previous revision of the PRA and not the latest revision. Supporting requirement QU-03 was characterized as Not Met because the SR requires the PRA results to be compared with those from similar plants. The VEGP PRA report cited the MSPI benchmark report as evidence of meeting this E1-13 L___

Enclosure 1 Introduction and Overall Summary requirement, which was an outdated comparison. Supporting requirement LE-GS was characterized as Not Met because limitations of the LERF calculations that could impact risk-informed applications were not identified. Resolution of the Findings HR-G6-01, QU-03-01, and LE-GS-01 resulted in SRs HR-G6, QU-03, and LE-GS being met at a Capability Category 1/111111. Thus, the VEGP internal events PRA (including flood) meets the requirements of RG 1.200. The 2013 addenda to the 2008 version of the PRA Standard used for the November 2014 peer review of the seismic PRA contains a total of 77 SRs in three technical elements. Of the 77 SRs, a total of 67 (87%) were met at Capability Category II or higher. The 10 SRs that were judged to be "Not Met" are listed in Table 1-S, along with the associated finding-level F&Os (16 total).

               - - VEGP Se1sm1c Ta bl e 1 5             .  . PRA N0 t Me t/CC -I SRs an dAssoc1a  . te d F"m d"mgs SR                                                       Findings SHA-C4                             12-18, 12-36 SHA-H1                             12-18, 12-36 SHA-11                             12-1S SHA-12                             12-1S SHA-J1                             12-1, 12-2, 12-11, 12-16 SHA-J3                             12-8 SFR-A2                             14-1 ' 14-7' 14-1 0 SPR-82                             16-4, 16-6 SPR-84                             16-1 SPR-F1                             12-31, 16-S As the table indicates, the following 10 SRs were assessed at less than CC-II.

• Six of the SRs are related to the seismic hazard analysis (SHA), for which the seven findings pertain to: (a) inadequate documentation of the hazard analysis; (b) demonstration that sufficient consideration has been given to more recent geologic events and associated modeling; and (c) sensitivity calculations for the models and parameters used in the site hazard. The documentation items have been addressed, as noted in Attachment 3 for the affected Findings listed in Table 1-S.
• One of the SRs is related to the seismic fragility analysis (SFR). Two of the three findings associated with this SR deal with conservatisms that have now been addressed within the analytical methodology. The remaining finding is associated with a specific polar crane fragility issue, and has also been addressed within the reviewed methodology, as noted in Attachment 3 for the affected findings listed in Table 1-S.
• Three of the SRs are related to the seismic plant response (SPR) model. Three of the five findings associated with this SR are related to implementation of the seismic performance shaping factor approach in the HRA. Those findings have been addressed and implemented in the seismic PRA model, without significant impact on E1-14

Enclosure 1 Introduction and Overall Summary the results. One finding is related to the relay chatter evaluation, and is resolved in the latest model update. The last finding is related to the SPR documentation, which has been updated to resolve the issue. The information provided in this section demonstrates that the VEGP internal events and seismic PRA models meet the technical adequacy requirements of RG 1.200, Revision 2 and is of sufficient quality and level of detail to support the risk informed approach for GSl-191. 3.0 Initiating Event Frequencies The initiating events relevant to GSl-191 at VEGP are LOCAs and SSBls. The LOCA and SSBI frequencies from the base PRA model were used (with some modifications for the GSl-191 evaluation as described below). The initiating event frequencies used in the VEGP base PRA model are consistent with the requirements of the ASME/ANS PRA Standard (Reference 10) as endorsed by RG 1.200 (Reference 7), and confirmed by the VEGP base PRA model peer review. The LOCA frequencies in the base PRA model are based on the geometric mean aggregation in NUREG-1829 (Reference 8) for medium and large LOCAs and the NRC initiating event database for small LOCAs (Reference 9). Although the small break LOCA frequency is nearly an order of magnitude lower than that produced from NUREG-1829 data, it has been demonstrated that the GSl-191 risk impact is not sensitive to the initiating event frequency for small LOCAs, because these breaks are not predicted to generate enough debris to cause strainer or core failures at VEGP (see , Section 14.1 ). The parametric uncertainty associated with using the mean frequency was addressed by a sensitivity analysis using the 5th and 95th percentile frequency to calculate the GSl-191 CFPs and LlCDF (see Enclosure 3, Section 14.2.2). In addition, the uncertainty associated with the use of the geometric aggregation for medium and large LOCA frequencies was assessed by performing a sensitivity analysis using the arithmetic mean aggregation (see Enclosure 3, Section 14.2.3). The SSBI frequency in the base PRA model is based on updated industry initiating event data (Reference 11 ). For the GSl-191 evaluation, the SSBI frequency was separated for main steam line breaks (MSLBs) and feedwater line breaks (FWLBs) due to the significant difference in debris quantities that could be generated by these breaks. Based on a closer review of the FWLB frequencies, two industry events between 1987 and 1995 actually occurred outside containment, and therefore the FWLB frequency contribution was recalculated using a Jeffrey's non-informative distribution for the GSl-191 evaluation. 3.1 SSBI Frequency Allocation As discussed in Enclosure 3, Section 14.1, the CFPs for SSBls were calculated independently for MSLBs and FWLBs. This evaluation conservatively assumed that all E1-15

Enclosure 1 Introduction and Overall Summary SSBI breaks were double-ended guillotine breaks (DEGBs). Therefore, the MSLB frequency was split evenly among all of the breaks analyzed on the main steam lines to calculate the MSLB GSl-191 CFP, and the FWLB frequency was split evenly among all of the breaks analyzed on the feedwater lines to calculate the FWLB GSl-191 CFP. 3.2 LOCA Frequency Allocation For the calculation of small, medium, and large LOCA CFPs, the LOCA frequencies were allocated to individual pipe welds using a top-down distribution methodology. The top-down LOCA frequency allocation methodology essentially treats all breaks of a similar size as having an equivalent LOCA frequency regardless of the weld size and associated degradation mechanisms. For specific break sizes within the small, medium, and large LOCA categories, the PRA model LOCA frequencies were interpolated using a semi-log interpolation scheme (i.e., linear interpolation between break sizes and logarithmic interpolation between frequencies). The uncertainty associated with the top-down LOCA frequency allocation was assessed using a sensitivity analysis with different weld-specific LOCA frequency allocation weighting schemes. For this sensitivity, welds were classified as having a high, medium, or low rupture probability based on the weld-specific degradation mechanisms, and the frequency allocations were weighted accordingly (see Enclosure 3, Section 14.2.3). Pipe LOCAs were postulated at weld locations. As described in Enclosure 2 Section 3.a.1, a range of break sizes and orientations were evaluated for all in-service inspection (ISi) welds in the unisolable portion of the Class 1 pressure boundary. Non-pipe LOCAs were not explicitly evaluated. Non-pipe components whose failure would result in a LOCA include nozzles, component bodies, pressurizer heater sleeves, manways, control rod drive mechanism penetrations, safety relief valves, reactor coolant pump seals, the reactor vessel, the pressurizer vessel, the steam generator vessels, welded caps on retired lines, and other components. It was reasonably assumed that breaks at any of these non-piping components would be bounded by already-analyzed breaks at pipe weld locations. With the exception of non-pipe components that are located in the reactor cavity, all of these non-pipe components are located at or near pipe welds. For example, there are many weld locations in lines around the pressurizer vessel including the surge line, spray lines, and the safety and relief valve lines that could be used to estimate debris generated from non-pipe components in that area of containment. In addition, there are many welds distributed along the cold legs, including those near the reactor coolant pumps, that could be used to estimate debris generated from non-weld locations in those areas. The modeled welds that are located at the safe ends on the nozzles at the reactor vessel, the pressurizer vessel, and the steam generator vessels are reasonably close to the associated nozzle welds and are close enough to the vessels to produce significant debris from the insulation around those vessels. Non-pipe components associated with E1-16

Enclosure 1 Introduction and Overall Summary the reactor vessel such as control rod drive penetrations, manways, and instrument lines connected to the reactor vessel, etc., are located away from the hot and cold leg nozzles and are not near modeled pipe weld locations. However, any quantity of debris generated by non-pipe component welds located in the reactor cavity will be bounded by a reactor vessel nozzle break. 3.3 Seismically Induced LOCAs In order to evaluate the risk impact from GSl-191 due to seismically-induced LOCAs, the VEGP Internal Events PRA model that was modified to perform the risk-informed GSl-191 evaluation was used as a guide to make corresponding modifications to the VEGP seismic PRA model. The GSl-191 risk impact presented in Table 1-1 therefore includes the risk increase from seismically-induced LOCAs. Enclosure 3 provides a description of the method used to calculate the LiCDF and LiLERF values for

                                                                                         \

seismically-induced LOCAs. 3.4 Water Hammer-Induced LOCAs The approach used to demonstrate that the risk of water hammer is acceptably low is to verify that the potential for water hammer is not likely to cause pipe rupture in the break locations that can produce and transport problematic debris. The portions of the VEGP RCS that are subject to a LOCA are designed to the Class 1 requirements of Section Ill of the ASME Boiler and Pressure Vessel Code, which includes consideration of appropriate transients. The reactor coolant pressure boundary (RCPB) is designed to accommodate the system pressures and temperatures attained under the expected modes of plant operation, including anticipated transients, with stresses within applicable limits. Consideration is given to loadings under normal operating conditions and to abnormal loadings, such as pipe rupture and seismic loadings. Pressurizer piping is a primary area of consideration due to its function during RCS pressure transients. The pressurizer safety valve, including valve supports, is designed for loads due to water relief, including the passage of a water slug and the effects of water hammer. The pressurizer is also instrumented to monitor for indications of RCS leakage that would contribute to creating a water hammer condition and the VEGP Technical Specifications (TS) impose limits on RCS operational leakage. Because the RCS is kept water-solid during operation, a water-hammer event can only be introduced from one of the systems that interact with the primary loop piping. At VEGP, the only systems that flow into the primary loop piping are the safety injection (SI) system, the residual heat removal (RHR) system, and charging from the chemical and volume control system (CVCS) (References 13, 14, and 15). The potential for gas accumulation in the ECCS, which includes the CVCS, RHR, and SI sub-systems, is addressed under VEGP's response to GL 2008-01 (Reference 13). To address GL 2008-01, VEGP performed a review of site documents, procedures, and E1-17

Enclosure 1 Introduction and Overall Summary equipment, and implemented modifications and document revisions as necessary. These changes included adding vent valves, revising procedures to include ultrasonic testing for gas voids, and creating/maintaining an active program to prevent, monitor, and trend gas voids in the ECCS and CSS (References 14, 15, 25, and 26). VEGP's documented resolution of GL 2008-01 was accepted by the NRC and deemed effective in precluding gas accumulation in the ECCS and CSS, and, therefore, preventing a water hammer in these systems. (References 27 and 28). Lastly, VEGP performed a search of corrective action program data for water hammer and found no issues in systems related to GSl-191 locations of concern. Based on the fact that the piping is designed to ASME Ill Class 1 standards, the implementation of an approved gas accumulation prevention/monitoring program, and the lack of historical data for water hammer events, the relevance of water hammer events in the context of GSl-191 is deemed insignificant. Therefore, LOCA frequencies are not impacted for VEGP Units 1 & 2 due to water hammer considerations in these systems. 4.0 Defense-in-Depth and Safety Margin As described in RG 1.174 (Reference 6), sufficient defense-in-depth and safety margin must be maintained. Both of these aspects were evaluated in detail as described in . 5.0 Uncertainty Uncertainty quantification is a key requirement in RG 1.174 for a risk-informed evaluation (Reference 6). As defined in RG 1.174 and explained in more detail in NUREG-1855 (Reference 16) and two corresponding EPRI reports (References 17 and 18), there are three types of uncertainty that should be addressed:

1. Parametric uncertainty

2. Model uncertainty

3. Completeness uncertainty Parametric uncertainty refers to the variability in input parameters that are used in the risk assessment. Due to the wide range of plant-specific post-LOCA conditions related to GSl-191 phenomena, this is a very important aspect for understanding the overall uncertainty.

Model uncertainty refers to the potential variability in an analytical model when there is no consensus approach. A consensus approach is a model that has been widely adopted or accepted by the NRC for the application for which it is being used (Reference 16). For example, the use of a spherical zone of influence (ZOI) to model the debris quantity generated by a high energy break is a consensus model that has been widely adopted and accepted by the NRC (References 19 and 20). In general, the VEGP GSl-191 evaluation has been performed using standard models that have been widely accepted for deterministic evaluations (e.g., accepted ZOI sizes and prototypical E1-18

Enclosure 1 Introduction and Overall Summary strainer module testing). By using these consensus approaches, the effort to address model uncertainty is minimized. Completeness uncertainty refers to a) the uncertainty associated with scenarios or phenomena that are excluded from the risk evaluation, and b) the uncertainty associated with unknown phenomena. Although it may not be practical to quantify the uncertainty associated with factors that are not explicitly modeled, their potential impact can be qualitatively assessed. Uncertainties associated with unknown phenomena, on the other hand, cannot even be qualitatively assessed. Uncertainties associated with unknown phenomena are the reason that it is important to maintain defense-in-depth and safety margins (see Enclosure 4). Because all of the cases that were evaluated for model uncertainty and parametric uncertainty resulted in a b.CDF less than 1x1 o-6 (see Section 5.1 and Section 5.2), it can be concluded with high confidence that the risk associated with GSl-191 is very low as defined by the acceptance guidelines in RG 1.174 (Reference 6). 5.1 Parametric Uncertainty The parametric uncertainties associated with the VEGP risk-informed GSl-191 evaluation were evaluated in a very conservative manner by analyzing the worst case combinations of input parameters associated with strainer and core failures. Although the scenario is hypothetically possible, the probability of all of the worst-case conditions occurring simultaneously is extremely unlikely. As described in Enclosure 3, Section 14.2.3, the results of this evaluation showed that the parametric uncertainties are low (i.e., the resulting b.CDF still falls within RG 1.174 Region Ill even under the worst-case combination of input parameter values). 5.2 Model Uncertainty The model uncertainties were quantified using sensitivity analysis for models where no consensus exists. Sensitivities were run for the following models:

• Break model (continuum vs. DEGB-only)
• LOCA frequencies (VEGP PRA vs. NUREG-1829 arithmetic mean)
• LOCA frequency allocation to individual welds (top-down vs. degradation mechanism probability weighting)
• Containment spray (CS) actuation (CS actuates for hot leg breaks larger than 15 inches vs. CS actuating for more or fewer breaks)
• Aluminum metal release equation (UNM vs. WCAP-16530)
• Fiber bed thickness required to apply chemical head loss (0.45 inches vs.

0 inches)

• LBLOCA size range discretization (base case allocation of frequencies vs.

allocations biased to smaller break sizes and larger break sizes) E1-19

Enclosure 1 Introduction and Overall Summary As described in Enclosure 3, Section 14.2.3, the uncertainty associated with each of these models is low (i.e., the resulting ~CDF still falls within RG 1.174 Region Ill for each sensitivity that was evaluated). 5.3 Completeness Uncertainty Completeness uncertainty was qualitatively determined to be low. As described below, the VEGP evaluation was rigorous and comprehensive, and the areas that were not explicitly evaluated have a low potential for any significant risk impact:

• The range of hazards, initiating events, and plant operating modes were considered as described in Section 2.1.
• LOCAs and consequential LOCAs from SSBls were directly evaluated in the risk quantification as described in Section 2.2.

o The SSBI evaluation included an analysis of DEGBs spaced no more than 5 ft apart along each of the main steam and feedwater lines. o The LOCA evaluation included pipe breaks on every ISi weld within the Class 1 pressure boundary inside the first isolation valve. o Break sizes ranging from %-inch to a full DEGB were postulated on each weld. o Partial breaks (i.e., breaks smaller than a DEGB) were evaluated in 45-degree increment orientations around the pipe for each break size. o Debris quantities were calculated for breaks on ISi welds outside the first I I isolation valve, and there is no significant difference between the type and quantity of debris generated for these breaks compared to similar size breaks inside the first isolation valve. Due to the low probability of isolation valve failure, breaks outside the first isolation valve are insignificant with respect to GSl-191 risk at VEGP. o Non-pipe LOCAs were shown to be reasonably represented or bounded by adjacent pipe breaks as described in Section 3.2. o High likelihood equipment configurations were explicitly evaluated. o Low likelihood equipment configurations were addressed using a bounding approach.

• The risk of seismic and water hammer-induced LOCAs was shown to be low as described in Sections 3.3 and 3.4.
• As described in Section 1.0 and Enclosure 2, all known GSl-191 phenomena and debris failure mechanisms were evaluated either in a bounding manner for phenomena not explicitly included in the VEGP risk model or in a reasonably conservative manner for phenomena that were included in the risk model.

Although there is also some uncertainty associated with unknown phenomena, this uncertainty is judged to be small. The nuclear industry has been actively addressing GSl-191 concerns for PWRs for well over a decade. In addition, the boiling water reactor (BWR) strainer performance issue dates back to 1992, and unresolved safety issue (USI) A-43 dates back to 1979. Numerous tests have been performed by the U.S. E1-20

Enclosure 1 Introduction and Overall Summary NRC and industry, as well as regulators and utilities around the world over the last 35 years to resolve issues related to debris and strainer performance. This testing has investigated nearly every aspect of GSl-191 including insulation and coatings destruction from break jets; unqualified coatings failure; blowdown and washdown debris transport; containment pool settling, tumbling, and lift-over-curb debris transport; debris erosion; chemical release, solubility, and precipitation; strainer head loss, vortexing, and penetration; ex-vessel component wear; and in-vessel core blockage and boron precipitation. Based on the extensive research that has been performed, it is unlikely that there are unidentified phenomena that would significantly increase the risk of GSl-191 related failures. 6.0 Monitoring Program VEGP has implemented procedures and programs for monitoring, controlling, and assessing changes to the plant that have a potential impact on plant performance related to GSl-191 concerns. These provide the guidance to inspect the condition of the sump strainers and the ability to assess impacts to the inputs and assumptions used in the PRA and the associated engineering analysis that support the proposed change. Programmatic requirements ensure that the potential for debris loading on the sump does not materially increase. In addition, programs and procedures have been implemented to evaluate and control potential sources of debris in containment. 7.0 Quality Assurance Most of the analysis and testing for the risk-informed GSl-191 evaluation was performed as safety related under vendor quality assurance (QA) programs compliant with 10 CFR 50 Appendix B. The following exceptions are noted:

• The aluminum release equation used for the NARWHAL CFP calculation was developed through testing at the University of New Mexico (UNM). Although the testing was not performed under an Appendix B QA program, it was conducted using standard practices for academic research at the same facility where the NRG-sponsored integrated chemical effects testing (ICET) was conducted (Reference 21). The test results (including the aluminum release equation) were peer reviewed and published in a scientific journal (Reference 22). The UNM aluminum release model was qualified for safety related use at VEGP as described in Enclosure 2, Section 3.o.2.9. Finally, a sensitivity calculation was performed to address the model uncertainty associated with the use of the UNM aluminum release equation (see Enclosure 3, Section 14.2.3).
• VEGP has a relatively high containment pressure setpoint for actuating containment sprays. The design-basis calculations show that this setpoint would be exceeded for a design-basis accident. However, best-estimate thermal hydraulic calculations performed for VEGP by Texas A&M University (TAMU) showed that hot leg breaks smaller than or equal to 15 inches and cold leg breaks up to DEGBs would not initiate containment sprays. Although the TAMU E1-21

Enclosure 1 Introduction and Overall Summary thermal-hydraulics work was not performed under an Appendix B QA program, it was prepared and peer reviewed using standard practices for academic research. Note that the results of the TAMU thermal-hydraulic analysis were only used to define which breaks would initiate containment sprays, and were not used to define the pressure and temperature profiles used in the NARWHAL CFP calculation. In addition, sensitivity calculations were performed to address the model uncertainty associated with the breaks that initiate containment sprays (see Enclosure 3, Section 14.2.3).

• The fiber penetration equations used for the NARWHAL CFP calculation were developed through testing at Alden Research Laboratory. Alden has a 10 CFR 50 Appendix B QA program. Although the testing was not officially conducted under the Alden QA program, it was performed using most of the same processes, reviews, and procedures in the QA program. In addition, sensitivity calculations were performed to evaluate the sensitivity to the fiber penetration fraction (see Enclosure 3, Section 14.2.2).
• The LOCA frequencies were allocated to individual welds using a top-down approach (see Section 3.2). To address the model uncertainty, a sensitivity analysis was performed using an alternate weighting scheme based on weld-specific degradation mechanisms. This weighting scheme was derived in part using information contained in a LOCA frequency evaluation prepared by KNF Consulting Services. Although the KNF evaluation was not performed under an Appendix B QA program, it provides a reasonable set of inputs for the purpose of the sensitivity analysis.
• The GSl-191 PRA calculations are not safety related, but were prepared as safety significant under the SNC QA program.

8.0 Periodic Update of Risk-Informed Analysis The risk-informed GSl-191 analysis will be updated within at least 48 months following initial NRC approval or since the last update. This update should include all parts of the risk-informed evaluation including the systematic risk assessment, consideration of defense-in-depth, and consideration of safety margin. The update should also include any new information on LOCA frequencies that may be developed. Reliability data, unavailability data, initiating event frequency data, human reliability data, and other similar PRA inputs are reviewed approximately every two fuel cycles to maintain the base VEGP PRA model consistent with the as-operated plant. In addition, existing procedures are in place for periodic updates of risk-informed applications. 9.0 Reporting and Corrective Actions Licensees are required to make a report to the NRC and take corrective action in the event that the risk of debris exceeds the NRC acceptance criteria or in the event that defense-in-depth or safety margins have decreased from NRG-approved analysis. The risk of debris is defined in terms of ~CDF and ~LERF, and the acceptance criteria are E1-22

Enclosure 1. Introduction and Overall Summary defined as the upper threshold for RG 1.174 Region 11 (i.e., 1x10-5 for ~CDF and 1x10-5 for ~LERF) (Reference 6). Defense-in-depth measures and safety margin are specifically defined for VEGP in . Any unacceptable changes in risk or reductions in defense-in-depth or safety margins that are identified through the monitoring program (see Section 6.0), the periodic updates (see Section 8.0), or other means will be reported to the NRC. In addition, these issues will be entered and tracked to resolution in accordance with the SNC corrective action program~ 10.0 License Application The specific requirements for the license application described in RG 1.174 (Reference

6) will be addressed at a later date.

11.0 References

1. SRM-SECY-10-0113, "Staff Requirements - SECY-10-0113 - Closure Options for Generic Safety Issue - 191, 'Assessment of Debris Accumulation on Pressurized Water Reactor Sump Performance'," December 23, 2010

2. SRM-SECY-12-0093, "Staff Requirements - SECY-12-0093 - Closure Options for Generic Safety Issue - 191, 'Assessment of Debris Accumulation on Pressurized-Water Reactor Sump Performance'," December 14, 2012

3. SNC Letter NL-13-0953 to NRC (ML13137A130), "Vogtle Electric Generating Plant Proposed Path to Closure of Generic Safety lssue-191, 'Assessment of Debris Accumulation on Pressurized Water Reactor Sump Performance'," May 16, 2013

4. Draft Regulatory Guide 1.229, Revision 0, "Risk-Informed Approach for Addressing the Effects of Debris on Post-Accident Long-Term Core Cooling"

5. NRC Letter (ML073110278), "Revised Content Guide for Generic Letter 2004-02 Supplemental Responses," November 2007

6. Regulatory Guide 1.174, Revision 2, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," May 2011

7. Regulatory Guide 1.200, Revision 2, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities,"

March 2009

8. NUREG-1829, "Estimating Loss-of-Coolant Accident (LOCA) Frequencies Through the Elicitation Process," April 2008

9. U.S. Nuclear Regulatory Commission, "Reactor Operational Experience Results and Databases, Initiating Events," http://nrcoe.inel.gov/resu ltsdb/I nitEvent/

(1988-2010 Summaries) 10.ASME/ANS RA-Sa-2009, "Addenda to ASME/ANS RA-S-2008 Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications," February 2009 E1-23

Enclosure 1 Introduction and Overall Summary

11. NUREG/CR-6928, "Industry-Average Performance for Components and Initiating Events at U.S. Commercial Nuclear Power Plants," Initiating Event Data Sheets Update 2010, January 2012

12. NUREG-1903, "Seismic Considerations for the Transition Break Size," February 2008

13. Generic Letter 2008-01, "Managing Gas Accumulation in Emergency Core Cooling, Decay Heat Removal, and Containment Spray Systems," January 11, 2008

14. NL-10-0062, "Vogtle Electric Generating Plant Unit 1 Nine-Month Supplemental (Post-Outage) Response to Nuclear Regulatory Commission Generic Letter 2008-01," January 20, 2010

15. NL-08-1921, "Vogtle Electric Generating Plant Unit 2 Nine-Month Supplemental (Post-Outage) Response to Nuclear Regulatory Commission Generic Letter 2008-01," January 21, 2009

16. NUREG-1855, Revision 1, "Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decisionmaking," March 2017

17. EPRI Report 1016737, "Treatment of Parameter and Model Uncertainty for Probabilistic Risk Assessments," December 2008

18. EPRI Report 1026511, Technical Update, "Practical Guidance on the Use of Probabilistic Risk Assessment in Risk-Informed Applications with a Focus on the Treatment of Uncertainty," December 2012

19. NEI 04-07, Revision 0, "Pressurized Water Reactor Sump Performance Evaluation Methodology, 'Volume 1 - Pressurized Water Reactor Sump Performance Evaluation Methodology'," December 2004

20. NEI 04-07, Revision 0, "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'," December 2004

21. NUREG/CR-6914, Volume 1, "Integrated Chemical Effects Test Project:

Consolidated Data Report," December 2006

22. Howe, Kerry J., et al., "Corrosion and solubility in a TSP-buffered chemical environment following a loss of coolant accident: Part 1 - Aluminum," Nuclear Engineering and Design, Volume 292, October 2015: 296-305

23. NRC Bulletin 2003-01, "Potential Impact of Debris Blockage on Emergency Sump Recirculation at Pressurized-Water Reactors," June 9, 2003

24. NRC Generic Letter 2004-02, "Potential Impact of Debris Blockage on Emergency Recirculation during Design Basis Accidents at Pressurized-Water Reactors,"

September 13, 2004

25. TSTF-523 (ML13053A075), Revision 2, "Generic Letter 2008-01, Managing Gas Accumulation," February 21, 2013

26. NRC Letter (ML16063A475), "Vogtle Electric Generating Plant, Units 1 and 2, Issuance of Amendments (CAC Nos. MF6213 and MF6214)," March 21, 2016 27.NL-10-1228 (ML102140115), "Vogtle Electric Generating Plant Response to NRC Generic Letter 2008-01 Response to Request for Additional Information,"

July 28, 2010

28. NRC Letter (ML11101A097), "Vogtle Electric Generating Plants, Units 1 and 2 -

Closeout of Generic Letter 2008-01, 'Managing Gas Accumulation in Emergency E1-24

Enclosure 1 Introduction and Overall Summary Core Cooling, Decay Heat Removal, and Containment Spray Systems' (TAC Nos. MD7892 and MD7893)," April 27, 2011 29.Topical Report PWROG-14001-P, Revision 1, "PRA Model for the Generation Ill Westinghouse Shutdown Seal," July 2014

30. NEI 00-02 (ML061510619), Revision 1, "Probabilistic Risk Assessment (PRA) Peer Review Process Guidance," May 2006 E1-25

Enclosure 1 Introduction and Overall Summary Vogtle Electric Generating Plant Supplemental Response to NRC Generic Letter 2004-02 Enclosure 1 Introduction and Overall Summary Attachment 1 Resolution of the VEGP Internal Events PRA Peer Review Findings E1-26

Enclosure 1 Attachment 1 - Resolution of the VEGP Internal Events PRA Peer Review Findings F&O Review cc F&O Description F&O Resolution Reference Number Element IE-A4-01 IE-A4 CCII The SR requires a systematic evaluation of each SR IE-A4 is met at Capability Category II V-RIE-IEIF-Met system to assess the possibility of an initiating event or equivalent level per Peer Review U00-001, occurring due to failure of the system. The reviewers Report. VEGP Electric could not find documentation of such a systematic Generating review. There is no technical issue associated Plant, Initiating with this F&O. Event Additional notes made by review team in response to Notebook, SNC's comments: When the reviewers asked for the A systematic systems evaluation of each June 2014, Initiating Events (IE) notebook (NB), they were told system, including support systems was Table 2 and that Chapter 2 of the main report is the IE NB. performed to assess initiating event Table 3 and Chapter 2 does not contain any evidence that a possibility due to system failure. The Appendix E systematic evaluation of each system was performed. results of this evaluation are documented and F. Nor does Chapter 2 contain a system failure modes in Table 2 and 3 and Appendix E and F and effects analysis (FMEA) as required by the of V-RIE-IEIF-U00-001, VEGP Electric Standard which would have been an acceptable Generating Plant, Initiating Event alternate. The fact that a systematic evaluation was Notebook, June 2014. performed during the Individual Plant Examination (IPE), in of itself, is not sufficient. The evaluation Discussed in Appendix D, table D.1 - performed for the IPE should have been reviewed and (page D.3) - Map of ASME Initiating a statement to that extent should have been Events (IE) SRs to the IE notebook. presented in the Chapter 2. In absence of such evidence, the review comment stays. This F&O is resolved. As noted elsewhere in the report, it is very important to have Qood documentation. IE IE-01 cc The lack of a central place for all the information SR IE-01 is met at Capability Category II V-RIE-IEIF-01 1/11/111 related to initiating events made it difficult for the or equivalent level per Peer Review U00-001, Met review team to review this topic. Most plants have all Report. VEGP Electric this information stored in a separate IE notebook. Generating The review team recommends that VEGP do the There is no technical issue associated Plant, Initiating same. with this F&O. Event Notebook, Additional notes in response to SNC's comments: The A review and update of the VEGP June 2014. review team disagrees with SNC's comments. The initiating events was completed in June Standard requires that the work be documented in a 2014. An IE Notebook was developed as manner that facilitates PRA applications, upgrades RIE calculation V-RIE-IEIF-U00-001 and peer review. The review team does not believe Initiating Events. The updated analysis that the work was documented a manner to facilitate notebook includes all the information E1-27

Enclosure 1 Attachment 1 - Resolution of the VEGP Internal Events PRA Peer Review Findings F&O Review cc F&O Description F&O Resolution Reference Number Element peer review. One could almost make a case for 'not related to the initiating events. met' categorization for this element as the Documentation of special initiators and documentation is the weakest link in this whole effort. system dependency analysis is included The F&O stays as written. in appendix E and F respectively. Appendix D provides mapping of ASME IE SRs to locations in the IE notebook. This F&O is resolved. AS-A11- AS-A11 cc Dependencies are not preserved for consequential SR AS-A 11 is met at Capability Category Documented 01 1/11/111 ATWS for the SLOCA initiator and the SGTR initiator. II or equivalent level per Peer Review in PRA-8C-V-Met The existing A TWS trees, based on a LOFW initiator, Report. 07-003 Rev were developed for transients that do not include a 4.0VEGP loss of RCS inventory or operator actions to mitigate a In the revised model, if ATWT occurred Internal Event SGTR. after a SLOCA or SGTR, the accident PRA model, sequence is treated in a similar way as a Chapter 5, Note: The review team decided to leave the F&O as is case with a stuck open PZR safety Section 5.2, after reviewing SNC's comments. valve(s) where inventory make up by item 27. high pressure injection and recirculation (Reference 4) as well as secondary heat removal is required to prevent core damage. This F&O is resolved. SY SY-83 cc The treatment of main or frontline system and SR SY-83 is met at Capability Category V-RIE-IEIF-01 1/11/111 supporting or mitigating system Common Cause II or equivalent level per Peer Review U00-013, Met Failure (CCF) event groupings do not appear to Report. VEGP Electric consistent in the current PRA documents. It appears Generating that for systems considered non-risk significant, CCF documentation is revised. Plant, reviews for CCF groups may not have been Common undertaken since the IPE modeling. A separate CCF Notebook (V-RIE-IEIF- Cause Factor U013 Common Cause Factors) was Notebook. The updated standards require a systematic treatment created. Potential CCF groups (for both of all systems, not just the main systems contributing risk significant and non-risk significant to core damage. New CCF groups may be required systems) are considered and or updated documentation as to why these groups are documented as identified. not required is needed. This F&O is resolved. Note: The review team decided to leave the F&O as is after reviewing SNC's comments. E1-28

Enclosure 1 Attachment 1 - Resolution of the VEGP Internal Events PRA Peer Review Findings F&O Review cc F&O Description F&O Resolution Reference Number Element HR-G6- HR-G6 Not Check of consistency and review for reasonableness This F&O is resolved. Reasonableness PRA-BC-V 01 Met is missing in the Revision 4 updated HRA draft and check for all HRAs for Revision 4 model 003 Rev 4.0, the prior revision document information related to was re-performed. All HRAs have been Chapter 8, these items is not appropriate to use in light of the determined to be reasonable or have Human updates performed and changes to the results. been appropriately revised. The Reliability Section 8 includes a table of human failure events reasonableness check is documented in Analysis for (HFEs) and human error probabilities (HEPs) but Section 8.2.2 of PRA-BC-V-07-003, VEGP PRA does not include HEP reasonableness check, as is Human Reliability Analysis for VEGP Model, Section documented in Section 8.3 of the November 2005 PRA Model Rev. 4.0 (Reference 4). 8.2.2. HRA update for Revision 3. This F&O is resolved. DA-C2- DA-C2 cc Generic data alone was used for the probability that a SR DA-C2 is met at Capability Category PRA-BC-V 01 1/11/111 PORV is blocked -- refer to Table 6.3.9. Since PORV II or equivalent level per Peer Review 003 Rev. 4.0 Met availability is a critical plant feature with respect to Report. VEGP Internal A TWS pressure control, the use of generic data for Events PRA this parameter is deemed a weakness. VEGP specific data was used for the model, probability that a PORV is blocked - Chapter 6, refer to Table 6.4-1 in Chapter 6, VEGP VEGP Data, Data, PRA-BC-V-07-003 Rev. 4.0 VEGP Table 6.4-1. Internal Events PRA model. This F&O is resolved. E1-29

Enclosure 1 Attachment 1 - Resolution of the VEGP Internal Events PRA Peer Review Findings F&O Review cc F&O Description F&O Resolution Reference Number Element IF-C2a- IF-C2a cc Because of a lack of well documented analysis, a lot SR IE-C2a is met at Capability Category V-RIE-IEIF-01 11111111 of information had to be obtained by talking to the II or equivalent level per Peer Review U00-008, Rev. Met analyst who performed the analysis, which is the Report. 1, Internal basis for the F&O. Flooding The VEGP Internal Flooding analysis Notebook Original F&O: From a more detailed review of the IPE uses the internal flooding assessment VEGP Electric flood calculations (which are the main input to defining conducted by ABS Consulting, and PRA- Generating the flood events and consequences), it is noted that BC-V-07-003, Rev.4.0 VEGP Internal Plant Internal successful operator mitigation of ALL flood events is Events PRA model, Chapter 5, Linked Flooding PRA, assumed to occur 30 minutes into any flood scenario Fault Tree, Section 5.3 Internal Flooding Rev. 2, dated and fully terminate the flood flow, and it appears to be IE Integration. These documents were 01/2009, based on assumptions only, as no detailed discussion combined into V-RIE-IEIF-U00-008, Rev. VEGP Design of the actual ability of operators to perform such 1, Internal Flooding Notebook. Manual# DC-actions is given. This appears to be in direct conflict Calculation V-RIE-IEIF-U00-008, Rev. 1, 1009 Flooding with the HFEs included (but not modeled in the PRA Internal Flooding Notebook resolves the lnterdiscipline, model) in the flooding report (assumed perfect comment of "lack of well documented PRA-BC-V response vs. HFE calculation). Also, the report lists analysis". 003, Rev.4.0 hundreds of pages of a detailed analysis approach Automatic and operator actions that have VEGP Internal using screening criteria, flow calculations, etc. and the ability to terminate or contain floods Events PRA only by locating very specific statements. are identified in V-RIE-IEIF-U00-008, model, Rev. 1, Internal Flooding Notebook, Chapter 5, There appear to be conflicts of the inputs to the Section 10, Evaluate Flood Mitigation Linked Fault flooding PRA and the subsequent discussions of Strategies. Actions to mitigate the flood Tree, Section operator mitigation as well as using the information are generally not credited. 5.3 Internal from the IPE calculations for propagation Flooding IE assessments. This is more than an editorial finding Bounding assumptions about flood Integration, and impacts the entire basis of using the older heights, propagation, and impact on VEGP Level 1 calculated results in the current analysis. equipment have been made (V-RIE-IEIF- PRA for At-U00-008, Rev. 1, Internal Flooding Power and Additional notes made in response to SN C's Notebook, Section 9.1, Flood Internal Floods comments: The lengthy flooding methodology outlined Characterization). Engineering (NRC). in the report is not used in the current VEGP flooding calculations for design basis flood results as mentioned in the original F&O. The conditions have been performed for each previous IPE flooding analysis is used as inputs to the flood area, but these calculations are not flooding targets and propagations for a bounding case directly referenced in the flooding estimation, and the more thorough analysis outlined in analysis. The flood water flow was the report is not undertaken but is in place for use (as successfully isolated at 30 minutes, and was explained by the SNC analyst in charge of the all calculations for flood volumes, flooding oroiect durinq the peer review and verv briefly oropaqations, etc. were done with the E1-30

Enclosure 1 Attachment 1 - Resolution of the VEGP Internal Events PRA Peer Review Findings F&O Review cc F&O Description F&O Resolution Reference Number Element mentioned in the flooding report). amount of water generated in this 30 minutes with considerations for system Each and every IPE flooding calculation reviewed characteristics". during the peer review contained the assumption that the flood water flow was successfully isolated at 30 While these actions are not explicitly minutes, and all calculations for flood volumes, modeled, the flooding model uses propagations, etc. were done with the amount of bounding assumptions with respect to water generated in this 30 minutes with operator actions (V-RIE-IEIF-U00-008, considerations for system characteristics. If any IPE Rev. 1, Internal Flooding Notebook, flooding calculations are done, which do not contain Section 9.1, Flood characterization). This this assumption, they were not seen during the peer resolves the comment "successful review. operator mitigation of ALL flood events is assumed to occur 30 minutes into any No change to the F&O is warranted. The problem with flood scenario and fully terminate the this scenario of using the IPE flooding calculations for flood flow". inputs to the described methodology is the following: If the flooding analysis was performed in accordance to The calculations were reviewed by ABS the methodology outlined in the current report, new and RIE analysts and an independent flooding volumes and propagation assessments would plant walkdown was performed. These be required that did not take into account successful activities support that model isolation at 30 minutes (as the IPE calculations do) assumptions are conservative. since another operator isolation assessment is outlined in the flooding report methodology for normal There is no technical issue associated HFE calculations for flow isolation. with this F&O. This F&O is resolved. QU QU-03 Not Reviewer asked the VEGP staff to provide evidence A new comparison study was performed V-RIE-IEIF-01 Met of comparison of the VEGP results to those from by comparing VEGP PRA results with U00-001, similar plants. The VEGP staff presented the two PWR PRAs (Callaway and Wolf Initiating benchmark report for MSPI as evidence of Creek), which are considered relatively Events comparison. Reviewers concluded similar to VEGP. In addition to the Notebook, that report is not sufficient evidence for demonstrating comparison of PRA reports, a plant visit Section 2, compliance to this SR. to Callaway was performed to identify Table 4 more details of Callaway systems and PRA modeling. The plant comparisons were again E1-31

Enclosure 1 Attachment 1 - Resolution of the VEGP Internal Events PRA Peer Review Findings F&O Review cc F&O Description F&O Resolution Reference Number Element updated with the 2014 initiating event update and V-RIE-IEIF-U00-001 Initiating Events Notebook was updated to reflect the changes. The comparison showed that all plants have loss of offsite power (LOSP)/station blackout (SBO) as the most dominant contributors which indicated that the VEGP PRA results are not an outlier, as compared to similar PWRs. Differences in dominant CDF contributors were investigated, and it was found that those differences are due to differences in details of system configuration/ operation and physical barriers for internal flooding and in the sources for generic initiating event frequency data (VEGP PRA used the latest generic (2010 - 2013) initiating frequency and failure data along with VEGP specific experience data for its data update). This F&O is resolved. QU-F5- QU-F5 cc In Chapter 10, there is insufficient documentation for SR QU-F5 is met at Capability Category PRA-BC-V-01: 1/11/111 the quantification process, which would impact II or equivalent level per Peer review 007-003, Met application (only EOOS). Reviews conclude that the Report. VEGP Internal documentation currently in Chapter 10 is not sufficient Event PRA, to meet this SR fully. Chapter 10 of PRA-BC-07-003 (VEGP Rev4.0, internal PRA model rev 4.) contains Chapter 10, detailed information for quantification VEGP PRA process and also identifies limitations Level 1 and which would affect application. The Level2 identified limitation in quantification Evaluation, process is that the average model Recovery assumed a specific plant system Analysis and alignment configuration. This assumed Uncertainty system alignment configuration is Analysis. chanQed to reflect actual system E1-32

Enclosure 1 Attachment 1 - Resolution of the VEGP Internal Events PRA Peer Review Findings F&O Review cc F&O Description F&O Resolution Reference Number Element alignment configuration when configuration specific risk is evaluated. There is no other limitation in quantification process. This F&O is resolved. LE-GS- LE-GS Not Limitations in the LERF analysis that would impact A comparison of VEGP LERF scenarios PRA-BC-V 01 Met applications are not identified. The LERF analysis with those in Table 4.S.9.3 of the ASME 003, VEGP documentation is incomplete because limitations in PRA standard revealed that the VEGP Internal Event the LERF analysis that would impact applications, as PRA included more potential LERF PRA, Rev 4.0, required by SR LE-GS, are not identified. scenarios than as required for a large dry Chapter 9, containment plant in ASME PRA VEGP Level 2 standard. PRA Modeling. The LERF scenarios modeled in VEGP PRA include containment bypass core damage scenarios (steam generator tube rupture and Interfacing systems LOCA), thermally or pressure induced steam generator tube rupture after core damage, containment isolation failure with core damage and various early containment failure modes. This F&O is resolved. MU MU-84 cc The VEGP plant procedures do not specifically call for Procedure RIE-014, Configuration See procedure 01 11111111 a peer review after a PRA upgrade has been Management of PRA Models, Qualitative RIE-014, Met completed. But the plant has had this peer review and Models and Software outlining Configuration other peer reviews in the past. This change is requirements dealing with PRA Management required by the SR. configuration control, as referenced in of PRA ASME/ANS RA-Sa-2009, Section 1-S, Models, have been developed to comply with Qualitative requirements of RITS Initiative 4b. Models and Software, Procedure RIE-001, Generation and Section 4.2 Maintenance of PRA Models and Associated Updates, Section 4.2.8 states See procedure the requirement for either a full or RIE-001, focused peer review. Generation E1-33

Enclosure 1 Attachment 1 - Resolution of the VEGP Internal Events PRA Peer Review Findings F&O Review cc F&O Description F&O Resolution Reference Number Element and The PRA model update process is Maintenance discussed in Section 5 of this licensing of PRA submittal. Models and Associated This F&O is resolved Updates, Section 4.2.8. AS ME/ANS RA-Sa-2009, Section 2-3 and Appendix 1-A. NEI 00-02, PRA Peer Review Process Guidance. E1-34

Enclosure 1 Introduction and Overall Summary Vogtle Electric Generating Plant Supplemental Response to NRC Generic Letter 2004-02 Enclosure 1 Introduction and Overall Summary Attachment 2 Comparison of RG 1.200 Revision 1 and Revision 2 Supporting Requirements E1-35

Enclosure 1 Attachment 2 - Comparison of RG 1.200 Revision 1 and Revision 2 Supporting Requirements SR in 2007 PRA Standard as SR in 2009 PRA Standard as Description of Change Resolution Amended by RG 1.200, Revision 1 Amended by RG 1.200, Revision 2 IE-C10: IE-C12: The sentences were NUREG/CR-6928 is used as clarifications provided in RG the source for generic data CC-1/11/111: CC-1/11/111: 1.200 Revision 1 and Revision 2, priors in Revision 5 of the ... ... respectively . VEGP internal events PRA. An example of an acceptable generic An example of an acceptable generic data sources is NUREG/CR-5750 data sources is NUREG/CR-6928 The updated SR cites a more [Note 1]. [Note 1]. recent example of an acceptable Qeneric data source. SY-815: SY-814: The sentences were As noted in Table 9.2-1 of the clarifications provided in RG internal events PRA CC-1/11/111: CC-1/111111: 1.200 Revision 1 and Revision 2, calculation, failure of the ... ... respectively . containment boundary due to (h) harsh environments induced by (h) harsh environments induced by venting is not applicable to the containment venting, or failure that containment venting, failure of the The updated SR explicitly VEGP large, dry, sub-may occur prior to the onset of core containment venting ducts, or failure requires consideration of atmospheric containment. damage. of the containment boundary that containment venting ducts and may occur prior to the onset of core failure of the containment damage boundary prior to core damage. DA-C1: DA-C1: Reference NUREG-1715 was NUREG/CR-6928 is used as added by RG 1.200 Revision 1; the source for generic data CC-1/11/111: CC-1/11/111: References NUREG-1715 and priors in Revision 5 of the ... . .. NUREG/CR-6928 were included VEGP internal events PRA. Examples of parameter estimates Examples of parameter estimates in the 2009 version of the PRA and associated sources include: and associated sources include Standard. (a) component failure rates and (a) component failure rates and probabilities: NUREG/CR-4639 [Note probabilities: NUREG/CR-4639 [2-7], The updated SR cites more (1)], NUREG/CR-4550 [Note (2)], NUREG/CR-4550 [2-3], NUREG- recent examples of acceptable NUREG-1715 [Note 71 1715 [2-21], NUREG/CR-6928 [2-20] generic data sources. QU-A2a: QU-A2: The LERF requirement was Section 10.3.2 of the internal added by RG 1.200 Revision 2. events PRA calculation CC-1/11/111: CC-1/11/111: presents estimates for PROVIDE estimates of the individual PROVIDE estimates of the individual The updated SR explicitly individual LERF sequence sequences in a manner consistent sequences in a manner consistent requires consideration of LERF. cutsets. with the estimation of total CDF ... with the estimation of total CDF (and LERF) ... E1-36

Enclosure 1 Attachment 2 - Comparison of RG 1.200 Revision 1 and Revision 2 Supporting Requirements SR in 2007 PRA Standard as SR in 2009 PRA Standard as Description of Change Resolution Amended by RG 1.200, Revision 1 Amended by RG 1.200, Revision 2 aU-A2b: aU-A3: The phrase, "from internal The peer review based on the events", was deleted from the 2007 version of the PRA CC-I: CC-I: 2009 version of the PRA Standard addressed these ESTIMATE the point estimate CDF ESTIMATE the point estimate CDF Standard. The LERF LERF requirements. Section from internal events. (and LERF). requirement was added by RG 10.3.2 of the internal events 1.200 Revision 2. PRA calculation presents the CC-II: mean CDF LERF results. ESTIMATE the mean CDF from CC-II: The SR explicitly requires internal events, accounting for the ESTIMATE the mean CDF (and consideration of LERF. "state-of-knowledge" correlation LERF), accounting for the state-of- However, per the note in 2007 between event probabilities [Note knowledge correlation between SR LE-E4 and LE-F3, LERF was (1 )]. event probabilities [Note (1)]. addressed in applicable requirements of Table 4.5.8, CC-Ill: which includes all au SRs. CALCULATE the mean CDF from CC-Ill: Thus, the peer review using the internal events by propagating the CALCULATE the mean CDF (and 2007 version of the PRA uncertainty distributions, ensuring LERF) by propagating the Standard addressed these LERF that the "state-of-knowledge" uncertainty distributions, ensuring requirements. correlation between event that the state-of-knowledge probabilities is taken into account. correlation between event probabilities is taken into account. aU-B6: aU-B6: The LERF requirement was The peer review based on the added by RG 1.200 Revision 2. 2007 version of the PRA CC-1111/111: CC-1/111111: Standard addressed these ACCOUNT for system successes in ACCOUNT for system successes in The SR explicitly requires LERF requirements. The addition to system failures in the addition to system failures in the consideration of LERF. Level 2 PRA event trees evaluation of accident sequences to evaluation of accident sequences to However, per the note in 2007 presented in Section 9.2 of the the extent needed for realistic the extent needed for realistic SR LE-E4 and LE-F3, LERF was internal events PRA calculation estimation of CDF. This accounting estimation of CDF or LERF. This addressed in applicable explicitly account for system may be accomplished by using accounting may be accomplished by requirements of Table 4.5.8, successes. numerical quantification of success using numerical quantification of which includes all au SRs. probability, complementary logic, or a success probability, complementary Thus, the peer review using the delete term approximation and logic, or a delete .term approximation 2007 version of the PRA includes the treatment of transfers and includes the treatment of Standard addressed these LERF among event trees where the transfers among event trees where requirements. "successes" may not be transferred the "successes" may not be between event trees. transferred between event trees. E1-37

Enclosure 1 Attachment 2 - Comparison of RG 1.200 Revision 1 and Revision 2 Supporting Requirements SR in 2007 PRA Standard as SR in 2009 PRA Standard as Description of Change Resolution Amended by RG 1.200, Revision 1 Amended by RG 1.200, Revision 2 QU-E3: QU-E3: The LERF ~equirement was The peer review based on the added by RG 1.200 Revision 2. 2007 version of the PRA CC-I: CC-I: Standard addressed these ESTIMATE the uncertainty interval of ESTIMATE the uncertainty interval of The SR explicitly requires LERF requirements. Section the CDF results. Provide a basis for the CDF (and LERF) results. consideration of LERF. 10.4 of the internal events PRA the estimate consistent with the Provide a basis for the estimate However, per the Note in 2007 calculation presents the characterization parameter consistent with the SR LE-E4 and LE-F3, LERF was uncertainty intervals for both uncertainties (DA-D3, HR-D6, HR- characterization parameter addressed in applicable CDF and LERF, with GB, IE-C15). uncertainties (DA-D3, HR-06, HR- requirements of Table 4.5.8, consideration of the state-of-GB, IE-C15). which includes all QU SRs. knowledge correlation. CC-II: Thus, the peer review using the ESTIMATE the uncertainty interval of CC-II: 2007 version of the PRA the CDF results. ESTIMATE the ESTIMATE the uncertainty interval of Standard addressed these LERF uncertainty intervals associated the CDF (and LERF) results. requirements. with parameter uncertainties (DA- ESTIMATE the uncertainty D3, HR-D6, HR-GB, IE-C15), taking intervals associated with into account the state-of- parameter uncertainties (DA-D3, knowledge correlation. HR-D6, HR-GB, IE-C15), taking into account the state-of-knowledge CC-Ill: correlation. PROPAGATE parameter uncertainties (DA-D3, HR-D6, HR- CC-111: GB, IE-C15) .... (no change) PROPAGATE parameter uncertainties (OA-03, HR-D6, HR-GB, IE-C15) .... (no change) E1-38

Enclosure 1 Attachment 2 - Comparison of RG 1.200 Revision 1 and Revision 2 Supporting Requirements SR in 2007 PRA Standard as SR in 2009 PRA Standard as Description of Change Resolution Amended by RG 1.200, Revision 1 Amended by RG 1.200, Revision 2 QU-E4: QU-E4: Separate requirements for CC-I, No action, CC-II met for 2007 II, and Ill were collapsed into a version of the PRA Standard. CC-I: CC-1/11/111: single requirement for CC-1/11/111 PROVIDE an assessment of the For each source of model in the 2009 version of the PRA impact of the model uncertainties uncertainty and related assumption Standard. The reference to Note and assumptions on the results of identified in QU-E1 and QU-E2, 1 was deleted by RG 1.200 the PRA. respectively, IDENTIFY how the Revision 2. PRA model is affected (e.g., CC-II: introduction of a new basic event, The updated SR assigns the EVALUATE the sensitivity of the changes to basic event probabilities, same requirement to all three results to model uncertainties and change in success criterion, CCs. Meeting CC-II: in the 2007 key assumptions using sensitivity introduction of a new initiating version of the PRA Standard analyses [Note (1)]. event). assures that the new SR is met. CC-Ill: EVALUATE the sensitivity of the results to uncertain model boundary conditions and other assumptions using sensitivity analyses except where such sources of uncertainty have been adequately treated in the quantitative uncertainty analysis [Note (1 )]. E1-39

Enclosure 1 Attachment 2 - Comparison of RG 1.200 Revision 1 and Revision 2 Supporting Requirements SR in 2007 PRA Standard as SR in 2009 PRA Standard as Description of Change Resolution Amended by RG 1.200, Revision 1 Amended by RG 1.200, Revision 2 LE-F2: LE-F3: Separate requirements for CC-I, No action, CC-II met for 2007 11, and Ill were collapsed into a version of the PRA Standard. CC-I: CC-1/11/111: single requirement for CC-1/11/111 PROVIDE a qualitative assessment IDENTIFY and CHARACTERIZE the in the 2009 version of the PRA of the key sources of uncertainty. LERF sources of model uncertainty Standard. Examples: and related assumptions, in a (a) Identify bounding manner consistent with the The updated SR assigns the assumptions. applicable requirements of Tables 2- same requirement to all three (b) Identify conservative treatment 2.7-2(d) and 2-2.7-2(e). CCs. Meeting CC-II: in the 2007 of phenomena. version of the PRA Standard assures that the new SR is met. CC-II: PROVIDE uncertainty analysis that identifies the key sources of uncertainty and includes sensitivity studies for the significant contributors to LERF. CC-Ill: PROVIDE uncertainty analysis that identifies the key sources of uncertainty and includes sensitivity studies. E1-40

Enclosure 1 Attachment 2 - Comparison of RG 1.200 Revision 1 and Revision 2 Supporting Requirements SR in 2007 PRA Standard as SR in 2009 PRA Standard as Description of Change Resolution Amended by RG 1.200, Revision 1 Amended by RG 1.200, Revision 2 IF-F2: IFPP-82: The requirement to document Section 5 and Appendix A of walkdowns performed in support the internal flooding PRA CC-11111111: CC-11111111: of plant partitioning was added document the walkdowns DOCUMENT the process used to DOCUMENT the process used to to the 2009 version of the PRA performed to validate identify ... flood areas, ... For identify flood areas. For example, Standard. information related to flood example, this documentation typically this documentation typically includes areas, flood sources, SSCs, includes , The updated SR cites examples mitigation and other flood ... of acceptable documentation of related features in the flood (b) flood areas used in the analysis (a) flood areas used in the analysis the process to identify flood areas. and the reason for eliminating areas and the reason for eliminating areas sources. from further analysis from further analysis ... (b) any walkdowns performed in Since documentation of support of the plant partitioning walkdowns was not in the 2007 version of the PRA Standard, it was not reviewed as part of the peer review conducted using that version of the PRA Standard. E1-41

Enclosure 1 Attachment 2 - Comparison of RG 1.200 Revision 1 and Revision 2 Supporting Requirements SR in 2007 PRA Standard as SR in 2009 PRA Standard as Description of Change Resolution Amended by RG 1.200, Revision 1 Amended by RG 1.200, Revision 2 IF-81: IFSO-A1: The requirement to include the Potential flood sources fire protection system in Item (a) identified in Section 5 of the CC-1111/111: CC-11111111: as a potential flooding source internal flooding PRA reviewed For each flood area, IDENTIFY the For each flood area, IDENTIFY the was added by RG 1.200 as part of 2009 peer review potential sources of flooding [Note potential sources of flooding [Note Revision 1. This requirement against 2007 version of the (1)]. INCLUDE: (1)]. INCLUDE was addressed in the peer PRA standard amended by RG (a) equipment (e.g., piping, valves, (a) equipment (e.g., piping, valves, review, which used the 2007 1.200, Revision 1 include pumps) located in the area that are pumps) located in the area that are version of the PRA Standard RCS-connected systems - connected to fluid systems (e.g., connected to fluid systems (e.g., amended by RG 1.200 Revision chemical and volume control circulating water system, service circulating water system, service 1. system (CVCS), containment water system, fire protection system, water system, fire protection system, spray (CS), residual heat component cooling water system, component cooling water system, The requirement to include the removal (RHR), reactor coolant feedwater system, condensate and feedwater system, condensate and reactor coolant system in Item system drain tank (RCSDT), steam systems) steam systems, and reactor coolant (a) as a potential flooding source safety injection (SI), and ... system) was added to the 2009 version reactor water makeup system

                                     ...                                    of the PRA Standard. Thus, it      (RMWS). The Containment was not reviewed as part of the    Building (and RCS peer review conducted using        components therein) is not that version of the PRA            included in the scope of the Standard.                          internal flooding analysis.

E1-42

Enclosure 1 Attachment 2 - Comparison of RG 1.200 Revision 1 and Revision 2 Supporting Requirements SR in 2007 PRA Standard as SR in 2009 PRA Standard as Description of Change Resolution Amended by RG 1.200, Revision 1 Amended by RG 1.200, Revision 2 IF-F2: IFS0-82: The requirement to document Section 5 and Appendix A of walkdowns performed in support the internal flooding PRA CC-1/11/111: CC-11111111: of the identification or screening document the walkdowns DOCUMENT the process used to DOCUMENT the process used to of flood sources was added to performed to validate identify applicable flood sources. For identify applicable flood sources. For 2009 version of the PRA information related to flood example, this documentation typically example, this documentation Standard. areas, flood sources, SSCs, includes typically includes mitigation and other flood (a) flood sources identified in the (a) flood sources identified in the The updated SR cites examples related features in the flood analysis, rules used to screen out analysis, rules used to screen out of acceptable documentation of areas. these sources, and the resulting list these sources, and the resulting list the process to identify flood of sources to be further examined of sources to be further examined sources . (f) screening criteria used in the Since documentation of analysis (b)screening criteria used in the walkdowns was not in the 2007 ... analysis version of the PRA Standard, it (j) calculations or other analyses was not reviewed as part of the used to support or refine the flooding (c) calculations or other analyses peer review conducted using evaluation used to support or refine the flooding that version of the PRA ... evaluation Standard. (d) any walkdowns performed in support of the identification or screeninQ of flood sources E1-43

Enclosure 1 Attachment 2 - Comparison of RG 1.200 Revision 1 and Revision 2 Supporting Requirements SR in 2007 PRA Standard as SR in 2009 PRA Standard as Description of Change Resolution Amended by RG 1.200, Revision 1 Amended by RG 1.200, Revision 2 IF-F2: IFSN-82: The requirement to document Section 5 and Appendix A of walkdowns performed in support the internal flooding PRA CC-1/11/111: CC-1/11/ll I: of the identification or screening document the walkdowns DOCUMENT the process used to DOCUMENT the process used to of flood scenarios was added to performed to validate identify applicable flood scenarios. identify applicable flood scenarios. 2009 version of the PRA information related to flood For example, this documentation For example, this documentation Standard. areas, flood sources, SSCs, typically includes typically includes mitigation and other flood ... The updated SR cites examples related features in the flood (c) propagation pathways ... (a) propagation pathways ... of acceptable documentation of areas. ... the process to identify flood (d) accident mitigating features and (b) accident mitigating features and scenarios. barriers credited ... barriers credited ... ... Since documentation of (e) assumptions or calculations used (c) assumptions or calculations used walkdowns was not in the 2007 in the determination of ... flood- in the determination of ... flood- version of the PRA Standard, it induced effects on equipment induced effects on equipment was not reviewed as part of the operability operability peer review conducted using ... that version of the PRA (f) screening criteria used in the (d) screening criteria used in the Standard. analysis analysis (g) flooding scenarios considered, (e) flooding scenarios considered, screened, and retained screened, and retained (h) description of how the internal (f) description of how the internal event analysis models were modified event analysis models were modified (j) calculations or other analyses (g) calculations or other analyses used to support or refine the flooding used to support or refine the flooding evaluation evaluation ... (h) any walkdowns performed in support of the identification or screening of flood scenarios E1-44

Enclosure 1 Attachment 2 - Comparison of RG 1.200 Revision 1 and Revision 2 Supporting Requirements SR in 2007 PRA Standard as SR in 2009 PRA Standard as Description of Change Resolution Amended by RG 1.200, Revision 1 Amended by RG 1.200, Revision 2 IF-F2: IFQU-82: The requirement to document Section 5 and Appendix A of walkdowns performed in support the internal flooding PRA CC-1/11/111: CC-1/111111: of internal flood accident document the walkdowns DOCUMENT the process used to DOCUMENT the process used to sequence quantification was performed to validate define the applicable internal flood define the applicable internal flood added in 2009 version of the information related to flood accident sequences and their accident sequences and their PRA Standard. areas, flood sources, SSCs, associated quantification. For associated quantification. For mitigation and other flood example, this documentation typically example, this documentation The updated SR cites examples related features in the flood includes typically includes of acceptable documentation of areas that are considered in ... (a) calculations or other analyses the process to identify flood flood sequence definition. (j) calculations or other analyses used to support or refine the flooding related features considered in used to support or refine the flooding evaluation flood sequence quantification. evaluation ... (b) screening criteria used in the Since documentation of (f) screening criteria used in the analysis walkdowns was not in the 2007 analysis version of the PRA Standard, it ... was not reviewed as part of the (i) flooding scenarios considered, (c) flooding scenarios considered, peer review conducted using screened, and retained screened, and retained that version of the PRA ... Standard . (k) results of the internal flood (d) results of the internal flood analysis, consistent with the analysis, consistent with the quantification requirements provided quantification requirements provided in HLR-QU-D in HLR-QU-D (e) any walkdowns performed in support of internal flood accident sequence quantification E1-45

Enclosure 1 Introduction and Overall Summary Vogtle Electric Generating Plant Supplemental Response to NRC Generic Letter 2004-02 Enclosure 1 Introduction and Overall Summary Attachment 3 Resolution of the VEGP Seismic PRA Peer Review Findings E1-46

Enclosure 1 Attachment 3 - Resolution of the VEGP Seismic PRA Peer Review Findings F&O Review cc F&O F&O F&O Disposition for GSl-191 Number Element Description Basis Resolution and Other Applications 11-3 SHA-E2 111111 While variability in To maintain hazard-consistent Expand documentation to There is an abundance of the mean base-case ground motion hazard at the demonstrate that a single site-specific Vs data from Vs profile is control point, the site response base-case Vs profile VEGP Units 3&4, which incorporated in the analysis needs to incorporate adequately represents reduces epistemic site response appropriate epistemic uncertainty the Units 1&2 site. Or if uncertainty to an analysis, no and aleatory variability in its that is not the case, insignificant level. epistemic inputs. The Vs profile for the include epistemic Additional discussion of uncertainty in the Vogtle Units 1&2 site is uncertainty in the the rationale for use of a base-case profile is represented by a single Vs profile, characterization of Vs single base-case Vs profile represented. indicating there is no epistemic profile and evaluate the for the site has been Documentation of uncertainty in the mean base- impact on control point included in the the justification for case profile. Documentation of ground motions. documentation. The added this assessment this assessment needs to be discussion demonstrates should be expanded. that a single base-case expanded. shear-wave velocity (Vs) Discussion with staff indicates profile adequately (This F&O originated that consideration of the represents the Vogtle site, from SR SHA-E2) combined data for the Vogtle site based on the availability of (Units 1&2, Units 3&4, ISFSI) Vs data, which reduces the provides sufficient confidence that epistemic uncertainty for a single mean base-case profile this particular parameter. characterizes the site. This This finding has been conclusion is based on the resolved with no significant quantity and quality of the impact to the SPRA results combined data and an evaluation or conclusions. showing the site is relatively uniform with respect to Vs. For some depth ranges, data from the nearby Savannah River Site (SRS) are used to support the profile interpretation. The documentation presents summaries of velocity data, but does not provide sufficient information to suooort the lack of E1-47

Enclosure 1 Attachment 3 - Resolution of the VEGP Seismic PRA Peer Review Findings F&O Review cc F&O F&O F&O Disposition for GSl-191 Number Element Description Basis Resolution and Other Applications I epistemic uncertainty at the Units 1&2 site over the complete depth range of the Vs profile. This would typically require multiple measurements throughout the depth range that provide a consistent picture of natural variability about a single mean base-case profile. The technical basis and justification that a single base-case profile is appropriate should be provided in more detail. This should include the basis for applying conclusions from other Vogtle locations to the Units 1&2 site. [A related Suggestion 11-2 addresses specifically potential epistemic uncertainty in the Blue Bluff Marl stratum.] 11-8 SHA-E2 11/111 Upper crustal site The Vogtle Site Response Provide a basis in the A discussion of the range attenuation of Analysis notes that the damping. documentation for of possible values of deep ground motion associated with the base-case representing base-case soil damping has been (kappa) is, profile corresponds to a total kappa at the site by a included in the generally, an kappa value for the soil column of single value. The basis documentation. uncertain parameter. 0.01 sec. The report does not might include sensitivity A sensitivity study on the Thus, to maintain address epistemic uncertainty in analyses to show the epistemic uncertainty of hazard-consistent kappa. impact of epistemic deep soil damping has ground motion at the uncertainty in kappa. been performed using control point, this In discussion with staff during the median, lower range, and uncertainty should peer review, it was noted that upper range alternatives be incorporated in randomization of the damping for deep rock damping. the site response associated with the profile layers Site response analysis analysis, or the represents both random variability was performed using 1E-4 basis for not and epistemic uncertainty. It was HF and LF rock input includinq it should also noted that kappa was motion. The resultinq E1-48

Enclosure 1 Attachment 3 - Resolution of the VEGP Seismic PRA Peer Review Findings F&O Review cc F&O F&O F&O Disposition for GSl-191 Number Element Description Basis Resolution and Other Applications be provided. In expected to be small for the amplification functions and either case, the Vogtle site and uncertainties in log-standard deviation technical basis and that small value would not be were weight-averaged and justification should expected to have a significant compared to the original be documented. impact on site amplification. Staff base case for each of BBM also noted that the approach High Pl and BBM Low Pl used had been reviewed by the soil columns. It was NRG for the Vogtle ESP and concluded that the (This F&O originated COLA. inclusion of alternative from SR SHA-E2) base cases for deep soil The SPID provides guidance damping to account accepted by the NRG for explicitly for the epistemic response to NTTF 2.1 uncertainty associated with Recommendation: Seismic that site kappa does not have indicates kappa is difficult to any significant effects on measure and thus subject to large the resulting seismic uncertainty (SPID Section B- hazard curves and UHRS. 5.1.3.2). The sensitivity study has been added to the SPRA Documentation of the technical documentation. basis for kappa characterization This finding has been should be expanded. resolved with no significant impact to the SPRA results or conclusions. 12-1 SHA-J1 Not As part of the PSHA The approach that was taken to Documentation should be A PSHA report has been Met implementation, the model earthquakes in the PSHA provided that describes prepared that describes analyst has different calculation was not identified. how seismic sources are how earthquake events alternatives for There are two basic alternatives modeled in the PSHA were modeled for area modeling the that can be used to model (i.e., how the SSC and sources in the PSHA earthquake earthquake events; as extended GMMs) were calculations. This was by occurrences in the fault ruptures, or as point sources. implemented in the modeling each earthquake calculations. The The approach that is used Vogtle PSHA. as a point source, and PSHA influences how the CEUS ground using correction factors for documentation does motion model is implemented. distance and ground not describe the motion uncertainty that approach that was No documentation is provided on modify the ground motion used to model either of these subjects estimate to include the E1-49

Enclosure 1 Attachment 3 - Resolution of the VEGP Seismic PRA Peer Review Findings F&O Review cc F&O F&O F&O Disposition for GSl-191 Number Element Description Basis Resolution and Other Applications earthquakes. (earthquake source modeling and effect of a closer distance use of the ground motion to a fault rupture (because attenuation models). From the rupture may be closer questions posed to the PSHA to the site than the single (This F&O originated analysts, it is our understanding point used to represent from SR SHA-J1) that earthquakes were modeled that event) and the as point sources and the uncertainty in ground appropriate ground motion motion because the aleatory uncertainty was used in azimuth of the rupture is the calculation. unknown. These correction factors were published by EPRI. This finding has been resolved with no significant impact to the SPRA results or conclusions. 12-11 SHA-J1 Not As part of the PSHA The PSHA analysts were asked Provide a description of A PSHA report has been Met implementation, the to describe the approach that was the earthquake modeling prepared that describes analyst has usedtomodelearthquakesinthe approach that was used how pseudo-faults were alternatives for Charleston RLME seismic source. to model the Charleston implemented to represent modeling the The response indicated that RLME seismic source the Charleston RLME earthquake earthquakes in the Charleston and how the approach source. This includes: 1. A occurrences in the RLME source were modeled was implemented. description of the pseudo-calculations. The using 'pseudo faults'. faults. 2. A definition of PSHA pseudo-faults as documentation does The PSHA report does not: constructed faults that not describe the 1. Describe that a 'pseudo fault' represent possible sources approach that was approach was used to model of future large used to model earthquakes in the Charleston earthquakes. 3. earthquakes in RLME source. Implementation of the RLME sources. 2. Provide a definition of 'pseudo pseudo-faults including faults'. spacing and limits at the

3. Describe how the 'pseudo fault' borders of the Charleston approach was implemented for source. 4. Documentation (This F&O originated the Charleston RLME seismic of the rupture area, length, from SR SHA-J1) source (e.g., what was the fault and width that were spacinQ that was used; how was estimated for possible E1-50

Enclosure 1 Attachment 3 - Resolution of the VEGP Seismic PRA Peer Review Findings F&O Review cc F&O F&O F&O Disposition for GSl-191 Number Element Description Basis Resolution and Other Applications the earthquake rate distributed to future earthquakes. 5. A the faults, etc.). description of how

4. Document the fault rupture earthquake ruptures are model that was used. distributed on the faults.

5. Describe how earthquake This finding has been events are distributed on the resolved with no significant faults. impact to the SPRA results or conclusions.

12-15 SHA-11, Not A screening A screening analysis was not A screening analysis for This evaluation was done SHA-12 Met assessment was performed for hazards such as other seismic hazards for the Vogtle 3&4 COLA performed for soil settlement, fault displacement, should be performed and and is noted in the ESP liquefaction and is tsunami, seiche, etc. documented as part of SAR. The Vogtle 3&4 described in seismic the PSHA and SPRA. evaluation is applicable to, fragility calculation. It is anticipated these other and has been cited in, the seismic hazards will be screened It is expected that Vogtle 1&2 SPRA Fragility A screening out. information in the FSAR report. assessment was not for Vogtle 1 & 2 and in performed for other the COLA for Units 3 & 4 This finding has been potential seismic can be used to support resolved with no significant hazards. this requirement. impact to the SPRA results or conclusions. (This F&O originated from SR SHA-11) 12-16 SHA-J1 Not The Vogtle PSHA The documentation of the PSHA- Prepare a complete and A PSHA report has been Met has gone through a is provided in a collection of up-to-date PSHA prepared that includes number of changes documents that were prepared in document that includes hazard results, and revisions since the 2012-2014 time frame. There all results, sensitivity uncertainties in hazard, 2012 due to does not exist a single document calculations, and sensitivities to input changes in models, that contains a set of results that deaggregation results, uncertainties; this input data, etc. As is based on the current PSHA etc. that is based on the summarizes hazard results new calculations model. current model. for the Vogtle site. were performed and This finding has been reports generated, resolved with no significant sensitivity results, impact to the SPRA results were not carried or conclusions. forward. As a result, there does not exist E1-51

Enclosure 1 Attachment 3 - Resolution of the VEGP Seismic PRA Peer Review Findings F&O Review cc F&O F&O F&O Disposition for GSl-191 Number Element Description Basis Resolution and Other Applications a current report that includes all PSHA results, deaggregations, etc. that is based on the current PSHA model. (This F&O originated from SR SHA-J1) 12-18 SHA-82, I/II The Vogtle PSHA is As part of a site-specific PSHA, A data gathering effort A detailed study of new SHA-C4, Not based on the CEUS there is a need to gather, review should be undertaken to geological, seismological, SHA-H1 Met SSC seismic source and evaluate new geological, identify new information and geophysical Not model which was seismological, or geophysical that post-dates the CEUS information was Met completed in 2012. information or information that is SSC data collection conducted, to determine if The SSC model was defined at a scale that was not effort. The data gathering any information a developed at a considered in the development of effort should also look for subsequent to the EPRI regional scale that the CEUS SSC model. As part of information local to the SSC model is available was based on data the Vogtle SPRA, no effort was Vogtle site region that that should be gathered up until made to gather up-to-date and was not considered, or at incorporated into the about 2010. (Note, local (local to the Vogtle site) a scale that was not seismic hazard results for the date when data information to evaluate whether addressed as part of the Vogtle. This study is was gathered any new information has become CEUS SSC regional described in the SPRA varied; for example available on active faulting and/or evaluation. documentation. While the the earthquake the development new seismic area around the site catalog was sources or the revision of sources Some of this information continues to be studied by complete through in the CEUS SSC model in the may be available in the many earth scientists, 2008.) In the sense vicinity of the Vogtle plant. COLA for Vogtle Units 3 there was no new that the CEUS SSC &4. information identified that model was not Since up-to-data was not would change the estimate specifically gathered, consideration of of seismic hazard for performed as a site- alternatives could not be Vogt le. specific PSHA for addressed. This finding has been the Vogtle site. resolved with no significant impact to the SPRA results (This F&O originated or conclusions. from SR SHA-82) E1-52

Enclosure 1 Attachment 3 - Resolution of the VEGP Seismic PRA Peer Review Findings F&O Review cc F&O F&O F&O Disposition for GSl-191 Number Element Description Basis Resolution and Other Applications 12-2 SHA-J1 Not The method that is For soil sites, the soil hazard is The documentation The methodology used for Met used in the Vogtle generally (though not exclusively, should include a the surface hazard PSHA to estimate since other methods could be description of the calculation has been the soil site hazard used) determined in two steps; methodology that is used described in detail, and a is not described or probabilistic rock hazard results to combine the rock comparison made between referenced. are estimated which are then hazard results and the the GMRS using the two combined with probabilistic site amplification factors approaches 2A and 3. estimates of the site response. to determine the soil Approach 2A was used for The method used in the Vogtle hazard at the Vogtle site. the calculation of SSI input (This F&O originated PSHA to estimate the soil hazard motions at foundation from SR SHA-J1) is not described. elevations and Approach 3 was used for the calculation of surface hazard and GMRS at the ground surface, as defined in NUREG/CR-6728. It was concluded that the use of Approach 2A USHRS as input to the SSI analysis of the Vogtle plant is considered acceptable and does not present any significant inconsistency with the seismic hazard curve and GMRS at the ground surface, which were calculated using Approach 3. This finding has been resolved with no significant impact to the SPRA results or conclusions. E1-53

Enclosure 1 Attachment 3 - Resolution of the VEGP Seismic PRA Peer Review Findings F&O Review cc F&O F&O F&O Disposition for GSl-191 Number Element Description Basis Resolution and Other Applications 12-22 SHA-E2 11/111 Versions 1 and 2 of The site response calculation A framework and A description of the the site response does not present a clear approach for evaluating methodology used to Calculation do not description of how aleatory and and modeling account for epistemic and describe a epistemic uncertainties are uncertainties in the site aleatory uncertainties in framework for identified and evaluated. As a response should be soil hazard has been evaluating and result it is difficult to track the developed and added to the characterizing propagation of uncertainties is implemented. The site documentation. sources of aleatory carried out in the site response response calculation This finding has been and epistemic analysis. documentation should resolved with no significant uncertainty and how fully describe the impact to the SPRA results the approach was It is worth noting that there is methodology and its or conclusions. implemented. some epistemic site response implementation. uncertainty that is accounted for (This F&O originated in the rock GMPEs. from SR SHA-E2) 12-23 SPR-E5 II The quantification The documentation presents the Develop and document Additional detail has been process has results of three different an understanding of the added to the SPRA included the uncertainty calculations for CDF earlier point estimate Quantification report to uncertainties in the and LERF. In addition, point results for CDF and LERF document the uncertainty, seismic hazard, estimates for CDF and LERF are and of uncertainty results. importance, and sensitivity fragility and calculated and reported in other analyses and relate the systems-analysis sections. Thus the documentation uncertainty analysis mean elements of the reports two estimates of the mean CDF and LERF to the SPRA. The results CDF and LERF respectively from point estimate values. presented are different uncertainty calculations internally and a 'Point Estimates' result for This finding has been inconsistent and are each. All of these results are resolved with no significant inconsistent with the different than the point estimate impact to the SPRA results results reported in (approximate mean) reported in or conclusions. other sections for other sections for CDF and LERF, CDF and LERF, respectively. The documentation . respectively. in the report does not describe the basis (inputs) for these (This F&O originated calculations, or offer an from SR SPR-E5) interpretation of the results. 12-24 SPR-E5 II The Quantification The uncertainty analysis is Provide documentation of Additional detail has been report does not presented with the results the uncertaintv analvsis added to the E1-54

Enclosure 1 Attachment 3 - Resolution of the VEGP Seismic PRA Peer Review Findings F&O Review cc F&O F&O F&O Disposition for GSl-191 Number Element Description Basis Resolution and Other Applications provide reported. The report provides that describes the results, documentation of the documentation of limited discussion of the results how they are being seismic plant response the uncertainty and the insights that might be interpreted and the model, model analysis results. gained from them. insights that are derived implementation, and from them. quantification in the QU The two sets of results that are report. In addition, the reported are not discussed in uncertainty, importance, (This F&O originated terms of their relationship to each and sensitivity analyses from SR SPR-ES) other. For instance the mean are described in more values should be the same (but detail. are not). The uncertainty estimates provide insight to the This finding has been total uncertainty and the resolved with no significant contribution of the basic event impact to the SPRA results uncertainty to the total. or conclusions. In addition, neither the table of results or the discussion identifies what is the 'final' uncertainty result that includes the propagation of uncertainties of all elements of the SPRA to the estimates of CDF and LERF. 12-26 SPR-ES II There are The report does not present the Document the results of Updated Monte Carlo differences in the results of sensitivity calculations sensitivity calculations on uncertainty runs have results for CDF and with regard to the number of the number of Monte been performed with LERF that are Monte Carlo simulations that are Carlo simulations 20,000 iterations for SCDF reported. A possible needed to produce stable results. required to produce and SLERF. This is a contributor to these stable results. sufficiently high number of differences may be It is our understanding from simulations to produce a due to the number of discussion with the PRA staff that stable result. The SPRA Monte Carlo these types of sensitivity documentation has been simulations that calculations were performed. updated to clearly were performed. indicates the results. (This F&O originated This finding has been from SR SPR-ES) resolved with no siqnificant E1-55

Enclosure 1 Attachment 3 - Resolution of the VEGP Seismic PRA Peer Review Findings F&O Review cc F&O F&O F&O Disposition for GSl-191 Number Element Description Basis Resolution and Other Applications impact to the SPRA results or conclusions. 12-27 SPR-F2 Met Documentation The current quantification Provide clear and The QU report should be provided document does not provide a complete documentation documentation has been that describes how clear description of the how the of the approach used to updated to describe the the plant model plant model is quantified. For quantify the seismic plant quantification process, analysis is example the discussion does not response model, to including the technique for quantified. identify how calculations are perform the risk combining cutsets over the performed, what the limitations of quantification, uncertainty 14 acceleration intervals, these quantifications are and how analysis, and importance and obtaining the they affect the results. analysis. importance measures. (This F&O originated from SR SPR-F2) This finding has been resolved with no significant impact to the SPRA results or conclusions. 12-29 SPR-E2 Met The Quantification There is limited documentation of Document the process Additional detail has been report provides the process and the numerical and methods that were added to the QU report to limited methods that were used to used to perform the document the uncertainty, documentation of perform the uncertainty analysis. uncertainty analysis. importance, and sensitivity the process and Based on the documentation that Where appropriate analyses and relate the methods that were is provided and discussions with document where uncertainty analysis mean used to perform the the PRA staff there is limited but consistencies and SCDF and SLERF to the uncertainty analysis. not complete understanding of potential inconsistencies point estimate values. the methods that were used and in results might be the relationship of these methods expected. This finding has been to the results were obtained. resolved with no significant (This F&O originated impact to the SPRA results from SR SPR-E2) In some cases (as described in or conclusions. the documentation) the results from the uncertainty analysis are not the same as the results reported in other sections of the documentation for CDF and LERF (though this connection is not E1-56

Enclosure 1 Attachment 3 - Resolution of the VEGP Seismic PRA Peer Review Findings F&O Review cc F&O F&O F&O Disposition for GSl-191 Number Element Description Basis Resolution and Other Applications clearly stated in the report). However, it would seem the results should be internally consistent. 12-31 SPR-F1 Not The standard There is limited documentation Documentation should be Additional detail has been Met requires a level of that describes the seismic plant provided in sufficient added to the documentation that response analysis and detail that describes the documentation of the provides an quantification; how the model was seismic plant model, how seismic plant response understanding of the implemented, how the it is implement and model, model seismic plant quantification was performed and quantified. implementation, and response model and a discussion of the analysis quantification in the QU the quantification. results. report. In addition, the This requirement is uncertainty, importance, not met. To meet this requirement, the and sensitivity analyses documentation must be in are described in more considerable detail in order to detail. support the review process and (This F&O originated future updates. Part of the This finding has been from SR SPR-F1) documentation should include a resolved with no significant detailed discussion of the results, impact to the SPRA results sensitivity calculations, and the or conclusions. uncertainty analysis. 12-32 SPR-F3 Met The documentation The purpose of this supporting Document and discuss The documentation of the of the sources of requirement is that documentation the contribution of the uncertainty analysis has model uncertainty should be presented that different sources of been expanded in the and a description of addresses the sources of uncertainty that are Quantification report. A the analysis epistemic (knowledge) uncertainty modeled in the SPRA. discussion of sources of assumptions is not that are modeled and their model uncertainty has complete in the contribution to the total been added to the report, SPRA quantification uncertainty in CDF and LERF. and potentially important report. In addition, sources have been there is not a clear In addition, the documentation addressed in the sensitivity description of the should discuss elements of the analysis. uncertainty analysis seismic plant model where there and the contributors may be latent sources of This finding has been to the total uncertainty that are not modeled resolved with no significant uncertainty bevond and assumptions that are made in impact to the SPRA results E1-57

Enclosure 1 Attachment 3 - Resolution of the VEGP Seismic PRA Peer Review Findings F&O Review cc F&O F&O F&O Disposition for GSl-191 Number Element Description Basis Resolution and Other Applications a simple report from performing the analysis. or conclusions. UN CERT. (This F&O originated from SR SPR-F3) 12-36 SHA-83, 1111, As part of a site- As part of the Vogtle PSHA an An up-to-date earthquake An update to the SHA-C4, Not specific PSHA, an effort was not made to gather catalog for the Vogtle site earthquake catalog was SHA-H1 Met up-to-date data on earthquakes that region should be prepared from the time of Not earthquake catalog occurred since 2008. As such, the developed to assess the CEUS SSC catalog Met should be used. The analysts did not assess whether whether modifications to (through 2008) through CEUS SSC study more recent seismicity is the seismic source February 2016. The rate involved the consistent with the recurrence parameters or of occurrence of development of a characterization parameters required. The updated earthquakes within 320 km comprehensive estimated as part of the CEUS catalog, resources used of the Vogtle site was earthquake catalog SSC study (NRC, 2012). in compiling the update compared to the rate of based on data and the results of the earthquakes represented through 2008. The We note that as part of the Vogtle evaluation should be by the CEUS SSC seismic Vogtle site-specific PSHA, calculations were documented as part of source model for that PSHA should performed to recompute the the PSHA. If more recent same area, this consider the impact seismic hazard at the site to take seismicity is not comparison being made SSC of any into account changes in the consistent with the for M>2.9. It was found additional seismicity CEUS SSC earthquake catalog existing CEUS SSC that the updated catalog since 2008 up to the through 2008 that were made seismic source implied a rate of time the study following the completion of the parameters, the earthquakes that is lower started. CEUS SSC study. These parameters should be than the mean rate from changes reflect the identification updated and the PSHA the CEUS SSC seismic (This F&O originated of reservoir induced seismicity should be updated. sources. Therefore, from SR SHA-C4) earthquakes and the re- incorporating the effects of interpretation of the location of a updated catalog on the some earthquakes in the hazard at Vogtle would Charleston, SC area that decrease the hazard occurred in the 1880's (EPRI, slightly, and was not 2014). undertaken. This comparison is documented References in the SPRA documentation. EPRI (2014). Review of EPRI This findinQ has been E1-58

Enclosure 1 Attachment 3 - Resolution of the VEGP Seismic PRA Peer Review Findings F&O Review cc F&O F&O F&O Disposition for GSl-191 Number Element Description Basis Resolution and Other Applications 1021097 Earthquake Catalog for resolved with no significant RIS Earthquakes in the impact to the SPRA results Southeastern U. S. and or conclusions. Earthquakes in South Carolina Near the Time of the 1886 Charleston Earthquake Sequence, transmitted by letter from J. Richards to R. McGuire on March 5, 2014. 12-8 SHA-J3 Not A foundational The documentation of the The resolution to this Sources of uncertainty in Met element of PSHA as sources of model uncertainty finding could involve: the seismic hazard it has evolved over analysis and a description of the analysis for Vogtle are the past 30 years is analysis assumptions is not 1. Documentation and discussed in the updated the development complete in the PSHA report in its discussion of the SPRA documentation. and implementation current form such that a clear contribution of different These include uncertainty of methods to understanding of the contribution sources of uncertainty in seismic source model identify, evaluate, of individual sources of that are modeled in the (for background and model sources uncertainty to the estimate of PSHA. The earthquake sources and of epistemic (model hazard are understood. Limited documentation of the for the Charleston RLME), and parametric). information on the contribution of contribution of different in maximum magnitude for uncertainty in the seismic sources to the total mean sources of uncertainty background seismic estimate of ground hazard is presented, but can be shown by means sources and for the motion hazards. As information on the contributors to of 'tornado plots' that Charleston RLME, in such fairly rigorous the uncertainty is not provided. quantify the sensitivity of ground motion prediction analyses are carried the hazard at different equation, in smoothing out (SSHAC studies) With respect to addressing model ground motion levels to assumptions for seismicity to quantitatively uncertainties and associated the various branches in parameters in background address model assumptions there are some the logic tree. These sources, and in site uncertainties. examples that can be identified in plots show which sources amplification model. the Vogtle PSHA. For example, in of epistemic uncertainty "Tornado plots" are At the same time the site response analysis the are most important. It included in the updated there is within any assumption is made that the 1D should include the source SPRA documentation that analysis sources of equivalent linear model (SHAKE model uncertainty, show the contribution to uncertainty that are type) to estimate the site ground motion model total uncertainty in seismic not directly modeled amplification and ground motion uncertainty, and site hazard from source model and assumptions input to plant structures is response uncertainty. uncertainty, maximum that are made for appropriate. Currently, the total magnitude uncertainty, E1-59

Enclosure 1 Attachment 3 - Resolution of the VEGP Seismic PRA Peer Review Findings F&O Review cc F&O F&O F&O Disposition for GSl-191 Number Element Description Basis Resolution and Other Applications pragmatic or other uncertainty is shown by ground motion prediction reasons. There are the hazard fractiles, but it equation uncertainty, also sources of is not broken down to smoothing assumptions for model uncertainty provide understanding as seismicity parameters in that are embedded to what is most important. background sources, and in the context of site response uncertainty. current practice that 2. Identification and These plots are presented are 'accepted' and discussion of model for 10 Hz and 1 Hz typically not subject assumptions that are spectral acceleration, for to critical review. For made. ground motion amplitudes instance, in the corresponding to mean PSHA it is standard annual frequencies of practice to assume exceedance of 1E-4 and that the temporal 1E-5. These "tornado occurrence of plots" show that ground earthquakes is motion prediction equation defined by a is the major contributor to Poisson process. seismic hazard uncertainty This assumption is for both 10 Hz and 1 hz well accepted spectral acceleration, and despite the fact that maximum magnitude of it violates certain the Charleston RLME fundamentally source is an important understanding of contributor for 1 Hz tectonic processes spectral acceleration. (strain The use of equivalent accumulation). A linear one-dimensional second practice is site response analysis, the fact that and its associated earthquake assumptions, and its aftershocks are not adequacy for the Vogtle modeled in the site are documented in the PSHA, even though hazard calculation. they may be This finding has been significant events resolved with no significant (depending on the impact to the SPRA results size of the main or conclusions. E1-60

Enclosure 1 Attachment 3 - Resolution of the VEGP Seismic PRA Peer Review Findings F&O Review cc F&O F&O F&O Disposition for GSl-191 Number Element Description Basis Resolution and other Applications event). In the spirit of the standard it seems appropriate that sources of model uncertainty that are modeled as well as sources of uncertainty and associated assumptions as they relate to the site-specific analysis should be identified/ discussed and their influence on the results discussed. As SPRA reviews and the use of the standard has evolved, it would seem the former interpretation is reasonable, but potentially incomplete. It is reasonable from the perspective that documentation of the sources of model uncertainty and their contribution to the site-specific hazard results is a valuable E1-61

Enclosure 1 Attachment 3 - Resolution of the VEGP Seismic PRA Peer Review Findings F&O Review cc F&O F&O F&O Disposition for GSl-191 Number Element Description Basis Resolution and Other Applications product that supports the peer review process and assessments in the future as new information becomes available). Similarly, documenting assumptions provides similar support for peer reviews and future updates. The notion that model uncertainties and related assumptions that are not addressed in the PSHA is at a certain level an extreme requirement that may not be readily met and may not be particularly supportive of the analysis that is performed. For purposes of this review, the following approach is taken with regard to this supporting requirement:

1. The E1-62

Enclosure 1 Attachment 3 - Resolution of the VEGP Seismic PRA Peer Review Findings F&O Review cc F&O F&O F&O Disposition for GSl-191 Number Element Description Basis Resolution and Other Applications documentation should present quantitative results and discussion the sources of epistemic uncertainty that are modeled and their contribution to the total uncertainty in the seismic hazard.

2. The documentation should discuss elements of the PSHA model where their may be latent sources of model uncertainty that are not modeled and assumptions that are made in performing the analysis.

(This F&O originated from SR SHA-J3) 14-1 SFR-A2 I The conservatisms SFR-A2 requires that seismic Account for conservatism Evaluation of anchorage that exist in fragilities be based on plant- in the building response has been updated to structural demand specific data and that they are analyses in the structure include clipping of in-were not properly realistic and median centered with response factor for structure response accounted for in the reasonable estimates of component fragility spectra, and the estimation of uncertainty. evaluations. methodology is component and documented in the fragility structure fragilities. The structural response factor Use clipped spectra for notebook. used in all component fragilities assessing anchorage Structure response is reviewed is reported as 1.0. This capacities. dominated by the soft soil factor will be greater than 1.0 on which Vogtle 1 and 2 (This F&O originated because of the conservatism structures are founded. E1-63

Enclosure 1 Attachment 3 - Resolution of the VEGP Seismic PRA Peer Review Findings F&O Review cc F&O F&O F&O Disposition for GSl-191 Number Element Description Basis Resolution and Other Applications from SR SFR-A2) . introduced in the demand through This would cause higher the structural analysis. Because damping at lower hazard of this, the component and frequency levels and lead structural fragilities are biased to stress similar to the low. stress calculated for the buildings at 1E-4. As a The fragilities developed for result the structural structures and components that response factor is close to are mounted in those structures 1 and is accounted for will be biased low because the appropriately in the fragility input structural demands include evaluations. conservatisms. Time histories The input motion at the used for the SSI analysis have control point in the SSI been processed such that each analysis has been record envelopes the target modified to reasonably UHRS. This will introduce some match the corresponding level of conservatism. The input 1E-4 UHRS from the site-motion at the control point has consistent input motion been scaled to produce resultant analysis. FIRS that envelopes the FIRS This finding has been coming out of the site-consistent resolved. input motion analysis. In structure response spectra coming out of the SSI analyses were not peak clipped when computing anchorage demands. Structure response at the calculated equipment fragility levels is considerably higher than the 1E-4 UHRS considered in the building response analyses. The structure will have additional cracked shear walls and higher associated levels of damping at these higher ground motions. 14-10 SFR-A2 I Significant In the fragility calculations of heat Realistic nozzle loads The CCW and ACCW heat conservatisms were exchangers (PRA-BC-V-14-009 should be determined for exchanger capacities have E1-64

Enclosure 1 Attachment 3 - Resolution of the VEGP Seismic PRA Peer Review Findings F&O Review cc F&O F&O F&O Disposition for GSl-191 Number Element Description Basis Resolution and Other Applications noted in several Appendix A}, nozzle loads fragility evaluation of heat been updated to reflect sampled fragility significantly contribute to the exchangers. realistic nozzle loads. The calculations. seismic demands which form the equipment fragilities have basis for the median capacities. been updated to account Based on in-plant walkdowns by for appropriate frequency, the peer review teams and also The equipment capacity and uncertainty has been (This F&O originated noted in the walkdown report, the factor should be based on considered in these from SR SFR-A2) piping is well supported in all the frequency range of updates. directions and will not impose interest. That frequency This finding has been significant nozzle loads during a range of interest is resolved. seismic event. The CCW and centered at the ACCW capacities are below the fundamental frequency of 2.Sg screening level and are the pump, and considers significant contributors to risk so some uncertainty in that more realistic fragilities are frequency. required. Battery rack 11806B3BN3 in calculation PRA-BC-V-14-010 Appendix J2 is governed by GERS capacity. The GERS capacity is taken to be 1g, which corresponds to a frequency of 1 Hz. This is not realistic. The actual capacity is about 4g. The median capacity reported in the calculation is well below the 2.Sg screening level and is not realistic. The median capacity reported for the Turbine Driven Auxiliary Feedwater Pump is reported in Calculation PRA-BC-V-14-008 as 1.56g. This fragility is based on the seismic qualification document. The frequency range E1-65

Enclosure 1 Attachment 3 - Resolution of the VEGP Seismic PRA Peer Review Findings F&O Review cc F&O F&O F&O Disposition for GSl-191 Number Element Description Basis Resolution and Other Applications of interest for the fragility evaluation should be centered around the fundamental frequency of the assembly and not consider the entire frequency ranQe. 14-14 SFR-G2 Met The iterative In review of the seismic fragility Add a description of the The description of the process used for calculation for the safety features iterative process for iterative process for developing realistic sequencer (11821U3001), it was computing the component computing fragilities has fragilities is not well discovered that an iterative fragilities in the SPRA been documented. documented. process was used. The initial documentation fragility is based on EPRI 6041 This finding has been screening methodology and an resolved with no significant equipment capacity factor that is impact to the SPRA results (This F&O originated equal to the EPRI 6041 median or conclusions. from SR SFR-G2) capacity divided by the peak in structure demand. If this value is less than the screening capacity (2.5g), then the fragility may be refined by examining the component fundamental frequency. The fragility may be further refined by examining component specific qualification test reports. However, the fragility used in the logic tree by the systems analyst is generally the highest of these computed. This is reasonable and appropriate, however, this process is not described in the fragility notebook or fragility calculations. 14-17 SFR-02 Met Inconsistencies and Fragilities for the Vogtle 1&2 Update SNC calculation The following changes errors in NSSS Nuclear Steam Supply System no. PRA-BC-V-14-015 to have been made: NSSS fragility (NSSS) are based on the results incorporate corrections fragility calculations have development. of the WestinQhouse analysis of and enhancements. been updated to reflect E1-66

Enclosure 1 Attachment 3 - Resolution of the VEGP Seismic PRA Peer Review Findings F&O Review cc F&O F&O F&O Disposition for GSl-191 Number Element Description Basis Resolution and Other Applications record (AOR) associated with the Westinghouse-provided safe shutdown earthquake (SSE). critical loads and support In general, fragilities are capacities represented in (This F&O originated developed through scaling of the the critical failure modes; from SR SFR-02) SSE demands to the RLE and the effect of inelastic using the AOR seismic margins. energy absorption is Various deficiencies were noted factored in and in the development of the documented in fragility fragilities associated with these calculation as appropriate; components. the Reactor Coolant Pump Basis: The NSSS Seismic fragility has been updated fragility evaluation (SNC to reflect the failure of the calculation no. PRA-BC-V pump associated with 015) includes detail calculations LOCA; the reactor for each of the major NSSS internals fragility has been components. It indicates that the updated in the calculation; critical failure modes for the and the new fragilities components are controlled by the have been reflected in the support capacities. updated SPRA model. During the Peer Review, the team members discussed these issues This finding has been with SNC staff to obtain insights resolved. and develop potential resolution paths. Key issues included: (a) Basis for assumption that the support capacities represented the critical failure mode was not documented. SNC indicated that this was based on input from Westinghouse and NUREG-3360 and will update the fragility evaluation of provide this information. (b) Inelastic energy absorption was not credited to increase the median capacities - this does not result in realistic median E1-67

Enclosure 1 Attachment 3 - Resolution of the VEGP Seismic PRA Peer Review Findings F&O Review cc F&O F&O F&O Disposition for GSl-191 Number Element Description Basis Resolution and Other Applications capacities (overly conservative). (c) Reactor Coolant Pump fragility was based on consideration of the failure of the attached CCW piping, due to an assumption that a small-break/RCP seal LOCA was critical. It was learned during the Peer Review that failure in the system model was linked to a large-break LOCA, so the failure mode considered in the fragility evaluation is not consistent with the system model - SNC indicated that they will revise the fragility evaluation. (d) Reactor Internal fragility evaluation determined the demand based an average spectral acceleration over the range of 2 to 3 Hz, rather than using the peak acceleration in this range of the ISRS, and did not consider the contribution of higher modes. SNC indicated that this was done to avoid an overly conservative capacity, but agreed that the contribution of higher modes should be addressed, and will revise the calculation. (f) Control Rod Drive Mechanism fragility evaluation assumed that material stresses were the critical failure mode, and did not address the potential impact of deflections on rod drop. SNC indicated that information provided by WestinQhouse (based on a E1-68

Enclosure 1 Attachment 3 - Resolution of the VEGP Seismic PRA Peer Review Findings F&O Review cc F&O F&O F&O Disposition for GSl-191 Number Element Description Basis Resolution and Other Applications Japanese testing program) indicated that the deflection levels associated with seismic loading does not impact rod drop, and agree to add this discussion to the calculation. 14-20 SFR-E4, Met Seismic induced fire The only mention for seismic Seismic induced fire is an The seismic-induced fire SPR-89 Met evaluations are not induced fire evaluation is important element of the and flood evaluations have documented in the contained in the quantification fragility evaluation been updated, and walkdown report or notebook. Based on discussions process and this should documented in the fragility fragility calculations. during the peer review, it is be clearly documented. and quantification report. understood that seismic induced This includes the details of fire was a key consideration the walkdown procedure during the walkdowns. However, used to evaluate the (This F&O originated detail of the walkdown procedure potential for seismically from SR SFR-E4) for fire following earthquake is induced fires, including the missing. The write up should methodology, screening include team composition, criteria and results. methodology, screening criteria, and results, This finding has been resolved with no significant impact to the SPRA results or conclusions. 14-4 SFR-01 Met A potential for SFR-01 requires that realistic Evaluate the potential for The evaluation for sloshing induced failure modes of structures and flood induced failure of potential flood induced inundation of the equipment that interfere with the the NSCW Pumps or failure of the NSCW NSCW Pumps operation of that equipment be NSCW discharge MOVs. pumps or the NSCW (11202P4007, identified. discharge MOVs has been 11202P408) and performed and associated The potential for earthquake documented in the fragility discharge motor induced sloshing of the water calculation for the NSCW operated valves within the NSCW tower exists. tower. There was no (1HV11600, 11606, From field walkdowns of the significant impact on the 11607, 11613) inthe NSCW it was observed that there pump or MOV fragilities. NSCW exists and is a potential for sloshing of was not identified contents to potentially splash onto This finding has been either in the or flood the pumps and or motor resolved with no significant E1-69

Enclosure 1 Attachment 3 - Resolution of the VEGP Seismic PRA Peer Review Findings F&O Review cc F&O F&O F&O Disposition for GSl-191 Number Element Description Basis Resolution and Other Applications walkdowns or operated valves on the attached impact to the SPRA results subsequent discharge piping. or conclusions. analysis. (This F&O originated from SR SFR-D1) 14-5 SFR-D1 Met The potential for Vogtle 1&2 is a soil site, with Develop estimates of the Documentation has been seismically-induced engineered fill from the rock differential settlements updated to include the differential interface to the finished grade. between adjacent effects of earthquake settlements between The in-scope Seismic Category I structures and assess the induced settlement; no structures was not structures have foundations with fragility of commodities significant differential addressed. varying embedment depths, based on their ability to settlements were ranging from surface founded accommodate the computed between the (elev. 220 ft.) to a foundation associated differential structures. embedment of 110 ft. (elev. 110 displacements. This finding has been (This F&O originated ft.). Since soils, including resolved with no significant from SR SFR-D1) engineered fill, will impact to the SPRA results consolidate/settle to some extent or conclusions. when subjected to high level earthquake ground motion, and the amount of settlement is proportional to the thickness of the soil layer under the foundation, the settlement of one structure relative to another structure is dependent on the depth of the foundation embedment. The Fragility Notebook (PRA-BC-V-14-025) does not address the potential differential settlement between buildings, or the potential effect on commodities (e.g., piping, electrical raceways, HVAC ducts, etc.) that cross the separation between adjacent E1-70

Enclosure 1 Attachment 3 - Resolution of the VEGP Seismic PRA Peer Review Findings F&O Review cc F&O F&O F&O Disposition for GSl-191 Number Element Description Basis Resolution and Other Applications structures. During the performance of the Peer Review, SNC personnel indicated that the consideration of differential settlements was not required, since the structures were founded on engineered fill. 14-6 SFR-G2 Met The results of the The walkdown guidance provided Provide documentation of As noted in the Finding seismic gap/shake in Appendix F (Checklists and the results of the seismic basis, inspection of the space walkdowns Walkdown Data Sheets) of EPRI gap walkdowns. seismic gaps was included are not documented. NP-6041 includes attributes of in the seismic walkdowns. seismic gaps between structures Piping across seismic which should be addressed in the gaps is designed with performance of the walkdowns. adequate flexibility to (This F&O originated These include the clearance accommodate building from SR SFR-G2) between adjacent structures and motions, and pipe sleeves the ability of any subsystems provide adequate gaps for (e.g., piping, cable trays, HVAC piping movement. The ducts) spanning the gap to documentation has been accommodate the differential updated to reflect the seismic displacements. inspections performed during the walkdowns. The Seismic Walkdown Report This finding has been (PRA-BC-V-14-005) does not resolved with no significant include documentation of the impact to the SPRA results results/findings/observations or conclusions. associated with the inspection of the seismic gaps between structures or the subsystems spanning the gap. During the performance of the Peer Review, SNC personal indicated that inspection of the seismic gaps was included in the seismic walkdowns, but not explicitly described in the report. The ability of components to E1-71

Enclosure 1 Attachment 3 - Resolution of the VEGP Seismic PRA Peer Review Findings F&O Review cc F&O F&O F&O Disposition for GSl-191 Number Element Description Basis Resolution and Other Applications accommodate potential differential movement at the building separations is implied in the discussion of rugged components (piping, cable trays, and HVAC ducts) in the section on Rationale for Screening in the report. In addition, information from the Vogtle IPEEE Report indicated that the seismic gaps had been inspected during the IPEEE. 14-7 SFR-A2, I, The fragility The determination of the Update the fragility The fragility evaluation of SFR-F4 Met evaluation for the fundamental frequency of evaluation for the polar the polar crane has been Containment Polar structures and components crane to address potential updated to address Crane (in fragility involves a certain degree of uncertainty in the potential uncertainty in the notebook) did not uncertainty. This uncertainty fundamental frequency fundamental frequency address the impact must be accounted for in the and the contribution of and contribution of higher of variation in the determination of the seismic higher modes. modes. fundamental accelerations from the applicable This finding has been frequency on the in-structure response spectra resolved. applicable seismic (ISRS). demand. The section of the Polar Crane of the Fragility Notebook evaluates the polar crane as a potential (This F&O originated seismic interaction source relative from SR SFR-A2) to the reactor vessel and other NSSS components inside the containment structure. In the determination of the vertical spectral acceleration applicable to the polar crane, the computed fundamental frequency falls within a valley in the applicable ISRS, on the low frequency side of the primary spectral peak. E1-72

Enclosure 1 Attachment 3 - Resolution of the VEGP Seismic PRA Peer Review Findings F&O Review cc F&O F&O F&O Disposition for GSl-191 Number Element Description Basis Resolution and Other Applications Uncertainty in the calculated frequency, and the contribution of high modes, could result in an increase in the applied vertical acceleration. During the performance of the Peer Review, SNC personnel provided a written response indicating that it is appropriate to increase the applied acceleration by 50%, which will result in a 20% decrease in the median capacity of the polar crane. 14-8 SFR-F3 111111 Relay fragility The relay evaluation for the Perform more realistic The relay fragilities have calculations include turbine driven auxiliary feedwater relay fragility evaluations. been updated using the conservative pump control panel in calculation appropriate response and assumptions. PRA-BC-V-14-008 is based on a in-cabinet amplification generic capacity for motor starters factors, and are realistic. and contactors (intended for This finding has been motor control centers) and an resolved. (This F&O originated amplification factor associated from SR SFR-F3) with center of door panel response. Based on walkdown observations the relay is not mounted on the door panel so is likely on an internal bracket. The median capacity of 0.627g is well below the screening level and is not realistic. The relay evaluations in calculation PRA-BC-V-14-009 are governed by response in the vertical direction, and the in-cabinet amplification factors used in the calculation are associated with horizontal response. The E1-73

Enclosure 1 Attachment 3 - Resolution of the VEGP Seismic PRA Peer Review Findings F&O Review cc F&O F&O F&O Disposition for GSl-191 Number Element Description Basis Resolution and Other Applications resulting median capacities of 0.762g (Appendix M1) and 1.026g (Appendix M2) are well below the screening level and are not realistic. 14-9 SFR-D2 Met The seismic The summary of the seismic Perform resolution of The noted walkdown walkdown report walkdowns documents a number open items and provide issues have been includes a number of issues identified during the documentation of the evaluated and reflected in of open items that performance of the walkdowns resolution associated with the revised documentation: are not are not that required follow-up actions each of the issues, either - potential piping traceable to a (31). These include spatial in the Fragility Notebook interaction; resolution interaction issues, housekeeping or the SPRA Database. - the difference in inverter issues, anchorage issues, valves anchorage configuration; having configurations that do not - potential interaction meet the EPRI guidelines, concerns with the (This F&O originated configuration issues, installation overhead heater; this from SR SFR-D2) errors, etc. evaluation is in the fragility notebook. The Seismic Walkdown Report Valve operator heights & does not document how the weights that were outside issues identified during the EPRI guidelines have walkdowns have been addressed, been taken into account in either in the field (e.g., correction the fragility analysis for of installation errors, resolution of these components. housekeeping issues) or in the The Diesel Generator fragility evaluations (e.g., valve Exhaust Silencer was re-configurations, anchorage evaluated to the as-issues). During the performance operated condition. of the Peer Review, the Peer The fragility analysis for Review Team provided a list of these components has the walkdown issues to SNC been completed for the as personnel, and SNC provided a built condition. summary of how they were This finding has been addressed. Most issues had resolved. been adequately addressed during the development of the SPRA, but it was determined that E1-74

Enclosure 1 Attachment 3 - Resolution of the VEGP Seismic PRA Peer Review Findings F&O Review cc F&O F&O F&O Disposition for GSl-191 Number Element Description Basis Resolution and Other Applications the following would require further effort for resolution: (a) Potential interaction between piping and deluge valve - follow-up walkdowns required. (b) Anchorage configuration on inverter - follow-up revision to ' fragility evaluation required (c) Overhead heater poses potential interaction issue - follow-up walkdown required. (d) Valve operator heights/weights outside of EPRI guidelines - follow-up walkdown required. (e) Diesel Generator exhaust silencer anchor bolt nuts - not addressed in fragility evaluation, further evaluation required. (f) Valve operator heights outside of EPRI guidelines and potential lack of yoke support - these valves are part of the unfinished scope described in the Fragility Notebook, which will be completed in the future. (g) Valve operator heights outside of EPRI guidelines - further evaluation required. 16-1 SFR-F3, 11/111, The model Relay chatter is consistently being Complete the analysis The approach to screening SPR-84, Not presented for peer observed as a significant and incorporate the and modeling of SPR-E5 Met, review did not contributor to risk profile in effects of relay chatter seismically-induced relay II incorporate the recently peer reviewed S-PRAs and similar devices in the failures and chatter was effects of relay and it is therefore realistic to PRA logic model. provided to the peer chatter as the expect that relay chatter is a review team and analysis was not yet potential significant contributor. determined to have been complete. DurinQ the peer review it was performed aooropriately; E1-75

Enclosure 1 Attachment 3 - Resolution of the VEGP Seismic PRA Peer Review Findings F&O Review cc F&O F&O F&O Disposition for GSl-191 Number Element Description Basis Resolution and Other Applications discussed that the SPRA team only the incorporation into does not believe relays will be a the model of the impacts of significant contributors but it was relay chatter from (This F&O originated also said that this conclusion/ unscreened relays was not from SR SPR-84) expectation is based on complete. The final potentially crediting operator screening resulted in only actions. Thus, the effects of relay 2 relays being chatter per se may be significant incorporated into the (and provide some insights) while model, with one having an the combination of relays and a operator action. Relay number of HEP may not be. chatter fragilities and impacts have been incorporated into the seismic model, in a manner consistent with that used for other failures. This finding has been resolved. 16-10 SPR-86 Met The documentation There is only a short sentence More detailed Walkdown documentation about the supporting the discussion on documentation is on accessibility for walkdowns in alternative access pathways. suggested to support the operator actions, including support to seismic conclusion on photos, has been impact on HRA accessibility, alternative improved. Potential failure appear limited. route, availability of of block walls has been tools/keys, clear reviewed and documented. identification of Required tools and equipment manipulated in equipment, such as (This F&O originated each local action. ladders, have been from SR SPR-86) identified with locations Obviously, the goal of the when needed. The enhanced documentation documentation supports is not to convince the the seismic HRA peer reviewer that the assumptions and walkdowns were modeling. performed but rather to ensure that the analyst is This finding has been E1-76

Enclosure 1 Attachment 3 - Resolution of the VEGP Seismic PRA Peer Review Findings F&O Review cc F&O F&O F&O Disposition for GSl-191 Number Element Description Basis Resolution and Other Applications fully convinced of the resolved with no significant conclusions. impact to the SPRA results or conclusions. Past SPRAs have shown examples of equipment needed for the HFE that was not in the SEL, or that has different actuators when manually actuated, or that needed ladders that were not easily accessible or that were close to block walls (or under ceiling that could collapse) that were not considered an issue because the block walls were not near safety related equipment (and therefore not addressed in the rest of the SPRA work). In this perspective, a more systematic documentation of the feasibility and accessibility analysis for each of the HFE credited in the SPRA is suggested. E1-77

Enclosure 1 Attachment 3 - Resolution of the VEGP Seismic PRA Peer Review Findings F&O Review cc F&O F&O F&O Disposition for GSl-191 Number Element Description Basis Resolution and Other Applications 16-11 SPR-E2 Met Missing review of It is understood that the As this exercise was A detailed quantitative the potential for investigation performed in internal apparently performed for HRA dependency analysis additional events to identify potential HFE the Fire PRA (as based on using the HRA dependencies dependency has been relied upon discussed during the peer calculator was performed introduced by the in the Vogtle SPRA. review), it is suggested and documented. There SPRA models (QU- that a review of the was no significant impact C1&2) The SPRA logic may identify potential for unforeseen on results since human additional dependencies trends dependencies trends is actions are not significant that were not identified in the performed. contributors in the Vogtle internal events. SPRA. (This F&O originated As it is understood that from SR SPR-E2) the plan is to transition to This finding has been a different dependency resolved. analysis method (based on HRA calculator), this may be addressed within the same transition as it is realistic to expect that not too many (if any) new dependencies would be identified. 16-12 SPR-E2 Met Missing It is an industry expectation (as It is understood that the The QU report has been documentation of discussed in NEI peer review task SPRA documentation will updated to document the the review of non force meetings) that review of the be revised to incorporate review of both dominant significant cutsets non significant cutsets is explicitly explicitly the two reviews cutsets and non-significant QU-05. documented. discussed in the basis for cutsets for both CDF and this F&O. It is also LERF. Based on discussion during the recommended to peer review, two reviews were document the review of This finding has been (This F&O originated performed to validate the overall cutsets following resolved with no significant from SR SPR-E2) model and cutsets. The first was guidance from the NEI impact to the SPRA results a random review of cutsets at peer review task force. or conclusions. midpoints and low significance for each of the %Gxx initiators to verify that the cutsets are valid cutsets, and that the patterns are appropriate. That is, if one cutset E1-78

Enclosure 1 Attachment 3 - Resolution of the VEGP Seismic PRA Peer Review Findings F&O Review cc F&O F&O F&O Disposition for GSl-191 Number Element Description Basis Resolution and Other Applications is valid, then another cutsetwith slightly different seismic failures (or random failures) should also be nearby. The second review, more importantly, lowered the median seismic capacity for each of the seismic initiators and some of the other seismic failures to ensure that the model would properly generate valid cutsets. For example, the LLOCA fragility was reduced to 0.5g to generate LLOCA cutsets. For ATWT, the fragility of the CRDs and RV internals were reduced to 0.5g to verify that valid A TWT cutsets were generated. 16-15 SPR-E6, Met, Documentation of The current documentation does Expand the The LERF documentation SPR-F2 Met LERF model not explain what are the basis for documentation to ensure in the QU report was applicability review. retaining the LERF logic and that the criteria used to expanded to describe the analysis unchanged within the retain the LERF analysis review of applicability of SPRA logic. in the SPRA is explained the internal events PRA so that the same LERF analysis to the (This F&O originated During the peer review the applicability review can seismic PRA. from SR SPR-F2) following explanation was be performed following provided by the SPRA team: future potential revisions This finding has been of the LERF modeling. resolved with no significant "The internal events Level 2 impact to the SPRA results notebook was reviewed to ensure or conclusions. that the definition of LERF would be appropriate for seismic events. The LERF definition includes the use of a 12 hour time period for release after event initiation, to allow for evacuation. This time E1-79

Enclosure 1 Attachment 3 - Resolution of the VEGP Seismic PRA Peer Review Findings F&O Review cc F&O F&O F&O Disposition for GSl-191 Number Element Description Basis Resolution and Other Applications period is considered to be valid for Vogtle seismic events, particularly due to the very low population density in the area. Other characteristics, such as bypass and scrubbing, are the same for seismic as for internal events. The logic for the internal events LERF model is very straightforward, with sequences from the GDF model ANDed with the appropriate LERF fault tree. This logic is also appropriate for seismic events." 16-18 SPR-88 Ill Very small LOCA The DB has a specific entry for To the peer review team Additional information on have been screened the incore thermocouples and knowledge Vogtle is the the walkdown for very from the analysis provides pictures of them. Still, in- only plant that has small LOCA has been based on core thermocouple tubing is not elected to perform added to fragility report to walkdowns but little the only possible source of very dedicated walkdowns in provide the basis for the documentation small LOCA that is envisioned support of not modeling VSLOCA screening. exists of such and the only documentation of very small LOCA. This walkdowns. addressing the other potential would be a best practice This finding has been sources is in the quantification but it also behooves to resolved with no significant notebook: the SPRA team to impact to the SPRA results provide detailed or conclusions. (This F&O originated "For Vogtle 1&2, the seismic documentation of such from SR SPR-88) walkdowns inspected and walkdowns and how they photographed a large sample of supported a systematic the small piping and tubing lines evaluation of the potential connected to the primary system sources of very small in order to identify any LOCA. weaknesses. The piping was iudQed to be ruQQed." E1-80

Enclosure 1 Attachment 3 - Resolution of the VEGP Seismic PRA Peer Review Findings F&O Review cc F&O F&O F&O Disposition for GSl-191 Number Element Description Basis Resolution and Other Applications 16-2 SFR-C1, I/II, Fragilities were not The 2014 hazard was only used During the peer review The fragilities have been SPR-E1 Met corrected to reflect as input to FRANX for the final the SNC staff answered a recalculated based on the the 2014 hazard quantification. It is understood question on this topic by 2014 hazard and the new used for that the fragility estimates have performing an initial values incorporated into quantification. (This been performed based on the limited investigation of the the SPRA model and F&O originated from 2012 hazard. While it is not effect on fragilities quantification. SR SPR-E1) expected nor recommended to correction to reflect the regenerate all the fragility work 2014 hazard and This finding has been with the new hazard, some concluded that the effect resolved. consideration on the possible of this scaling is not change in fragility due to the use insignificant (especially of the newer hazard should be for LERF). It is made. recommended to continue and expand this investigation to make the quantification fully consistent with the fragility values. 16-4 SPR-82 Not The effect of seismic There is no assessment of the While it is recognized that The methodology used for Met impact on effect of changing the breaking the industry is still the seismic HRA analysis performance points in the Surry method. The developing methods in is based on defining PSFs shaping factors is Surry method is based on support to this particular as a function of seismic considered in the methods used in the past at topic (e.g., recently hazard level (bins), which analysis by the SONGS and Diablo Canyon and published EPRI HRA is consistent with the EPRI usage of the Surry the 0.8g breaking point was method for external seismic HRA guidance in method. developed for California events), some additional EPRI 3002008093. The earthquakes. In the Vogtle considerations should be Integrated PSFs and bins analysis there is no indications on done to understand the (breaking points) have whether the breaking point at effect of HEPs in the been updated with (This F&O originated 0.8g is.also applicable to Vogtle. model rather than simply additional breaking points from SR SPR-82) There are also no sensitivity implementing the Surry and integrated PSFs to analyses that would support method as is. reflect seismic binning whether a change in the breaking applicable to Vogtle, in points is significant or not. Three examples for accordance with this addressing this finding finding and consistent with may be the following: the EPRI guidance. The

1. Perform sensitivities on updated values have been E1-81

Enclosure 1 Attachment 3 - Resolution of the VEGP Seismic PRA Peer Review Findings F&O Review cc F&O F&O F&O Disposition for GSl-191 Number Element Description Basis Resolution and Other Applications the values of the applied to both internal multipliers and the g events HFEs and seismic-levels where the breaking unique HFEs within the point happens. plant response model.

2. Use a different There was no significant multipliers method with impact on the SPRA more breaking points. results.

3. Apply the impact of seismic specific PSF at This finding has been the individual PSF level resolved.

(i.e., timing, stress, etc.) in the HRA calculator. 16-5 SPR-81, Met, LOCA modeling and The selection of the fragility data Documentation on the LOCA basis has been re-SPR-F1 Not fragility selection not used for all LOCA is discussed in use of fragility in support evaluated and updated. Met clearly documented. Appendix 8.2 of the quantification to LOCA should be This was partially due to notebook but is confusing in the clarified to better seismic fragility update mapping of selected fragilities represent the rationale and partially a matter of with specific failures. selected and potentially adding amplifying (This F&O originated addresses the modeling information to the LOCA from SR SPR-F1) It appears that the fragility uncertainties associated basis. The quantification selected to represent LOCA with this selection. report includes updated sequences are coming from documentation. Although specific components but then While this finding is LOCAs are a significant they are used to represents sort expected to be addressed contributor to the SPRA of surrogate events for potential via documentation, some results, the VEGP SCDF failures along the piping network. additional suggestions and SLERF are sufficiently are provided, such as: small that further LOCA Using localized events as modeling sensitivity surrogate for pipe network failure 1. Perform a sensitivity to beyond what has been is probably conservative and may show that the modeling provided in the updated not be fully consistent with the approach described is not model quantification is not system success criteria and significantly skew the warranted. modeling in the internal events results for seismic; modeling. For example, the This finding has been seismic-induced MLOCA fragility 2. Modify the logic by resolved. seems to be based on failure of mapping the seismic-the pressurizer surge line, which induced MLOCA to a E1-82

Enclosure 1 Attachment 3 - Resolution of the VEGP Seismic PRA Peer Review Findings F&O Review cc F&O F&O F&O Disposition for GSl-191 Number Element Description Basis Resolution and Other Applications is a localized failure. The seismic- different position in the induced MLOCA initiator is logic (e.g., a dummy mapped to the internal events event can be entered in MLOCA initiator. The internal the model to provide a events logic for MLOCA has a target for the FRANX split fraction that divides MLOCA injection). (and LLOCA) in four 25% contributors impacting all four CL/HL. Since the seismic-induced MLOCA is a localized failure, the internal events logic is not fully applicable (probably slightly conservative). Because the documentation is potentially leading to a misunderstanding of the selected approach (thus impacting ease on update), this F&O is considered a finding against the documentation SR. 16-6 SPR-B2 Not The effect of seismic The Vogtle SPRA elected to use Expand the IPSF The methodology used for Met impact on Integrated Performance Shaping approach to all the the seismic HRA analysis performance Factors (IPSF) multipliers. While operator actions credited is based on defining PSFs shaping factors is this approach was used for the in the SPRA. as a function of seismic not considered for HEPs that were carried over from hazard level (bins), which any action that was internal events, it was is consistent with the EPRI explicitly added for systematically not done for all the seismic HRA guidance in the SPRA (e.g., actions explicitly added for EPRI 3002008093 flood isolation or DG seismic. [9]. The Integrated PSFs output breaker and bins (breaking points) closure). Based on discussion during the have been updated to peer review, the analyst believed reflect seismic binning that having designed these applicable to Vogtle, in actions for specific scenarios accordance with this following a seismic event, the finding and consistent with (This F&O originated impact of seismic specific PSF is the EPRI Quidance. The E1-83

Enclosure 1 Attachment 3 - Resolution of the VEGP Seismic PRA Peer Review Findings F&O Review cc F&O F&O F&O Disposition for GSl-191 Number Element Description Basis Resolution and Other Applications from SR SPR-B2) already included. updated values have been applied to both internal The objection to this conclusion is events HFEs and seismic-that the seismic specific PSF unique HFEs within the should realistically change with plant response model. the magnitude of the event. This There was no significant change addresses the change in impact on the SPRA the overall context of the plant results. when a small seismic event happens as opposed to when a This finding has been very large seismic event happens. resolved. This seems not to be captured by the approach selected for the Vogtle SPRA. One example of this is that an action that has a 30 minute Tsw (S-OA-BKR-LOCAL) maintains an HEP of 1.60E-03 at all g levels, including the %G14 interval (i.e., >2g). It is understood that this is not expected to be quantitatively significant because failure of the recovered equipment is taken care by the logic model. 16-7 SPR-E2 Met Base case seismic Both CDF and LERF are LERF at 1E-11 truncation LERF truncation, which LERF does not meet truncated at 1.0E-09 with 1000 meets the QU-83 was already considered in the truncation cutsets managed by ACUBE. This truncation requirement. sensitivity studies, has requirements from meets the QU-B3 requirement for Rename LERF at 1E-11 been revised appropriately QU-B3. CDF but not for LERF. as the base case for to meet QU-B3. A new LERF. LERF truncation limit has been established consistent with the LERF (This F&O originated results. Quantification is at from SR SPR-E2) 1E-12, which is a suitably low value. E1-84

Enclosure 1 Attachment 3 - Resolution of the VEGP Seismic PRA Peer Review Findings F&O Review cc F&O F&O F&O Disposition for GSl-191 Number Element Description Basis Resolution and Other Applications This finding has been resolved. 16-8 SPR-E2 Met Missing There is a description of the most While it is understood that The QU report has been documentation of important scenarios but there is the Draft. B version of the updated to document the cutsets review (cfr. no cutset-by-cutset review. quantification notebook is review of both dominant QU-D1) still somewhat a work in cutsets and non-significant process, it is expected cutsets for both CDF and that when the model LERF. reaches a more stable (This F&O originated state documentation of This finding has been from SR SPR-E2) the review of the cutsets resolved with no significant is going to be part of the impact to the SPRA results documentation. or conclusions. 16-9 SPR-B1, Met, Screening values At the time when the An appropriate resolution The seismic HRA analysis SPR-B4b Met used for the HEPs documentation was provided for of this F&O is pending the has been revised to be that (at the time of peer review, the most significant current evolution of the consistent with the EPRI the provided operator actions (i.e., flood model and the seismic HRA guidance in documentation) isolation of ACCW HX) were all importance of operator EPRI 3002008093. The were in the most screening values, which would actions in the SPRA. original screening HEPs significant cutsets. only meet CCI for HR-G1 (directly Given the expectation have been updated using called through SPR-B1 ). that operator actions will the HRA Calculator, be needed to mitigate the consistent with the In addition, there is little importance of relay approach used in the (This F&O originated documentation or supporting chatter (not yet included VEGP internal events from SR SPR-B1) evidence to justify screening in the SPRA logic model) PRA. The Documentation values as low as 3.00E-2 this F&O was provided to has been updated. ensure care is used in the Operator response to relay generation of HEPs if chatter has been they appear in important addressed and evaluated cutsets and also to within the same process, provide more justification and not found to be for screening values less important. than 1.00E-1 because a low screening value may This finding has been indeed skew the actual resolved. importance of the newly qenerated HEP. E1-85

Enclosure 1 Attachment 3 - Resolution of the VEGP Seismic PRA Peer Review Findings F&O Review cc F&O F&O F&O Disposition for GSl-191 Number Element Description Basis Resolution and Other Applications 17-1 SPR-81 Met The documentation The modeling approach injected A separate section in the The discussion of accident does not specifically seismic fragilities into fault trees documentation that sequences and success address the that were modified from the specifically addresses criteria has been applicability of the internal events PRA model. It can accident sequences and expanded, and specific internal events be inferred from this approach, success criteria is needed descriptions of the flooding accident sequences and it was verified by discussions to collect the information scenarios has been added. and success criteria with the staff, that the internal in one logical place, and This finding is to the SPRA model, events sequences and success is needed to support documentation only and and does not criteria were considered to be effective peer reviews does not impact Seismic properly document applicable to the SPRA model. and future model PRA model results. the accident This was not specifically stated in updates. sequences created the documentation. This finding has been specifically for the resolved with no significant SPRA model. Further, several additional impact to the SPRA results seismic flooding sequences were or conclusions. added to the fault tree. These sequences are not discussed (This F&O originated from an accident sequence and from SR SPR-81) success criteria perspective. Inspection of the fault tree and discussions with the staff indicate that the sequences were appropriately developed with specific success criteria that is different from other internal events sequences. The development of these sequences needs to be included in the documentation. Including event trees for these sequences would also aid in a reader's understandinQ. E1-86

Enclosure 1 Attachment 3 - Resolution of the VEGP Seismic PRA Peer Review Findings F&O Review cc F&O F&O F&O Disposition for GSl-191 Number Element Description Basis Resolution and Other Applications 17-2 SPR-E2, Met, The processes used Examples include: Expand the Documentation for QU SPR-F2 Met to create the documentation to clearly results has been im-presented The top cutsets shown in the explain the post- proved to describe the quantification results quantification report are produced processing of the results processes used to ag-are not fully by combining the cutsets from all generated by CAFTA and gregate results over the 14 documented. the seismic interval cutsets in a FRANX. Examples hazard intervals. The process that is not documented. include: importance calculations have been re-quantified While the process used to obtain - Explain how the cutsets and the method for (This F&O originated the importance measures is generated by FRANX are presentation documented. from SR SPR-F2) documented, discussions with the combined into g-level-PRA staff indicated that independent cutsets. This finding has been importances for some of the basic resolved with no significant events were obtained in a - Explain the post- impact to the SPRA results different manner (setting to one or processing used to or conclusions. zero and requantifying). This is generate importance not documented in the notebook. measures, especially focusing on the deviation from a normal practice that is currently only mentioned in the notebook. 17-3 SPR-83, I/II, Subdividing To account for similar equipment The impact of the The non-minimal cutsets in SPR-E4 I/II correlation groups that has different fragilities due to retention of these non- the peer reviewed model based on different building locations, minimal cutsets on were identified and weaker/stronger certain correlation groups were CDF/LERF and reviewed for impact, and components subdivided to assign a seismic importance measures determined to be non-resulted in retention capacity to a weaker component should be assessed and significant to risk. The of non-minimal that only failed that component. the results documented, results were very slightly cutsets in some The higher capacity was then or a method to remove conservative due to these cases, which could assigned to both components, the non-minimal cutsets non-minimal cutsets. The impact CDF/LERF and was effectively the correlated should be devised. Each issue has been addressed results as well as failure of both components. This subdivided correlation in the updated model, such model importance can result in the retention of non- group should be non-minimal cutsets no measures. The minimal cutsets in some cases. investigated for similar longer appear. magnitude and For example, for the Containment effects. acceptability of Fan Cooler Units there are This finding has been E1-87

Enclosure 1 Attachment 3 - Resolution of the VEGP Seismic PRA Peer Review Findings F&O Review cc F&O F&O F&O Disposition for GSl-191 Number Element Description Basis Resolution and Other Applications these impacts was cutsets in which, due to other resolved with no significant not documented. failures, only one containment fan impact to the SPRA results cooler needs to seismically fail to or conclusions. cause core damage. Inspection of the cutsets shows that two otherwise identical cutsets, are retained: one in which the 1Fan (This F&O originated 'group' occurs, and one in which from SR SPR-E4) the 4Fans group occurs. The 4Fans cutset is not minimal, and should not be included in the results. Discussions with the staff indicated that these non minimal cutsets were noted during the quantification review process, but were thought to not greatly impact overall results. No formal assessment was done, however, and no record of the informal assessment was included in the documentation. 17-4 SPR-E6 Met No quantitative A quantitative analysis is required Perform the analysis and The quantitative analysis analysis of the to meet CCII for LE-F1 & LE-G3, include the results in the of significant LERF plant relative contribution which are directly called from quantification notebook. damage states and to LERF from Plant SPR-E6. contributors has been Damage States and performed. A table and Significant LERF associated discussion of contributors was plant damage states and presented in the significant contributors has quantification been added to the LERF results. QU documentation to resolve this finding. (This F&O originated from SR SPR-E6) This finding has been resolved. E1-88

Vogtle Electric Generating Plant - Units 1 & 2 Supplemental Response to NRC Generic Letter 2004-02 Enclosure 3 Risk Quantification _ I -

Enclosure 3 Risk Quantification Table of Contents 1.0 Introduction 2.0 Scope of Risk Assessment 3.0 Failure Mode Identification 4.0 PRA Model Changes 5.0 Sub-Model Development 6.0 Scenario Development 7.0 Debris Sources 8.0 Chemical Effects 9.0 Debris Transport 10.0 Strainer Debris Impact Evaluation 11 .0 Debris Penetration Evaluation 12.0 Debris Penetration Effects 13.0 Sub-Model Integration 14.0 Systematic Risk Assessment 15.0 References E3-1

Enclosure 3 Risk Quantification 1.0 Introduction Figure 3-1 shows a high-level summary of the GSl-191 risk quantification including the interface between the probabilistic risk assessment (PRA) evaluation, the phenomenological evaluation, and the risk guidelines from RG 1.174 (Reference 1).

4. PRA quantification of CDF & LERF with no GS!-191 failures

2. Evaluation of GSI*

1. PRA Identification 191 phenomena for of accident scenarios each high likelihood 3. PRA quantification

5. Calculate llCDF &

and high likelihood configuration to of CDF & LERF with equipment lllERF determine CfPs for GSl-191 failures configurations each strainer/core failure basic event 7a. Identification of

8. Output to analytical refinements to submittal analysis documentation 7b. Identification of potential plant modifications (physical or procedural)

Figure 3 Flow Chart Illustrating Risk-Informed GSl-191 Evaluation The following steps are included in this flow chart:

• Step 1: PRA identification of accident scenarios and high likelihood equipment configurations Step 1 is an important part of the overall evaluation because it defines the scope of the GSl-191 problem. This step has two important parts: identification of accidents requiring recirculation through the emergency core cooling system (ECCS) strainers, and identification of high likelihood equipment configurations.

As discussed in Enclosure 1, Section 2.1, the VEGP risk-informed GSl-19'1 evaluation considered all breaks that require strainer recirculation, which includes loss of coolant accidents (LOCAs) and secondary side breaks inside containment (SSBls). For the accidents that need to be evaluated, it is important to identify the high likelihood equipment configurations. For example, because VEGP has two E3-2 ________ _J

Enclosure 3 Risk Quantification 100 percent capacity independent trains of ECCS and containment spray system (CSS) including a containment spray (CS) pump, a residual heat removal (RHR) pump, a safety injection (SI) pump, and a charging pump, there could be a large number of random failure combinations unrelated to GSl-191 issues (i.e., failure to start or failure to run). However, some equipment failure combinations may have sufficiently low probability that they can be addressed with a bounding analysis. The high likelihood equipment configurations are described in Section 6.3.

• Step 2: Evaluation of GSl-191 phenomena for each high likelihood scenario to determine CFPs for each basic event Step 2 is the process of using NARWHAL to calculate each of the conditional failure probabilities (CFPs) required by the PRA model. The details of this analysis are expanded upon in Section 13.0.
• Step 3: PRA quantification of CDF and LERF with GSl-191 failures Step 3 is a straightforward quantification of the core damage frequency (CDF) and large early release frequency (LERF) based on the GSl-191 failures for the as-built/as-operated plant.

As discussed in Enclosure 1, Section 2.2, some minor modifications to the base PRA were necessary to perform this evaluation. Specifically, the GSl-191 PRA model includes basic events for both strainer failures and core failures to capture the GSl-191 CFPs. The modifications are described in more detail in Section 4.0.

• Step 4: PRA quantification of CDF and LERF with no GSl-191 failures Step 4 is essentially identical to Step 3, but all GSl-191 CFPs were set to zero to represent a hypothetical modification to the plant that would prevent any GSl-191 failures (e.g., removal of all sources of fiber debris).
• Step 5: Calculate ~CDF and ~LERF Step 5 is a simple comparison of the values calculated in Steps 3 and 4. The change in CDF (~CDF) and change in LERF (~LERF) values represent the risk associated with GSl-191.
• Step 6: Comparison to RG 1.174 Guidelines Step 6 is a comparison of the CDF, LERF, ~CDF, and ~LERF values to the acceptance guidelines defined in RG 1.174 (Reference 1). The ~CDF limits are defined as 1x1 o-5/yr for the boundary between Regions I and II (low risk), and 1x1 o-6/yr for the boundary between Regions II and Ill (very low risk). The ~LERF E3-3

Enclosure 3 Risk Quantification limits are one order of magnitude lower than the ~CDF limits (1 x1 o-6/yr for the boundary between Regions I and II, and 1x1 o-7/yr for the boundary between Regions II and Ill). Ideally, the evaluation would show that the risk associated with GSl-191 is in Region Ill. However, as described in RG 1.174, Region II is acceptable for a risk-informed submittal (Reference 1).

• Step 7a: Identification of analytical refinements to analysis In general, risk-informed evaluations should not contain significant conservatisms because conservatism skews the results in a way that might mask the true sources of risk. However, in practice, some level of bias that is generally accepted to be conservative is incorporated in risk-informed evaluations to avoid the extensive analysis and testing that would be required to develop more refined inputs and models. Therefore, if the results of the risk-informed evaluation are unacceptable, Step 7a can be used to identify specific conservatisms and implement refinements that may reduce the calculated risk to an acceptable level.

Over the course of developing the VEGP risk-informed GSl-191 evaluation, a variety of refinements and simplifications were incorporated for different aspects of the model to balance realism and conservatism.

• Step 7b: Identification of potential plant modifications (physical or procedural)

Step 7b is very similar to Step 7a, but is predicated on the analysis being as refined as practical (i.e., the inputs and analytical models used in the risk-informed evaluation are considered to be a reasonable representation of the post-accident conditions). In this case, if the risk is determined to be unacceptably high, it is necessary to make one or more plant modifications. The modifications could include physical changes (strainer replacement, insulation replacement, buffer change, degraded coatings remediation, etc.) and/or procedural changes (securing pumps earlier in the event, initiating hot leg recirculation earlier, etc.). If a plant modification is necessary, various options can be evaluated to determine the impact on risk, as well as the cost-benefit of implementing each option. Note that even if Step 6 shows that the risk is acceptably low, it may be beneficial to evaluate and implement some minor modifications that significantly reduce risk. For VEGP, a combination of physical and procedural modifications were determined to be beneficial. As described in Enclosure 2, Section 3.j, the RHR strainer height will be reduced by removing disks, and the procedures for switching over from refueling water storage tank (RWST) injection to sump recirculation are being modified to inject more water into containment for breaks that do not initiate containment sprays. These modifications will help ensure that the strainers are completely submerged for an increased number of postulated E3-4

Enclosure 3 Risk Quantification* LOCA scenarios, which significantly reduces the likelihood of debris-related failures.

• Step 8: Output to submittal documentation Once the risk has been determined to be acceptable, a licensing submittal can be prepared to document the risk quantification, defense-in-depth, safety margin, etc. This submittal is documented for VEGP in Enclosures 1 through 5.

2.0 Scope of Risk Assessment As discussed in Enclosure 1, Sections 2.0 and 3.0, the scope of the risk model includes:

• Reactor coolant system (RCS) pipe breaks resulting in small, medium, and large LOCAs (includes breaks ranging from %" partial breaks to double-ended guillotine breaks (DEGBs) on every Class 1 ISi weld within the first isolation valve) o This includes breaks due to seismically-induced LOCAs.
• SSBls that result in a consequential LOCA upon failure to terminate safety injection or a stuck open power operated relief valve (PORV) (includes DEGBs at least every 5 ft on main steam line and feedwater line piping inside containment)

Water hammer LOCAs, non-piping LOCAs, and breaks past the first isolation valve were qualitatively addressed or screened out as described in Enclosure 1. In addition, as discussed in Enclosure 1, the risk model was evaluated for Mode 1, which is bounding for Modes 2 through 6. 3.0 Failure Mode Identification The specific GSl-191 failure modes that were included in the risk model are:

• Strainer head loss exceeds the net positive suction head (NPSH) margin for the RHR and CS pumps when the strainer is fully submerged.
• Strainer head loss exceeds half of the submerged strainer height when the strainer is partially submerged.
• Strainer head loss exceeds the strainer structural margin.
• Gas voids downstream of the strainers exceed the acceptable void fraction limits of the RHR or CS pumps.
• Debris accumulation on the strainer exceeds tested debris limits. Although this by itself is not a unique failure mode, it represents an unknown head loss condition where any of the failure modes above may occur.
• Debris accumulation in the core exceeds debris limits for core blockage and boric acid precipitation.

Note that containment sprays are not required for containment cooling. Therefore, although CS pump failures were evaluated in the NARWHAL model, these failures were E3-5

Enclosure 3 Risk Quantification not included in the CFP results. As discussed in Enclosure 1, Section 1.0, other failure modes (upstream blockage, vortexing, ex-vessel downstream effects, and the LOCA deposition model (LOCADM) portion of the analysis of in-vessel effects) were addressed in a bounding manner for the range of possible breaks with no issues of concern, and were therefore not explicitly modeled in NARWHAL. Note that no significant direct or indirect effects associated with these excluded failure modes have been identified with respect to the risk model. 4.0 PRA Model Changes To perform the risk-informed GSl-191 evaluation, a few relatively minor changes were required for the base PRA model to incorporate the events for the GSl-191 sump strainer and core blockage failures, along with the associated LOCA initiating events and equipment configurations. The following configurations were represented in the NARWHAL evaluations to determine CFPs for GSl-191 related strainer and core failures, and were included in the modifications to develop the VEGP GSl-191 PRA model:

• No equipment failed (all ECCS trains operating)
• One RHR train (RHPA or RHPB) failed
• One charging (CP) train (CPPA or CPPB) failed
• One SI train (SIPA or SIPS) failed
• One nuclear service cooling water (NSCW) train (NSCWA or NSCWB) failed, causing failure of one ECCS train
• One CS train (CSPA or CSPB) failed
• Two CS trains (CSPA and CSPB) failed Each potential sump strainer and core blockage failure can be represented in the PRA model with a basic event that is combined with the appropriate LOCA initiating event and pump failure logic to represent the equipment configurations listed above. Some simplifications of the modeled configurations were performed based on the assumptions described below:
• The CS pumps were assumed not to actuate during a medium or small LOCA; thus, configurations with one or both CS trains failed were modeled for large LOCA initiating events, and were also modeled for the medium LOCA initiating events for sensitivity purposes only.
• CS is not required to be modeled in the VEGP PRA for containment cooling; therefore, the sump strainers for the CS pump recirculation suction, and associated GSl-191 failures, were also not included in the VEGP GSl-191 PRA model.
• Because the charging and SI pumps piggyback off the RHR pumps, failure of one charging or one SI train has the same impact on debris transport to the RHR sump strainer as the configuration with all ECCS trains operating; therefore, these configurations were grouped together.

E3-6

Enclosure 3 Risk Quantification The VEGP base PRA model currently includes events to represent independent and common cause plugging of ECCS containment sumps A and B. For convenience, these existing sump-plugging events were combined under new logic gates NON-GSl-191-A and NON-GSl-191-B. These existing sump plugging events may ultimately be removed from the VEGP PRA model of record and replaced with the more detailed representation of GSl-191 sump strainer failure logic described below. The logic developed for GSl-191 RHR sump strainer failures was modeled under new fault tree gates GSl-191-SUMP-A, GSl-191:.SLJMP-B, and GSl-191-SUMP-AB. Although the sump strainer CFPs may be zero for some of the scenarios included in the logic, these events were retained for potential sensitivity studies. The new logic gate for GSl-191 failure of RHR Strainer A (GSl-191-SUMP-A) was included under the existing PRA model gates for loss of flow from the ECCS containment sump A (ECCS-SUMP-A and ECCS-SUMP-A-ACR). Similarly, the new logic gate for GSl-191 failure of RHR sump strainer B (GSl-191-SUMP-B) was included under the existing PRA model gates for loss of flow from the ECCS containment sump B (ECCS-SUMP-B and ECCS-SUMP-B-ACR). Finally, the new logic gate for GSl-191 failure of both RHR sump strainers A and B (GSl-191-SUMP-AB) was included under the existing PRA model gates for loss of flow from sump A and from sump B (ECCS-SUMP-A, ECCS-SUMP-A-ACR, ECCS-SUMP-B, and ECCS-SUMP-B-ACR). The logic developed for GSl-191 core blockage was modeled under new fault tree gate GSl-191-CORE. As with the sump strainer CFPs, the core blockage CFPs may be zero for some of the scenarios included in the logic; however, these events were also retained for potential sensitivity studies. The new logic gate for GSl-191 core blockage (GSl-191-CORE) was added under the existing PRA model top gate for core damage (GDF-TOTAL). A GSl-191 core blockage logic gate was also added under the existing PRA model gates for LERF end states 01 through 08 (gates LERF-01 through LERF-08). End states LERF-01 through LERF-06 can result from small LOCAs only; therefore, the small LOCA core blockage gate (GSl-191-CORE-SL) was combined with the containment event tree logic associated with each of those end states (e.g., gate LERF01 X ... LERF06X). End state LERF-07 can result from medium or large LOCAs; therefore, the medium and large LOCA core blockage gates (GSl-191-CORE-ML "OR" GSl-191-CORE-LL) was combined with the containment event tree logic for end state LERF-07 (gate LERF07X). Finally, end state LERF-08 can result from small, medium, or large LOCAs; therefore, the core blockage gate for all LOCAs (GSl-191-CORE) was combined with the containment event tree logic for end state LERF-08 (gate LERF08X). Proceduralized operator actions for switching from RWST injection to sump recirculation, switching from cold leg recirculation to hot leg recirculation, and securing containment sprays were all included in the risk-informed GSl-191 evaluation. However, no operator actions were credited to recover from the effects of debris-related failures on the strainer or the core in the b.CDF and b.LERF calculation.* E3-7

Enclosure 3 Risk Quantification 4.1 Seismic PRA Model To quantify the contribution of seismically-induced LOCAs to GSl-191 risk, the initiating event frequency for such LOCAs was determined by considering the seismic fragilities for several RCS components whose failure could cause a LOCA. While LOCAs are generally related to piping failures for the internal events PRA, the piping fragility analyses for VEGP demonstrated that the RCS loop piping and the Nuclear Class 1 piping inside containment have high seismic capacity. Other systems with LOCA sensitive piping, including the chemical volume and control system (CVCS), RHR, SI, reactor vessel head vent, and reactor vessel level indication system (RVLIS), were also considered for the potential occurrence of a seismically-induced LOCA within the size range of concern. This ensured seismically-induced pressure boundary failure was considered for a range of components to identify the bounding median seismic capacity for potential LOCAs. For development of a seismically-induced LOCA frequency, the seismic capacity of the reactor coolant pump (RCP) was selected. The RCP has the lowest seismic capacity among the nuclear steam supply system (NSSS) components. The selected bounding seismic capacity is based on an indirect seismically-induced LOCA due to the failure of the RCP column assembly support, which provides the least seismic margin of safety. Due to the flexibility and strength of the primary system piping and its supports, and the uncertainty attached to the postulated break size, this indirect seismically-induced failure could range in size from a small LOCA to a large LOCA. Therefore, the failure frequency is divided equally among the three LOCA initiating event sizes modeled in the VEGP PRA. The VEGP internal events PRA model modifications made to perform the risk-informed GSl-191 evaluation were used as a guide to make corresponding modifications to the VEGP seismic PRA model. These modifications included the incorporation of the CFPs calculated by NARWHAL for strainer and core failures due to the effects of debris generated by a large break LOCA. It is assumed that these CFPs are also applicable to a (direct or indirect) seismically-induced LOCA occurring in one location. The indirect seismic LOCAs that are postulated are for a break in the cold leg or crossover leg piping next to any one of the RCPs. The RCP break location is postulated based on a collapse of the RCP support structure following a seismic event. The resulting break could be any size from a small break up to a DEGB of the cold leg or crossover leg. The CFP values were calculated based on a range of break sizes (from Y2 inch to a DEGB) postulated at all Class 1 ISi welds within the first isolation valve. The NARWHAL evaluation showed that none of the small and medium breaks generate sufficient debris quantities to cause GSl-191 failures (including small and medium breaks on the cold legs and crossover legs). Many large breaks also do not generate sufficient debris quantities to cause failures. The large break GSl-191 CFP values were calculated based on the LOCA frequencies for a random pipe break (with a higher frequency for 6-inch breaks and a much lower frequency for 31-inch DEGBs). Although the frequency as a function of break size would likely be different for a seismically-induced LOCA, it is E3-8

Enclosure 3 Risk Quantification reasonably conservative to use the large break CFP values, because a seismic event could result in a small or medium break that would not cause any debris related failures. A seismic event also has the potential to dislodge insulation from piping and components inside containment. The insulation is contained inside a fabric cover; therefore, any dislodged insulation would be expected to fail in relatively-large intact pieces that would be unlikely to transport to the sumps or have a significant effect on the strainers. 5.0 Sub-Model Development The evaluation of failures related to the effects of debris was performed outside the VEGP PRA model using the NARWHAL software. The VEGP NARWHAL model includes the following sub-models:

• Post-accident conditions o Plant configuration o Plant states o Random equipment failures o Water volume and level o Flow rates o Pressure and temperature o pH
• Debris sources o Zone of influence (ZOl)-generated insulation, fire barrier, and qualified coatings debris o Unqualified coatings debris o Latent debris o Miscellaneous debris
• Chemical effects o Release via corrosion/dissolution o Solubility o Precipitate debris quantity
• Debris transport o Slowdown o Washdown o Pool fill o Recirculation o Erosion
• Strainer head loss o Debris groups o Clean strainer head loss o Conventional debris head loss o Chemical debris head loss o Head loss correction o Head loss extrapolation E3-9

Enclosure 3 Risk Quantification

• Strainer air intrusion
• Strainer and pump acceptance criteria o Strainer flashing o Strainer structural margin o Strainer partial submergence limit o Pump void fraction limit o Pump NPSH margin o Debris limits
• Strainer penetration o Fiber o Particulate
• Core acceptance criteria These sub-models are described in more detail in the following sections.

6.0 Scenario Development The post-accident conditions are described in the following sub-sections. These conditions include the plant configuration, plant state changes (due to automatic and manual operator actions), and various thermal-hydraulic parameters that were used in the VEGP NARWHAL model. For all breaks evaluated, a 30-day mission time was used in the NARWHAL model. This is consistent with the mission time used for deterministic GSl-191 evaluations (Reference 2). Note that the VEGP PRA model uses the typical PRA mission time of 24 hours. Any RHR strainer and core failures predicted by NARWHAL (regardless of failure time) were included in the CFP values that were used in the GSl-191 PRA calculation. Although additional operator actions and compensatory measures (such as refilling the RWST) could be taken to mitigate late failures, these actions were not included in either the GSl-191 PRA model or the NARWHAL model. 6.1 Plant Configuration The plant configuration determines the flow paths during the different phases of accident mitigation. At VEGP, there are a total of four separate sump strainer assemblies and two engineered safety features (ESF) trains. An ESF train consists of pumps associated with the ECCS and the CSS. A single train of ECCS consists of a high head centrifugal charging pump, a medium head SI pump, and a low head RHR pump. In addition to these pumps, each train of ECCS contains two SI accumulators. A single train of CSS contains a CS pump. There are two strainers dedicated to each ESF train, one for the ECCS pumps and one for the CS pump. Note that the strainers dedicated to the ECCS pumps are referred to as RHR strainers, and the strainers dedicated to the CSS are referred to as CS strainers. Figure 3-2 shows the connection diagram from NARWHAL that was used to model VEGP. Valves are used in NARWHAL to define whether a flow path is currently active. E3-10

Enclosure 3 Risk Quantification For example, each pump has a connection to the RWST and the strainer. During injection, the valves connecting the RWST to each of the pumps would be open and the valves connecting the pumps to the sump would be closed. The activity state and flow paths associated with each pump are discussed further in Section 6.2. As shown in the figure, the equipment was arranged in a manner that is consistent with the plant-specific emergency operating procedures (EOPs). The ECCS pumps operate in parallel while taking suction from the RWST, and the charging and SI pumps take suction from the RHR discharge during recirculation. Additionally, these pumps can provide flow to the RCS through either the cold legs or the hot legs. The accumulators are aligned to provide rapid cooling through the cold legs. The CS pumps provide flow to the CS headers and can take suction from the RWST during injection and the sump during recirculation. E3-11

Enclosure 3 Risk Quantification Figure 3 VEGP Connection Diagram E3-12

6.2 Plant States A plant state represents each of the different plant modes of post-accident operation and is defined by the activity state of the valves and pumps in the connection diagram. The plant states are consistent with the plant EOPs and are a function of the initiating event. For VEGP, two procedures mandate unique states as a function of break size and break side (i.e., hot leg vs. cold leg side). These procedures address SI accumulator injection and CS activation. As discussed in Section 13.2, the SI accumulators were modeled to only inject for breaks greater than or equal to 2 inches. Also as discussed in Section 13.2, the containment sprays were modeled to activate for hot leg breaks larger than 15 inches and operate for a duration of 24 hours. Note that different states are required for cold leg (CL) recirculation and hot leg (HL) recirculation. During HL recirculation, the RHR and SI pumps are aligned to discharge into the HLs, while the charging pumps continue to discharge into the CLs. Because the charging pump alignment after switchover to HL recirculation does not have any effect on the overall results, the NARWHAL model was simplified to align all ECCS pumps to discharge into the HLs during HL recirculation. Accident mitigation at VEGP was modeled using the states shown in Table 3-1 and Table 3-2. Note that Table 3-1 describes the initiators that dictate which breaks enter a given state, and Table 3-2 describes the operating equipment during each state. In Table 3-1, the term "Full Recirc" is used to describe the state in which all active pumps are taking suction through the sump strainers (or in piggy-back mode). Table 3 Plant State Initiators Min Break Max Break Break State Name Initiating Variable Size Size Side Injection (small break 0 1.999 Both Time= 0 minutes no CS) Injection 1.999 15.001 Both Time= 0 minutes (2<Break<15, no CS) Injection 15.001 31.001 Cold Time= 0 minutes (CL Break> 15, no CS) Injection (CS) 15.001 31.001 Hot Time= 0 minutes ECCS Switchover Level RHR Recirc (CS) 15.001 31.001 Hot in RWST (CS active) Full Recirc ECCS Switchover Level 0 15.001 Both (Break < 15, no CS) in RWST (CS inactive) Full Recirc ECCS Switchover Level 15.001 31.001 Cold (CL Break> 15, no CS) in RWST (CS inactive) CSS Switchover Level in Full Recirc (CS) 15.001 31.001 Hot RWST Hot Leg Switchover 0 15.001 Both Time= 450 minutes (Break< 15, no CS) Hot Leg Switchover 15.001 31.001 Cold Time = 450 minutes (CL Break > 15, no CS)

Enclosure 3 Risk Quantification Min Break Max Break Break State Name Initiating Variable Size Size Side Hot Leg Switchover 15.001 31.001 Hot Time= 450 minutes (CS) CS Termination 0 31.001 Both Time= 1440 minutes Table 3 Plant State Component Activity State Name Component Activity Description RHR Pumps Active Suction from RWST SI Pumps Active Suction from RWST Injection (small break Chan::iing Pumps Active Suction from RWST no CS) CS Pumps Inactive Not Operating SI Accumulators Inactive Not Operating Reactor Vessel Flow CL N/A RHR Pumps Active Suction from RWST Injection SI Pumps Active Suction from RWST (2<Break<15, no CS) Charging Pumps Active Suction from RWST Injection CS Pumps Inactive Not Operating (CL Break>15, no CS) SI Accumulators Active Injection into CL Reactor Vessel Flow CL N/A RHR Pumps Active Suction from RWST SI Pumps Active Suction from RWST Charging Pumps Active Suction from RWST Injection (CS) CS Pumps Active Suction from RWST SI Accumulators Active Injection into CL Reactor Vessel Flow CL N/A RHR Pumps Active Suction from Sump SI Pumps Active Suction from RWST Charging Pumps Active Suction from RWST RHR Recirc (CS) CS Pumps Active Suction from RWST SI Accumulators Inactive N/A Reactor Vessel Flow CL N/A RHR Pumps Active Suction from Sump Full Recirc SI Pumps Active Suction from RHR (Break<15, no CS) Charging Pumps Active Suction from RHR Full Recirc CS Pumps Inactive Not Operating (CL Break>15, no CS) SI Accumulators Inactive N/A Reactor Vessel Flow CL N/A Full Recirc (CS) RHR Pumps Active Suction from Sump E3-14

Enclosure 3 Risk Quantification State Name Component Activity Description SI Pumps Active Suction from RHR Charging Pumps Active Suction from RHR CS Pumps Active Suction from Sump SI Accumulators Inactive N/A Reactor Vessel Flow CL N/A RHR Pumps Active Suction from Sump Hot Leg Switchover SI Pumps Active Suction from RHR (Break<15, no CS) Charging Pumps Active Suction from RHR Hot Leg Switchover CS Pumps Inactive Not Operating (CL Break>15, no CS) SI Accumulators Inactive N/A Reactor Vessel Flow HL N/A RHR Pumps Active Suction from Sump SI Pumps Active Suction from RHR Hot Leg Switchover Charging Pumps Active Suction from RHR (CS) CS Pumps Active Suction from Sump SI Accumulators Inactive N/A Reactor Vessel Flow HL N/A RHR Pumps Active Suction from Sump SI Pumps Active Suction from RHR Charging Pumps Active Suction from RHR CS Termination CS Pumps Inactive Not Operating SI Accumulators Inactive N/A Reactor Vessel Flow HL N/A 6.3 Random Equipment Failures Random equipment failures are defined as failure to start or failure to run due to issues unrelated to GSl-191. Based on symmetry and the inputs used in the VEGP NARWHAL model, pump failures in one train are analytically identical to the same pump failures in the other train. In addition, although there is a slight difference during the RWST injection phase, the VEGP NARWHAL CFP calculation showed that the GSl-191 CFP results (i.e., the breaks that fail) were identical for the cases with no equipment failures, a single charging pump failure, and a single SI pump failure, because the charging and SI pumps piggyback off of the RHR pumps during recirculation. The switchover of the CS pumps from RWST injection to sump recirculation requires a manual operator action. Due to the human failure probability associated with this action, the probability of losing both CS pumps at the start of recirculation is higher than the probability of a single CS pump randomly failing to start or failing to run. E3-15

Enclosure 3 Risk Quantification Table 3-3 shows the functional failure probabilities for the different equipment configurations from the VEGP GSl-191 PRA model. Note that these overall probabilities are based on the logic described above and have been normalized to 100 percent. Table 3 Functional Failure Probabilities Equipment Configuration Functional Failure Probability No Equipment Failures 91.50% 2 CS Pump Failures 5.31% 1 RHR Pump Failure 1.46% 1 CS Pump Failure 1.26% 1 RHR Pump+ 1 CS Pump Failures 0.39% 1 RHR Pump+ 2 CS Pump Failures 0.07% Total 100% Due to the low functional failure probability, the case with one RHR pump and two CS pump failures was not evaluated explicitly, and the CFP was conservatively set to 1.0 for the risk quantification. 6.4 LOCA Frequencies Table 3-4 shows the mean LOCA frequencies taken from the VEGP GSl-191 PRA model. As discussed in Enclosure 1, Section 3.0, the medium and large LOCA frequencies are based on the geometric aggregation from NUREG-1829 (Reference 3). Table 3 Mean LOCA Frequencies from GSl-191 PRA Model Break Exceedance Size Frequency (yr- 1) 0.375 4.73E-04 2 1.39E-04 6 1.85E-06 31 1.50E-08 Note that the frequencies were used to calculate conditional failure probability as a function of PRA initiating event break size ranges. There are three break size ranges evaluated in the VEGP PRA. The small LOCA category comprises random breaks in the RCS in the range of 3/8-inch to 2-inch equivalent diameter. The medium LOCA initiating event is defined as a break in the RCS that is greater than or equal to 2 inches and less than 6 inches equivalent diameter. The large LOCA initiating event is defined as a break in the RCS that is greater than or equal to 6 inches up to a DEGB of the largest pipe (31 inches) in the RCS. Because the equivalent break size for a 31-inch DEGB is 43.8 inches (31 *-'12 =43.84), the frequency for each PRA size category is: E3-16

Enclosure 3 Risk Quantification

• FsLocA =Fo.375" - F2" =4.73E 1.39E-04 =3.34E-04 yr-1
• FMLOCA =F2" - Fa" =1.39E 1.85E-06 =1.37E-04 yr-1
• FLLOCA =Fa" - F43.B" 1 =1.85E 0 =1.85E-06 yr- 1 6.5 Water Volume and Level The height of the pool is a function of the total quantity of water in the sump. The pool level is calculated as a linear function of pool volume in gallons.

Hpool -- 1.20071 (ft)

• 10 -5 gal
• Vpool - 0.058 ft Equation 1 Nomenclature:

Hpool =the height of the sump pool in ft Vpoo1 = the volume of the recirculation pool in gallons At VEGP, three sources of water contribute to the recirculation pool inventory. These sources are the RWST, the SI accumulators, and the RCS.

• The total quantity of water delivered from the RWST is the difference between the initial and final levels. There are two final levels that are important in the analysis, the low-low level alarm (ECCS recirculation level) and the empty alarm (CS recirculation level). Note that this is different for breaks that do not activate containment sprays. The empty level alarm is the ECCS recirculation level for these breaks.
• Four SI accumulators provide immediate cooling to the core for breaks large enough to rapidly depressurize the RCS.
• A portion of the water initially in the RCS will be released as steam or spill to the pool through the break opening at the beginning of a LOCA.

Several hold-up volumes reduce the sump pool height:

• The amount of water held up as steam (vapor) in the containment atmosphere is a function of time.
• During recirculation, the total hold-up in the RCS is a function of break size and elevation. *
• The containment spray falling through the containment building represents a transitory hold-up volume. Note that this hold-up was only applied while containment spray pumps were operating. After containment spray was terminated, this quantity was returned to the sump pool.

1 Because the exceedance frequency for a 43.8-inch break is not available, the frequency was set to zero, which conservatively maximizes the frequency for large LOCAs. E3-17

Enclosure 3 Risk Quantification

• The break flow falling through containment also represents a transitory hold-up volume that was applied to all breaks.
• The volume of the CS discharge piping is another hold-up that was applied for breaks that initiate containment sprays.
• Other miscellaneous hold-up volumes include the containment sump pits, the elevator pit, and the containment floor drains.
• The reactor cavity and in-core tunnel is one of the largest hold-up volumes. Due to the restricted flow paths into the reactor cavity, this hold-up volume was applied in a time-dependent manner based on the break location and whether containment sprays are activated.

The long-term water level is high enough to fully submerge the RHR and CS strainers for all breaks. For some reactor cavity breaks, there is a short period where the RHR strainers are not fully submerged just after the RHR pumps are switched to recirculation. However, for these breaks, the water level rises enough to submerge the strainers before the CS pumps finish drawing down the RWST. 6.6 Flow Rates The pump flow rates that were used in the VEGP NARWHAL model are design flow rates for the SI pumps, charging pumps, and CS pumps. A flow rate approximately 20 percent higher than the design value was used for the RHR pumps. The break flow rate is the sum of the flow through the RHR, SI, and charging pumps during RWST injection, and is the sum of the RHR pump flow rates during recirculation (when the charging and SI pumps are piggybacking off of the RHR pumps). No credit was taken for reduced RHR flow rates for smaller break sizes. Note that for secondary side breaks in a feed and bleed scenario, only the charging and containment spray pumps were assumed to be active, which results in a significantly lower flow rate. As discussed in Section 6.5, the reactor cavity and in-core tunnel have restricted flow paths. Because of the restricted flow paths and the fact that the volume is fairly large, a function was implemented to model the flow rate into the reactor cavity as a function of time. Note also that the flow rate into the reactor cavity is a function of break location. For the breaks in the reactor cavity, the entire cavity is assumed to fill before the start of recirculation. In addition, the total reactor cavity hold-up volume is larger due to the flow rate inside of the cavity and the height of the flow paths that connect the cavity to the sump. For breaks outside the reactor cavity, the cavity would fill relatively slowly and would not be completely filled until after the start of recirculation. E3-18

Enclosure 3 Risk Quantification The core boil-off flow rate used in the NARWHAL model was calculated based on the core power and the ANSl/ANS-5.1-1979 decay heat curve (Reference 4). An additional 20 percent margin was added to the boil-off flow rate. 6.7 Pressure and Temperature The sump temperature, containment temperature, and containment pressure profiles were used in NARWHAL to determine time-dependent thermal-hydraulic properties. Design basis temperature profiles were used for both sump temperature and containment temperature. These temperature profiles were based on a DEGB in the primary loop piping with minimum safeguards, and therefore represent the maximum temperature profiles. Although the actual temperature profiles would be significantly lower for smaller break sizes, the same temperature profiles were conservatively used for all break sizes. Note that there are competing factors associated with sump temperature, which could result in a lower temperature resulting in more failures. However, based on sensitivity analysis (see Section 14.2.2), it was determined that maximizing the temperature is more conservative for the VEGP model. Containment accident pressure was not credited for NPSH margin calculations in the VEGP NARWHAL model. Because the technical specification minimum containment pressure is -0.3 psig at VEGP, the containment pressure profile was specified to be saturation pressure at pool temperatures above 210.96 degrees F, and 14.396 psia at pool temperatures below 210.96 degrees F. For the purpose of degasification and flashing calculations, up to 3.5 psi of accident pressure was credited. Both flashing and degasification are most problematic when the sump temperature is near or above 212 degrees F. The pool temperature is greater than 212 degrees F for approximately the first 120 minutes after a LOCA. Additional details are provided in Enclosure 2 Sections 3.f.14 and 3.g.14. 6.8 Sump and Spray pH The maximum sump recirculation pH was conservatively rounded up to 7.8 for the VEGP NARWHAL model. Use of the maximum pH provides bounding chemical release quantities from submerged materials. The minimum sump pH was rounded down to 7.0 for the VEGP NARWHAL model. The minimum sump pH was used to calculate the aluminum solubility limit in NARWHAL. Note that using different pH values to calculate release and solubility results in an over-prediction of the actual precipitate quantity. The spray pH during RWST injection was specified to be 5.72, which is the maximum pH of the RWST. Using the maximum RWST pH maximizes chemical release from E3-19 I

Enclosure 3 Risk Quantification unsubmerged sources during the injection phase. After switchover to recirculation, the spray pH is equivalent to the sump pH. 7.0 Debris Sources As described in Enclosure 2, Section 3.a.3, the types of debris in the NARWHAL model include Nukon fiberglass insulation, lnteram fire barrier, and qualified epoxy and IOZ coatings debris generated inside the ZOI. The model also includes unqualified epoxy, IOZ, and alkyd coatings, latent dirt/dust and fiber, and miscellaneous debris, which are present in containment or generated by the post-accident environmental conditions. The debris sources generated inside the ZOI range from essentially zero debris for the smallest break sizes up to 2,229 ft 3 of Nukon, 60 lbm of lnteram, 220 lbm of qualified epoxy, and 65 lbm of qualified IOZ debris for the bounding breaks. The other debris sources are independent of the break location and size and were therefore applied to all breaks. The debris quantities used in the NARWHAL model (including some operating margin) were 2,729 lbm for unqualified epoxy, 56 lbm for unqualified IOZ, 59 lbm for unqualified alkyd, 30 lbm for latent fiber, 170 lbm for latent dirt/dust, and 50 ft2 for miscellaneous debris. A four-category size distribution (fines, small pieces, large pieces, and intact blankets) was used for the Nukon debris based on the guidance in the appendices of NEI 04-07 Volume 2 (Reference 2). A more conservative two-category size distribution (fines and small pieces) was used for the lnteram debris. All of the qualified and unqualified coatings debris was conservatively treated as fines. The latent debris was also treated as fines. The miscellaneous debris was simply treated as a reduction in the total strainer surface area. Note that 25 percent overlap of the miscellaneous debris was credited in the NARWHAL model, which is consistent with the guidance in NEI 04-07 Volume 2 (Reference 2). Debris from all sources was essentially treated as being generated at the beginning of the event. The miscellaneous debris surface area reduction was applied prior to other debris transporting to the strainer. Unqualified coatings were modeled as failing after the pool fill phase (no transport to inactive cavities), but were available to transport at the start of recirculation. This is described further in Section 9.0. 8.0 Chemical Effects As described in Enclosure 2 Section 3.o.2, the formation of chemical products was analyzed as a function of the temperature, pH, and pool volume inputs, as well as debris quantities and exposed aluminum and concrete surface areas. There are two parts to the chemical product generation model: the elemental chemical release from materials in containment, and chemical precipitate formation. Note that these E3-20

Enclosure 3 Risk Quantification processes are based on break-dependent conditions and were therefore analyzed separately for each postulated break. 8.1 Elemental Chemical Release The Nukon and lnteram debris both contribute to chemical release, which was quantified in NARWHAL using the WCAP-16530 release equations (Reference 6) and the break-specific debris quantities. Note that lnteram debris only releases silicon. Therefore, the lnteram has no effect on the chemical product generation because the only aluminum precipitate that is tracked in the VEGP NARWHAL model is sodium aluminum silicate (SAS) (see Section 8.2), and NARWHAL conservatively assumes an infinite source of silicon when SAS is the only aluminum precipitate tracked (Reference 7). The amount of elemental chemical release from a given debris source is limited by the quantity of that element within the source. E-Glass (which includes Nukon) has 1.95 percent aluminum and 2.16 percent calcium by mass. The exposed surfaces include aluminum metal and concrete, which would either be submerged in the containment pool or exposed to containment sprays. The same surface areas were analyzed for each break. The chemical release from exposed concrete was evaluated using the WCAP-16530 release equations (Reference 6), and the chemical release from aluminum was evaluated using the University of New Mexico (UNM) release equations (Reference 8). 8.2 Chemical Product Formation The chemical precipitates analyzed in the VEGP NARWHAL model were SAS and calcium phosphate. The calcium phosphate was modeled in NARWHAL as precipitating immediately when calcium is released in solution (Reference 7). The SAS precipitates when the concentration of aluminum in the pool exceeds the aluminum solubility limit as calculated with the Argonne National Laboratory (ANL) solubility equation (Reference 7). Note that if precipitation of SAS was not predicted before 24 hours, then precipitation was forced at that time. Also note that aluminum was assumed not to remain dissolved in the pool after precipitation occurred (i.e., the aluminum solubility limit was only credited for precipitate timing). Forcing precipitation at 24 hours, as well as not taking credit for aluminum remaining dissolved in the pool, are conservative factors in the chemical product formation model. The effects of the chemical precipitates on strainer head loss are described in Section 10.0, and the effects on core blockage are described in Section 12.0. E3-21

Enclosure 3 Risk Quantification 9.0 Debris Transport As described in Enclosure 2, Section 3.e, debris transport includes the transport of debris during the blowdown, washdown, pool fill, and recirculation phases, as well as debris erosion. Transport can vary significantly as a function of flow rate, water level, etc. However, in the VEGP NARWHAL model, many of these parameters were not explicitly modeled and were conservatively represented using bounding conditions (e.g., the bounding flow rates for large break conditions were used to calculate recirculation transport fractions that were applied to breaks of all sizes). The specific factors affecting transport that were included in the VEGP NARWHAL model include debris type and size, break location (i.e., breaks in the steam generator compartment on the Loop 1/3 side, breaks on the Loop 2/4 side, breaks in the reactor cavity, breaks in the pressurizer compartment, or breaks in the annulus), whether sprays are initiated or not, and whether one or two trains are operating. The blowdown transport fractions are a function of the break location, as well as the size of debris. The only debris transported during the blowdown phase would be debris generated inside the ZOI. Fine debris was transported with the blowdown flow, with no credit for retention on structures. The transport of small and large pieces of Nukon and small pieces of lnteram was dependent on the break location as well as the location of grating that debris would have to be blown through to reach upper containment or the containment floor. The washdown transport fractions were based on containment spray initiation as well as the size of debris. If containment sprays were initiated, 100 percent of fine debris was modeled as being washed down to the containment floor. Some credit was taken for small pieces of Nukon and lnteram being retained in upper containment, and most large pieces of Nukon that were transported to upper containment were modeled as being retained in upper containment. If containment sprays were not initiated, the transport would be significantly reduced. However, 10 percent of fine debris was still modeled as washing down from upper containment due to condensation drainage. For pool fill transport, a relatively small fraction of debris was modeled as transporting to inactive cavities and the ECCS strainers as the sump cavities were filled. These transport fractions were applied to all debris that was in the containment pool at the end of the blowdown phase. This includes debris generated inside the ZOI as well as latent debris, but not unqualified coatings. Recirculation transport fractions were developed based on computational fluid dynamics (CFD) modeling. Several simulations were run to determine the transport fractions for the various types and sizes of debris corresponding to different break locations, number of trains operating, and whether containment sprays were initiated. Small and large pieces of Nukon and lnteram debris that are retained in upper containment would be subject to erosion due to containment sprays. A one percent E3-22

Enclosure 3 Risk Quantification erosion fraction was used for this debris for breaks where containment sprays were initiated. Similarly, small and large pieces of Nukon and lnteram debris in the containment pool would be subject to erosion and a 10 percent erosion fraction was used. Because the debris size is important with respect to penetration, NARWHAL applies the pool erosion for both the debris that settles in the pool as well as the debris that transports to the strainers (Reference 7). This conservatively maximizes the quantity of fine debris. Each of the transport processes described above were used to determine the total quantity of debris that reaches the strainer. However, the timing was not assumed to be instantaneous for all of these processes. Slowdown was treated as an instantaneous process at the beginning of the event. Washdown was treated as occurring after the inactive and sump cavities were filled during the pool fill phase, but before the start of recirculation. Debris that was transported to the ECCS strainers during the pool fill phase was modeled on the strainers at the start of recirculation (note that a fraction of the fine fiber debris penetrates prior to the start of recirculation as described in Section 11.0). For VEGP, failure time was not credited for unqualified coatings, and all of the unqualified coatings debris was treated as being in the pool at the start of recirculation. Although erosion is a time-dependent process, all of the erosion fines were treated as being generated at the start of recirculation. The actual transport to the strainers during the recirculation phase was modeled as a time-dependent process, where debris arrives on the strainers as a function of the pool turnover time (i.e., as a function of the pool volume and strainer flow rates). The debris accumulation on each strainer was proportional to the flow split to each strainer. This flow split was evaluated at each time step. Therefore, the relative accumulation of debris changes over time when various pumps were switched from RWST injection to sump recirculation, containment sprays were secured, etc. 10.0 Strainer Debris Impact Evaluation As discussed in Enclosure 2 Section 3.f.4, the head loss associated with debris accumulation on the strainer was determined using the results of prototypical strainer module testing. There are a total of four separate sump strainer assemblies for each unit at VEGP. There are two for the CSS (i.e., CS strainers) and two for the ECCS (i.e., RHR strainers). Each RHR and CS strainer assembly consists of four parallel vertical stacks connected to a plenum installed over each sump pit. The modified height of each RHR strainer (not including the curb) will be approximately 3.77 ft, and the effective surface area will be 677.6 ft2. The height of each CS strainer (not including the curb) is approximately 3.3 ft, and the effective surface area is 590 ft2. E3-23

Enclosure 3 Risk Quantification 10.1 Strainer Head Loss The conventional and chemical head loss values were corrected based on the sump thermal-hydraulic conditions compared to the test conditions. As specified in the March 2008 Nuclear Regulatory Commission (NRC) guidance (Reference 9), flow sweep data was used to develop the flow and temperature scaling. This scaling was performed at each time step using the time-dependent approach velocity and temperature in the VEGP NARWHAL model. Additionally, the test results were extrapolated to 30 days in accordance with the March 2008 NRC guidance (Reference 9) to account for chemical head loss that had not leveled off by the end of the test. For convenience, the 30-day head loss extrapolation value (which represents a gradual increase over time) was instantaneously applied at 7.5 hours in the VEGP NARWHAL model. Note that the head loss extrapolation value was also corrected based on the time-dependent approach velocity and temperature. The bounding clean strainer head loss of 0.375 ft at 4,500 gpm was used in the VEGP NARWHAL model for all cases. NARWHAL uses a rule-based approach to calculate head loss based on the results of head loss testing with a prototypical strainer module. For the VEGP NARWHAL model, if the fiber debris load on the strainer was less than the tested quantity from the thin bed test, then the maximum thin bed conventional head loss (0.625 ft) was returned. If the fiber quantity is greater than what was tested in the thin bed test, then the maximum full load conventional head loss (5.46 ft) was returned. The head loss effects of calcium phosphate and SAS were each analyzed separately from the conventional debris head loss. Chemical head loss was only applied if the fiber debris quantity on the strainer was greater than 0.45-inch equivalent (see , Section 3.f.10). If this condition was met, and any calcium phosphate accumulated on the strainer, the head loss corresponding to the full quantity of calcium phosphate debris (1.11 ft) was added. Similarly, given the accumulation of sufficient fiber and any SAS on the strainer, the head loss corresponding to the full SAS debris quantity (5.24 ft) was added. In addition, because the extrapolation constant was associated with the chemical head loss, this constant was only applied if the fiber bed was thick enough for the chemical head loss to be added (i.e., greater than 0.45 inches). Because the head loss results are only applicable for debris quantities up to what was tested, debris limits were specified in the VEGP NARWHAL model corresponding to the tested quantities. Separate debris limits were specified for fiber, particulate, lnteram, calcium phosphate, and SAS debris. If any one of these debris limits were exceeded, the strainer was assumed to fail. E3-24

Enclosure 3 Risk Quantification 10.2 Degasification and Flashing NARWHAL calculates degasification and flashing based on the total strainer head loss and other important parameters (containment pressure, sump temperature, water level, etc.) (Reference 7). For the degasification calculations; 2.5 psi of accident pressure was credited. The midpoint of the strainer was used to calculate the average degasification across the entire strainer. NARWHAL performs the flashing calculation at the same reference elevation used to calculate degasification (i.e., the midpoint of the strainer in the VEGP model). However, because the pressure would be lower at the top of the strainer, using the midpoint of the strainer for the flashing calculations is equivalent to crediting additional accident pressure equal to the hydrostatic head from the top of the strainer to the midpoint (approximately 1 psi). Therefore, the VEGP NARWHAL model credits 2.5 psi accident pressure for degasification and approximately 3.5 psi accident pressure for flashing. Note that no accident pressure was credited for NPSH in the VEGP NARWHAL model. The strainer was automatically assumed to fail if any flashing occurs. The degasification calculation results in a gas void fraction that was compared against the pump limits (a void fraction limit of 0.02 was used for all pumps). If the void fraction exceeds a pump limit, the pump was assumed to fail. 10.3 Structural and NPSH Margin The NPSH margin is the NPSH available (excluding strainer head losses) minus the NPSH required. The NPSH available was calculated in the VEGP NARWHAL model based on the time-dependent containment pressure, sump temperature, water level, and major and minor losses in the pump suction piping. As discussed in Section 6.7, a bounding pressure and temperature profile was used, which conservatively minimizes NPSH available. The NPSH required was calculated as a function of the time-dependent flow rate based on the pump curves. The NPSH required was also adjusted as a function of the void fraction as described in RG 1.82 (Reference 10). The strainer head loss was compared against the strainer structural margin (24.0 ft) and the pump NPSH margin at each time step to determine whether a failure occurred. 11.0 Debris Penetration Evaluation As described in Enclosure 2, Section 3.n.1, fiber penetration correlations were developed based on testing with a prototypical strainer module. These equations were used to calculate the fine fiber quantity that passes through the strainer from both prompt penetration and longer-term shedding. A penetration fraction of 0.48 was also applied to the fine fiber that transports to the . strainers during pool fill. As shown in Enclosure 2, Section 3.n.1, this penetration fraction bounds the prompt penetration corresponding to clean strainer conditions. E3-25

Enclosure 3 Risk Quantification 12.0 Debris Penetration Effects As discussed in Enclosure 1, Section 1.0 and Enclosure 2, Section 3.m, ex-vessel downstream effects were addressed in a bounding manner and therefore were not included in the VEGP NARWHAL model. As described in Enclosure 2, Section 3.n.1, core blockage and boron precipitation were evaluated using assumed fiber debris limits and acceptance criteria. VEGP uses Westinghouse fuel and the reactor vessel has an upflow barrel/baffle design with pressure relief holes in the core plates that are not currently credited in the long-term core cooling analyses. Any debris that does not accumulate in the reactor vessel was modeled as automatically returning to the sump pool where it would be available to transport and potentially pass through the strainers again. 12.1 Cold Leg Breaks The fiber debris that penetrates the RHR strainers transports to the reactor vessel through the ECCS flow. During cold leg recirculation, a portion of the ECCS flow that enters the reactor vessel through the cold legs would travel through the core inlet as make-up for boil-off, while the rest of the cold-leg flow would spill from the break. The fraction of fiber debris that was caught on the core inlet was the ratio of the boil-off flow rate to the ECCS flow rate into the cold leg. Note that margin was added to the boil-off flow rate (see Section 6.6). Once switchover to hot leg recirculation had occurred, debris no longer accumulated at the core inlet. Instead, any penetrated debris that entered the reactor vessel was captured incore. However, by the time hot leg recirculation was initiated, most of the fiber fines had bee.n captured on the strainers, and very little additional fiber was transported to the core. Any fiber that was captured in the reactor vessel was assumed to remain there for the duration of the event. For cold leg breaks, an in-vessel failure was recorded if the core inlet fiber load was greater than the specified limit (see Enclosure 2, Section 3.n.1 ). 12.2 Hot Leg Breaks For hot leg breaks, a significant quantity of fiber could accumulate on both the core inlet as well as incore. Flow would either enter the core inlet or the alternate flow path based on the head loss due to debris-related resistance across the core inlet at the bottom core plate. To determine the flow split for the ECCS flow that entered the reactor vessel through the cold leg, several preliminary calculations were performed. Initially, a Ksplit variable was calculated based on an assumed function of ECCS flow. Next, the current Ktactor variable was calculated based on an assumed function of the fiber quantity on the core inlet and the presence of chemical precipitates. If chemicals E3-26

Enclosure 3 Risk Quantification had already precipitated, then the current Ktactor was set to a very high value. Similarly, no matter the time, if the quantity of fiber on the core inlet was greater than an assumed threshold then the current Ktactor was also set to a very high value. Note that in the context of core fiber accumulation, the chemical precipitation timing refers to the time at which aluminum precipitation occurs (see Section 8.2). For all other conditions, the Ktactor was calculated based on an assumed piecewise function of the current fiber debris load. Using the !<split and Ktactor values, an msplit variable was calculated. If the Ktactor was less than l<split, then msplit was set to 0. If the Ktactor was very high, then the msplit was set to 1. If neither of these conditions were met, then the msplit variable was calculated based on an assumed function of the Ktactor and Ksplit values. Note that an assumed maximum curve fit was used, which results in the minimum msplit value. With these variables calculated, the fraction of debris that was caught on the core inlet and within the core for a hot leg break was calculated using the following equations. Qcold leg Core Inlet (HL Break) = ( 1 - ffisplit ) *------- Qhot leg + Qcold leg Qcold leg Qhot leg lncore (HL Break) = msplit * +------- Qhot leg + Qcold leg Qhot leg + Qcold leg Following switchover to hot leg recirculation, all of the fiber carried with the ECCS flow was modeled as accumulating in the core. For hot leg breaks, an in-vessel failure was recorded if any of the following failure criteria were met:

1. The calculated Ktactor exceeded the specified limit before the specified tb1ock time.

2. The incore fiber load was greater than the specified limit.

3. The reactor vessel fiber load (sum of the incore and core inlet fiber quantities) was greater than the specified limit.

See Enclosure 2, Section 3.n.1 for additional details on the hot leg break failure criteria. 13.0 Sub-Model Integration The following flow diagrams show an overview of the analytical models that were used to determine how water was transported, how the conventional debris was generated and transported, how chemical precipitates formed and transported, how the strainer failure criteria were analyzed, and how the core failure criteria were analyzed. All of these models were addressed holistically in a time-dependent manner. E3-27

Enclosure 3 Risk Quantification

5. Water held-up in the

                                                                         ,-4   RCS or other geometric hold-up volumes

2. ECCS pumps provide .__

                             .....-     water to cool the core

1. LOCA occurs, 7. RWST is drained, injection from the RWSTbegins

                       --                                               _____. 4. Water spills directly to the sump recirculation from the sump begins

3. CSS pumps are used

                             '---I                               >---

to cool containment

6. Water is held-up in

                                                                          .__.      transitory and geometric hold-up volumes Figure 3 Water Transport and Accumulation

3. Debris quantities outside ZOI

{unqualified coatings, latent, miscellaneous)

4. Blowdown 5. \Vashdown

2. ZOI debris quantities

1. Select unique transport to transport from from BADGER break location, size, ~ }----? containment ~

H containment (insulation and and orientation qualified coatings) compartments and sump pool compartments to sump pool

6. Pool fill transport 1

7. Recirculation from sump pool to L'

transport from sump strainers and inactive pool to strainers cavities l

9. Debris accumulation S. Debris penetration on core ~- through strainers Figure 3 Debris Generation and Transport E3-28

Enclosure 3 Risk Quantification

1. Corrosion/

dissolution of metals, ~

3. Precipitate solubility concrete, and debris limit byes II

2. Corrosion/

dissolution of metals, ' 4. Formation of concrete, and debris chemical precipitates in sump pool

5. Recirculation transport from sump <E------

pool to strainers

                                                                                \[/

6. Debris penetration through strainers

                                                                                                   - 7. Debris accumulation on core Figure 3 Chemical Product Formation and Transport E3-29

Enclosure 3 Risk Quantification O.Doe

14. Does quantity No quantity No

9. Strainer debris exceed 13. Core debris exceed accumulation tested accumulation blockage debris limits?

limits? Yes 12. Pass Yes 16. Pass core Strainer criteria criteria

11. Fail Strainer criteria 15. Fail core criteria Figure 3 Comparison of Strainer Head Loss and Core Debris Loads to Failure Criteria Any failures that occurred were binned as strainer failures or core failures, and these results were used to calculate the GSl-191 CFP values as described in more detail below. The NARWHAL software, which was used to integrate all of the sub-models, is described in Section 13.1, and Section 13.2 provides a summary of the assumptions used in the VEGP NARWHAL evaluation.

In order to calculate the CFP values, the following steps were taken:

1. GSl-191 failures were computed for each break as described above. For VEGP, the failures for input into the PRA were computed in the following 12 categories (Strainer A and B correspond to the RHR strainers):

a. Small breaks

i. Core failures E3-30

Enclosure 3 Risk Quantification ii. Strainer A and Strainer B failures (without core failures) iii. Strainer A failures (without core or Strainer B failures) iv. Strainer B failures (without core or Strainer A failures)

b. Medium breaks

i. Core failures ii. Strainer A and B failures (without core failures) iii. Strainer A failures (without core or Strainer B failures) iv. Strainer B failures (without core or Strainer A failures)

c. Large breaks

i. Core failures ii. Strainer A and B failures (without core failures) iii. Strainer A failures (without core or Strainer B failures) iv. Strainer B failures (without core or Strainer A failures)

2. Overall plant-wide LOCA frequencies were allocated to individual welds and break sizes using a top-down LOCA frequency methodology.

a. Plant-wide LOCA frequencies were based on the small, medium, and large break frequencies in the VEGP GSl-191 PRA model with log-linear interpolation for intermediate break sizes.

b. The frequency for a given break size was allocated to individual welds (that can experience a break of that size).

3. The PRA model category for large breaks was broken up into size ranges to more accurately calculate the overall CFP. Smaller breaks within a given size range were assumed to have the same probability as larger breaks within the size range.

4. The CFP for a PRA category (e.g., large breaks) was calculated based on the combined CFP and LOCA frequency weight for each size range.

Figure 3-7 shows an example of how the size ranges were used to calculate the strainer CFP for the case with no random equipment failures. In this example, the overall frequencies result in a probability weight of 82.3 percent in Size Range 1, 15.1 percent in Size Range 2, and 2.6 percent in Size Range 3. The corresponding strainer CFP values are 0.0 percent for Size Range 1, 3.0 percent for Size Range 2, and 27.8 percent for Size Range 3. Therefore, the overall strainer CFP for large* breaks is 1.2 percent as shown below: CFPLarge = 0.823

• 0.000 + 0.151*0.030+0.026
• 0.278 = 0.012 E3-31

_ _ _ _ _J

Enclosure 3 Risk Quantification Size Range Definition 1.E-02 0.9 1.E-03 0.8 Large Large La rge Size Rangel Size Range 2 Size Range 3 1.E-04 0.7

  ~                      P{SR,)=82 .3%                               P(SR2)=15.1%                                    P(SR1 )=2.6%
  ~                      CFP(SR 1)=0.06                          CFP(SR )=3.0                                   CFP(SR )=27 8          0.6 u   1.E-05 c

J CT 0.5 ~

   ~

u. 1.E-06 CFP(Large)=l.2% 0.4 50 1.E-07 0.3 02 l.E-08 0.1 1.E-09 0 0 .5 2 4 6 8 10 12 14 15 16 17 18 19 20 21 22 23 24 25 26 27 27.5 28 28.S 29 29.5 30 30.5 31 Break Size (Inches)

                       -   RHR StrainPr Ostbri'i limit failurpc: tAfl Pump   Av~il.1blel    -   VfGP PRA Mf'<ln lOC'A fteflUPncv Figure 3 Illustration of CFP Calculation for Large Breaks Based on Three Size Ranges 13.1 NARWHAL Softwa re NARWHAL is an object-oriented program that models the connections between important plant components (pumps , strainers, tanks , etc.) based on user-defined inputs . The software performs mass balance calculations that determine the time-dependent quantity of water, debris , and chemical solutes associated with each physical object. Using these time-dependent quantities along with other user-specified conditions , each aspect of GSl-191 can be evaluated in an integrated manner. The software can be used to simulate a single break or many thousands of breaks to evaluate the risk associated with GSl-191 .

At any given time during the simulation , the state of the plant can be defined by a fixed set of parameters (i.e., the on/off states of components , the quantity of debris stored by components , etc.). This collection of parameters is called the state vector. NARWHAL updates the state vector by determining the amount of change in each variable given a change in time. The amount of change in each variable is determined using a series of algorithms called "marching algorithms", which are graph traversal algorithms that maintain consistent flow through the plant system . For example , if a 10,000 gpm pump is fed by a pump that is limited to 1,000 gpm , the algorithms will determine that the high capacity pump can only pump 1,000 gpm . These algorithms are essentially a way for NARWHAL components to communicate information to one another. E3-32

Enclosure 3 Risk Quantification A single NARWHAL simulation relies on a series of marching algorithms (Figure 3-8). The algorithms were designed to handle generic configurations of components, meaning that the user can design arbitrary networks as long as the configuration is valid (i.e., there is a source, active pumps, and a sink). The first algorithm, the activity march, exists solely to determine the health of the network and its components. This algorithm is responsible for detecting valid and invalid paths of flow through the system. For instance, it is necessary to detect the failure of a pump if the strainer that feeds the pump fails. Conversely, this algorithm determines that a strainer will not receive flow if the pumps feeding from it shut down or fail. The second algorithm, the source march, determines that flow through the network and its components is consistent. This is important when, for instance, pump flow rates are a function of other pumps (e.g., piggy-backing). It is also important when considering flow restrictions in the system, such as break size dependent flow rates. The third algorithm, the water march, determines the flow rates and storage balances on all components in the network. This algorithm implements the mass balance equations inherent to the NARWHAL base component. After this algorithm has run, water is moved from the network sources, through all active components, to the network sinks. The fourth and final algorithm, the debris march, uses information generated by the previous algorithms to determine the mass balance of debris and chemicals in the network. This includes determining the release of chemicals, the movement of debris, the capture of debris, and the formation of precipitates. E3-33

Enclosure 3 Risk Quantification Simulation Initialization Reporting Activity March Yes Simulation No Source Termination? March Component Water March Tests Debris March Figure 3 NARWHAL's Basic Procedure In each time step , after the marching algorithms have run and a new state vector has been calculated , a series of tests are run against a number of failure criteria (i .e. , strainer structural margin , component debris limits , strainer submergence, etc.). If a failure occurs , it is noted in the results , but the simulation is allowed to continue running as if the failure had not occurred (e.g ., a pump that fails due to loss of NPSH is allowed to continue running normally for the remainder of the simulation). At the end of a simulation , NARWHAL reports the outcome of the simulation in one of three ways . If a single break simulation is being run , NARWHAL outputs a list of time-dependent vectors representing a number of variables including core debris quantities, strainer debris quantities, component failure states, and component flow rates . If a bulk simulation is being run , NARWHAL will not report time vectors. Instead , it reports summary variables such as failure times , total debris on each strainer at the end of the simulation , maximum head loss for each strainer, and total fiber on the core at switchover to hot leg recirculation. In addition , descriptive break information is reported including the break location and size . In this mode , each time a simulation for a given break is completed , a new record containing the summary information is automatically entered into the results file. If a probability or sensitivity simulation is being run , NARWHAL outputs the summary CFP values for each bulk simulation . Additionally, descriptive information about the variables being modified is entered into the results file . In addition to calculating which breaks pass and fail the GSl-191 acceptance criteria , NARWHAL can also be used to post-process the results to calculate the CFP values based on the approach described in Section 13.0. E3-34

Enclosure 3 Risk Quantification 13.2 NARWHAL Model Assumptions The following assumptions were used for the VEGP NARWHAL model evaluation:

1. It was assumed that breaks less than 2 inches do not result in rapid, full depressurization of the RCS. Therefore, injection by the SI accumulators is not required for these breaks. This assumption is consistent with the minimum water level calculation.

2. The containment sprays were assumed to only be activated for hot leg breaks greater than 15 inches. Note that this includes all partial breaks and DEGBs greater than 15 inches on the hot legs; however, no failures on the cold or intermediate legs are assumed to actuate containment sprays. Although there is some uncertainty in which breaks initiate containment sprays, the relatively high containment pressure required to actuate sprays (21.5 psig) significantly reduces the likelihood that most breaks will exceed the set point and actuate containment sprays. The assumption that only very large hot leg breaks will initiate containment sprays is consistent with the results of best-estimate thermal-hydraulic modeling for a range of potential break sizes on the hot and cold leg piping. This modeling showed that a hot leg DEGB initiated containment sprays, while all other evaluated breaks (including a cold leg DEGB and partial 15-inch breaks on both the hot and cold legs) did not. Assuming that hot leg breaks greater than 15 inches activate containment sprays reasonably represents what was learned in the best-estimate thermal hydraulic modeling.

3. It was assumed that containment sprays would be secured at 24 hours. Although there is some uncertainty in the spray duration, this is a reasonably conservative assumption because sprays are required to operate for at least 2 hours once they are initiated, and running sprays longer than 2 hours would significantly increase the quantity of aluminum released from unsubmerged sources.

4. The RHR flow rate used in the NARWHAL CFP calculation was assumed to be 3,700 gpm. The design flow rate for the RHR pumps is 3,000 gpm. However, it is not expected that the actual plant flow rate would be this low. The use of a higher flow rate is generally conservative in terms of recirculation timing, flashing calculations, and head loss correction. The use of 3,700 gpm for the RHR flow rate is consistent with what was used in the deterministic NPSH calculation and the single train value used in the ECCS system head curve.

5. The minimum sump recirculation pH was assumed to be 7.0. This conservatively bounds the calculated minimum value of 7.12 and is consistent with the minimum acceptable pH documented in the buffer verification calculation. Note that the minimum sump recirculation pH was used to calculate the aluminum solubility limit in the NARWHAL model.

6. The spray pH during injection was assumed to be 5.72 as determined from the conservative maximum pH of the RWST. This maximizes chemical release from unsubmerged sources during injection.

E3-35

Enclosure 3 Risk Quantification

7. It was assumed that the breaks downstream of the first isolation valve do not need to be addressed in this quantification because they are not risk significant. This is a reasonable assumption because there would have to be a coincidental failure of the valve along with the pipe break, which is a low probability event. Additionally, there are no known quantities of localized problematic insulation types or any other factors that are unique to the isolable weld locations that would significantly increase the probability of debris-related failures.

8. The amount of latent debris documented in the debris generation calculation is 60 lbm. This value was conservatively increased to a total of 200 lbm in containment for operating margin.

9. The amount of miscellaneous debris documented in the debris generation calculation is 4 ft2. This value was conservatively increased to a total quantity of 50 ft2 in containment for operating margin.

1O. It was assumed that there is a total quantity of 926.6 ft2 of unsubmerged aluminum metal and 348.4 ft2 of submerged aluminum metal in containment. This includes some margin for future additions of aluminum, as described in the chemical product generation calculation.

11. The total amount of exposed concrete that would be submerged in the pool was assumed to be 10,000 ft2. This is a conservatively large surface area used to maximize the potential for chemical release and is consistent with the quantity used in the chemical product generation calculation. Note that the chemical product generation calculation did not evaluate the chemical effects associated with unsubmerged concrete due to the conservative quantity used for submerged concrete.

12. VEGP requires modifications to the existing RHR strainers to maintain long-term submergence during recirculation for all breaks. Two disks per disk stack will be removed from each of the RHR strainer assemblies to achieve submergence.

13. The containment pressure was assumed to be saturation pressure when the pool temperature is greater than 210.96 degrees F, and 14.396 psia after.the pool temperature has dropped below 210.96 degrees F. The pressure of 14.396 psia was calculated based on the minimum containment pressure of -0.3 psig per VEGP Technical Specification 3.6.4 and an atmospheric pressure of 14.696 psia. The temperature of 210.96 degrees F is the corresponding saturation temperature at 14.396 psia determined from linear interpolation. This is a conservative assumption because it only credits the accident pressure necessary to keep the pool as a liquid.

Note that a small amount of containment accident pressure was credited for degasification and flashing calculations (see Assumption 16).

14. It was assumed that head loss due to chemical precipitates is only applied once the theoretical fiber bed is greater than 0.45 inches. This is supported by the 2009 thin bed head loss test.

15. It was assumed that all random equipment failures evaluated for the different NARWHAL CFP evaluations occurred at the beginning of recirculation. This is a conservative assumption because it results in a quicker switchover to recirculation when compared to failure at the beginning of the event. Additionally, for CS pump E3-36

Enclosure 3 Risk Quantification and/or RHR pump failure cases, it results in more debris accumulation on the remaining active strainers.

16. For the purpose of degasification and flashing calculations, it was assumed that up to 3.5 psi of accident pressure would be available. As described in Assumption 13, no credit was taken for accident pressure in the NPSH available calculations. The basis for this assumption is described in more detail in Enclosure 2 Sections 3.f.14 and 3.g.14. Note that the actual accident pressure credited for degasification is 2.5 psi, and the accident pressure credited for flashing is 2.5 psi plus the pressure head from the top of the strainer to the midpoint of the strainer (approximately 1 psi).

17. It was assumed that the assumed maximum msplit curve fit for core blockage calculations should be used in the NARWHAL CFP evaluation. This is a conservative assumption because it maximizes the quantity of fiber that accumulates on the core inlet. Although the curve fit results in a higher msplit value initially, it results in a lower msplit value for greater resistances across the core inlet. This allows more overall fiber to accumulate at the core inlet for breaks that challenge the core inlet fiber limit.

18. The accumulators were assumed to not inject for any secondary side break. This is a reasonable assumption because secondary side breaks do not result in rapid depressurization of the RCS, which would trigger accumulator injection.

19. Both trains of containment spray were assumed to operate for all secondary side breaks. This provides a reasonable starting point for evaluating the secondary side breaks. To address the uncertainty in this assumption, several different equipment configurations were evaluated, including single CS pump failure and failure of both CS pumps.

20. For a secondary side break, the total flow to the ECCS was assumed to be provided by both charging pumps at a flow rate of 230 gpm/pump. This is a reasonable assumption because the PORV water-relieving capacity is 230 gpm per PORV.

Note that recirculation for secondary side breaks would occur due to overflow of the pressurizer relief tank through the two PORVs. The use of the maximum flow rate through the PORVs is conservative because it maximizes the flow split to the RHR strainers during recirculation. 21.All secondary side breaks were assumed to accumulate fiber in-vessel in the same way as a hot leg break. This is a reasonable assumption because the RCS would be bled through the pressurizer relief tank during a feed and bleed scenario. The pressurizer relief tank is connected downstream of the hot leg.

22. The general inputs used to calculate the CFPs for LOCAs were assumed to be applicable for secondary side breaks. These inputs include the following:

       - Thermal hydraulic inputs (water sources, flow rates, temperature profiles, pH, etc.)
       - Debris inputs (unqualified coatings quantities, debris transport fractions, latent debris, etc.)
       - Chemical precipitate debris
       - Strainer geometry
       - Strainer head loss
       - Strainer failure options E3-37

Enclosure 3 Risk Quantification

       - Strainer penetration equations
       - Core blockage equations This provides a reasonable result set of conditions to evaluate the risk impact of secondary side breaks.

23. It was assumed that a LOCA that occurs during full power operation (i.e., Mode 1) is equivalent or bounding compared to the other operating modes. This is a reasonable assumption because the RCS pressure and temperature (key inputs affecting the ZOI size) would either be approximately the same or significantly lower for Modes 2 through 6. In addition, the flow rate required to cool the core (a key input affecting core blockage) would be reduced significantly for low power or shutdown modes.

14.0 Systematic Risk Assessment As described in Section 4.0, the VEGP GSl-191 PRA model includes the necessary modifications to calculate the risk impact associated with the effects of debris on the strainers and in the core. No new human failure events due to the effects of debris were identified, and no human actions to mitigate the effects of debris were credited in the risk calculation. The only common inputs that were used in both the NARWHAL model and the GSl-191 PRA model are the LOCA frequencies and the equipment configurations. As discussed in Section 6.4, the GSl-191 CFPs were calculated using the same LOCA frequencies that were used in the GSl-191 PRA model. In addition, as discussed in Section 6.3, the high likelihood equipment configurations were explicitly evaluated in NARWHAL to calculate separate sets of CFPs for each high likelihood configuration. These CFPs were entered for the appropriate equipment configurations in the GSl-191 PRA model. Therefore, the VEGP NARWHAL model and GSl-191 PRA model are consistent. 14.1 VEGP NARWHAL.CFP Evaluation The VEGP NARWHAL CFP calculation showed that there were no small or medium break LOCAs that fail for any equipment configurations. Table 3-5 shows the NARWHAL CFP results for large break LOCAs for each equipment configuration. E3-38

Enclosure 3 Risk Quantification Table 3 NARWHAL CFP Results for Large Break LOCAs Equipment Strainer A Strainer A Strainer B Core Configuration and B only only No Equipment Failure 0 0.0118 0 0 RHR Pump B Failure 0 NIA 0.0679 NIA Charging Pump B Failure 0 0.0118 0 0 SI Pump B Failure 0 0.0118 0 0 Train B Failure 0 NIA 0.0736 NIA CS Pump B Failure 0 0.0139 0 0 Both CS Pumps Failure 0 0.0177 0 0 Table 3-6 and Table 3-7 show the CFP results for DEGBs on the feedwater line piping and main steam line piping, respectively. None of the feedwater line breaks produced a sufficient quantity of debris to fail, and main steam line breaks only failed under the equipment configuration where both CS pumps fail due to issues unrelated to debris. Table 3 Feedwater Line DEGB CFP Results Equipment Strainer A Strainer A Strainer B Core Configuration and B only only No Equipment Failure 0 0 0 0 Charging Pump B Failure 0 NIA 0 NIA CS Pump B Failure 0 0 0 0 Both CS Pumps Failure 0 0 0 0 Table 3 Main Steam Line DEGB CFP Results Equipment Strainer A Strainer A Strainer B Core Configuration and B only only No Equipment Failure 0 0 0 0 Charging Pump B Failure 0 NIA 0 NIA CS Pump B Failure 0 0 0 0 Both CS Pumps Failure 0 0.475 0 0 The total baseline risk (from internal events, internal fire, internal flood, and seismic events) for the VEGP Unit 1 PRA model is 4.39x10-5 per reactor-year (ry- 1) for GDF and

1. 73x10- 5 ry- 1 for LERF. The total baseline risk for the VEGP Unit 2 PRA model is 5.05x10- 5 ry- 1 for GDF and 1.90x1 o-6 ry- 1 for LERF. Using the CFP results described above, the change in risk calculated with the VEGP GSl-191 PRA model is shown in Table 3-8.

E3-39

Enclosure 3 Risk Quantification Table 3 VEGP Total Risk Impact due to GSl-191 Failures ACDF ALE RF Case (ry-1) crv-1> Risk increase from GSl-191 failures for high-likelihood 2.32x1Q*B 3.10x10- 11 LOCA configurations Bounding risk increase from GSl-191 failures for 1.41x10-9 4.09x10- 12 unlikely LOCA configurations Risk increase from GSl-191 failures for seismically-1.50x10-9 1.50x10-10 induced LOCAs Risk increase from GSl-191 failures for SSBls 1.39x10-9 8.25x10-11 Total risk increase associated with GSl-191 2.75x10-s 2.68x10-10 These CDF, LERF, flCDF, and flLERF values fall well within the RG 1.174 Region Ill guidelines. Therefore, the effects of debris have very low risk at VEGP. Table 3-9 provides a summary of information on each Class 1 ISi weld inside the first isolation valve. The results shown in this table (specifically the maximum transported fiber and whether a failure was observed at the weld) are based on the single train failure equipment configuration. Table 3 Weld Information List Max Weld Break Weld Transported Degradation Failure Compartment ID Side Identifier Fiber to RHR Mechanism  ? A Strainer (ft3 ) 11201-004-29 hot SG_1-4 639.75 D&C yes 6-RB 11201-001-29 hot SG_1-4 637.70 D&C yes 5-RB 11201-001-29 hot SG_1-4 620.24 D&C yes 3-RB 11201-004-29 hot SG_1-4 605.83 D&C yes 4-RB 11201-003-29 hot SG_2-3 482.44 D&C yes 5-RB 11201-002-29 hot SG_2-3 476.86 D&C yes 5-RB 11201-002-29 hot SG_2-3 473.54 D&C yes 3-RB 11201-003-29 hot SG_2-3 462.00 D&C yes 3-RB 11201-008-31 cold SG_1-4 196.90 D&C yes 1-RB 11201-005-31 cold SG_1-4 194.24 D&C yes 1-RB 11201-008-31 cold SG_1-4 186.47 D&C yes 2-RB E3-40

Enclosure 3 Risk Quantification Max Weld Break Weld Transported Degradation Failure Compartment ID Side Identifier Fiber to RHR Mechanism  ? A Strainer (ft3 ) 11201-005-31 cold SG_1-4 185.94 D&C yes 2-RB 11201-006-31 cold SG_2-3 168.92 D&C yes 1-RB 11201-005-31 cold SG_1-4 168.81 D&C yes 3-RB 11201-007-31 cold SG_2-3 167.20 D&C yes 1-RB 11201-008-31 cold SG_1-4 162.99 D&C yes 3-RB 11201-006-31 cold SG_2-3 158.19 D&C yes 2-RB 11201-007-31 cold SG_2-3 156.75 D&C yes 2-RB 11201-006-31 cold SG_2-3 150.85 D&C yes 3-RB 11201-001-29 hot Reactor Cavity 150.42 D&C yes 1-RB 11201-004-29 hot Reactor Cavity 148.06 D&C yes 1-RB 11201-V6-29 hot Reactor Cavity 148.06 D&C yes 001-W37-RB 11201-V6-29 hot Reactor Cavity 147.50 D&C yes 001-W36-RB 11201-003-29 hot Reactor Cavity 142.41 D&C yes 1-RB 11201-V6-29 hot Reactor Cavity 141.36 D&C yes 001-W33-RB 11201-002-29 hot Reactor Cavity 141.13 D&C yes 1-RB 11201-V6-29 hot Reactor Cavity 139.03 D&C yes 001-W40-RB 11201-005-31 cold SG_1-4 137.98 D&C yes 4-RB 11201-008-31 cold SG_1-4 137.56 D&C yes 4-RB 11201-007-31 cold SG_2-3 136.77 D&C yes 3-RB 11201-006-31 cold SG_2-3 131.69 D&C yes 4-RB 11201-008-31 cold SG_1-4 124.63 D&C yes 8-RB 11201-005-31 cold SG_1-4 124.57 D&C yes 8-RB E3-41

Enclosure 3 Risk Quantification Max Weld Break Weld Transported Degradation Failure Compartment ID Side Identifier Fiber to RHR Mechanism  ? A Strainer (ft 3 ) 11201-005-31 cold SG_1-4 123.83 D&C yes 7-RB 11201-008-31 cold SG_1-4 122.59 D&C yes 7-RB 11201-012-27.5 cold SG_1-4 122.33 D&C yes 1-RB 11201-009-27.5 cold SG_1-4 120.98 D&C yes 1-RB 11201-007-31 cold SG_2-3 114.03 D&C yes 4-RB 11201-007-30.5 cold SG_2-3 103.72 D&C yes 7-RB 11201-006-31 cold SG_2-3 103.70 D&C yes 7-RB 11201-006-31 cold SG_2-3 100.47 D&C yes 8-RB 11201-007-31 cold SG_2-3 100.21 D&C yes 8-RB 11201-011-27.5 cold SG 2-3 98.14 D&C yes 1-RB 11201-010-27.5 cold SG 2-3 97.10 D&C yes 1-RB 11201-053-12.814 hot SG 1-4 54.67 TF, D&C yes 2-RB 11201-053-12.814 hot SG_1-4 44.09 TF,D&C yes 3-RB 11201-009-27.5 cold Reactor Cavity 42.36 D&C yes 8-RB 11201-V6- PWSCC, 27.5 cold Reactor Cavity 41.09 no 001-W35-RB D&C 11201-009-27.5 cold Reactor Cavity 41.09 D&C no 9-RB 11201-012-27.5 cold Reactor Cavity 40.77 D&C no 8-RB 11201-010-27.5 cold Reactor Cavity 39.41 D&C no 6-RB 11201-V6- PWSCC, 27.5 cold Reactor Cavity 39.27 no 001-W38-RB D&C 11201-012-27.5 cold Reactor Cavity 39.27 D&C no 9-RB 11201-010-27.5 cold Reactor Cavity 38.45 D&C no 7-RB 11201-V6- PWSCC, 27.5 cold Reactor Cavity 38.43 no 001-W34-RB D&C E3-42

Enclosure 3 Risk Quantification Max Weld Break Weld Transported Degradation Failure Compartment ID Side Identifier Fiber to RHR Mechanism  ? A Strainer (ft 3 ) 11201-011-27.5 cold Reactor Cavity 37.77 D&C no 7-RB 11201-011-27.5 cold Reactor Cavity 36.70 D&C no 8-RB 11201-V6- PWSCC, 27.5 cold Reactor Cavity 36.70 no 001-W39-RB D&C 11201-053-12.814 hot SG_1-4 36.37 TF,D&C yes 1-RB 11201-004-12.814 hot SG 1-4 35.50 TF,D&C no 2-RB 11201-053-12.814 hot SG_1-4 26.42 TF,D&C no 4-RB 11204-124-8.75 cold SG_1-4 26.20 D&C no 16-RB 11204-124-8.75 cold SG 1-4 25.64 D&C no 17-RB 11204-127-8.75 cold SG_1-4 25.48 D&C no 20-RB 11201-036-10.5 hot SG_1-4 25.01 D&C no 5-RB 11201-036-10.5 hot SG_1-4 24.65 D&C no 6-RB 11204-127-8 cold SG_1-4 24.62 D&C no 21-RB 11201-049-10.5 hot SG_1-4 24.35 D&C no 1-RB 11201-004-10.5 hot SG_1-4 24.32 D&C no 3-RB 11201-036-10.5 hot SG_1-4 24.31 D&C no 4-RB 11204-124-8.75 cold SG_1-4 24.27 D&C no 18-RB 11204-:126-8.75 cold SG_2-3 23.80 D&C no 16-RB 11201-001-10.5 hot SG_1-4 23.44 D&C no 2-RB 11204-126-8.75 cold SG_2-3 23.42 D&C no 17-RB 11204-127-8.75 cold SG_1-4 23.40 D&C no 22-RB 11201-036-10.5 hot SG_1-4 22.82 D&C no 1-RB 11201-009-8.75 cold SG_1-4 22.70 D&C no 6-RB E3-43

Enclosure 3 Risk Quantification Max Weld Break Weld Transported Degradation Failure Compartment ID Side Identifier Fiber to RHR Mechanism  ? A Strainer (ft3 ) 11204-126-8.75 cold SG_2-3 22.27 D&C no 18-RB 11204-125-8.75 cold SG_2-3 21.99 D&C no 16-RB 11201-012-8.75 cold SG 1-4 21.99 D&C no 6-RB 11201-049-10.5 hot SG_1-4 21.53 D&C no 2-RB 11204-125-8.75 cold SG_2-3 21.47 D&C no 17-RB 11201-049-10.5 hot SG_1-4 20.97 D&C no 6-RB 11201-011-8.75 cold SG_2-3 20.39 D&C no 5-RB 11204-125-8.75 cold SG_2-3 20.34 D&C no 18-RB 11201-036-10.5 hot SG_1-4 20.25 D&C no 3-RB 11201-036-10.5 hot SG_1-4 20.18 D&C no 2-RB 11204-021-10.5 hot SG 1-4 20.08 D&C no 27-RB 11201-049-10.5 hot SG_1-4 20.05 D&C no 3-RB 11201-053-11.188 hot SG_1-4 20.00 TF,D&C no 5-RB 11204-021-10.5 hot SG_1-4 19.83 D&C no 28-RB 11201-049-10.5 hot SG_1-4 19.60 D&C no 4-RB 11201-049-10.5 hot SG_1-4 19.36 D&C no 5-RB 11201-010-8.75 cold SG_2-3 18.68 D&C no 4-RB 11201-V6- Pressurizer 11.188 hot 16.54 D&C no 002-W22-RB Compartment 11201-058- Pressurizer 5.189 cold 14.94 D&C no 6-RB Compartment 11201-V6- Pressurizer 5.189 cold 14.93 D&C no 002-W17-RB Compartment 11201-059- Pressurizer 5.189 cold 14.87 D&C no 7-RB Compartment 11201-V6- Pressurizer 5.189 cold 14.78 D&C no 002-W18-RB Compartment E3-44

Enclosure 3 Risk Quantification Max Weld Break Weld Transported Degradation Failure Compartment ID Side Identifier Fiber to RHR Mechanism  ? A Strainer (ft3 ) 11201-059- Pressurizer 5.189 cold 14.75 D&C no 6-RB Compartment 11201-056- Pressurizer 5.189 cold 14.73 D&C no 5-RB Compartment 11201-V6- Pressurizer 5.189 cold 14.73 D&C no 002-W19-RB Compartment 11201-059- Pressurizer 5.189 cold 14.69 D&C no 2-RB Comoartment 11201-V6- Pressurizer 5.189 cold 14.69 D&C no 002-W20-RB Compartment 11201-059- Pressurizer 5.189 cold 14.66 D&C no 5-RB Comoartment 11201-059- Pressurizer 5.189 cold 14.58 D&C no 4-RB Comoartment 11201-057- Pressurizer 5.189 cold 14.56 D&C no 2-RB Comoartment 11201-059- Pressurizer 5.189 cold 14.53 D&C no 8-RB Comoartment 11201-057- Pressurizer 5.189 cold 14.53 D&C no 7-RB Comoartment 11201-058- Pressurizer 5.189 cold 14.48 D&C no 2-RB Compartment 11201-059- Pressurizer 5.189 cold 14.48 D&C no 3-RB Comoartment 11201-056- Pressurizer 5.189 cold 14.41 D&C no 2-RB Compartment 11201-058- Pressurizer 5.189 cold 14.39 D&C no 5-RB Comoartment 11201-058- Pressurizer 5.189 cold 14.33 D&C no 7-RB Comoartment 11201-058- Pressurizer 5.189 cold 14.30 D&C no 3-RB Comoartment 11201-057- Pressurizer 5.189 cold 14.30 D&C no 6-RB Comoartment 11201-057- Pressurizer 5.189 cold 14.28 D&C no 8-RB Compartment 11201-059- Pressurizer 5.189 cold 14.27 D&C no 9-RB Comoartment 11201-057- Pressurizer 5.189 cold 14.21 D&C no 3-RB Comoartment 11201-056- Pressurizer 5.189 cold 14.14 D&C no 4-RB Comoartment 11201-058- Pressurizer 5.189 cold 14.14 D&C no 4-RB Comoartment E3-45

Enclosure 3 Risk Quantification Max Weld Break Weld Transported Degradation Failure Compartment ID Side Identifier Fiber to RHR Mechanism  ? A Strainer (ft3 ) 11201-059- Pressurizer 5.189 cold 14.00 D&C no 10-RB Compartment 11201-056- Pressurizer 5.189 cold 13.99 D&C no 6-RB Compartment 11201-057- Pressurizer 5.189 cold 13.91 D&C no 5-RB Compartment 11201-057- Pressurizer 5.189 cold 13.87 D&C no 4-RB Compartment 11201-012-3.438 cold SG 1-4 13.68 D&C no 4-RB 11201-056- Pressurizer 5.189 cold 13.67 D&C no 3-RB Compartment 11201-030-3.438 cold SG_1-4 13.57 D&C no 1-RB 11201-009-3.438 cold SG_1-4 13.56 D&C no 4-RB 11201-029-3.438 cold SG 1-4 13.52 D&C no 1-RB 11201-V6- Pressurizer 3.438 cold 13.48 D&C no 002-W21-RB Compartment 11201-059- Pressurizer 5.189 cold 12.97 D&C no 11-RB Compartment 11201-059- Pressurizer 5.189 cold 12.84 D&C no 12-RB Compartment 11201-059- Pressurizer 5.189 cold 12.83 D&C no 13-RB Compartment 11201-060- Pressurizer 5.189 cold 12.82 D&C no 1-RB Compartment 11201-060- Pressurizer 5.189 cold 12.82 D&C no 2-RB Compartment 11201-009-2.626 cold SG 1-4 12.80 VF,TF, D&C no 5-RB 11201-011-2.626 cold SG_2-3 12.78 VF,D&C no 4-RB 11208-009-2.626 cold SG_1-4 12.77 VF,TF, D&C no 6-RB 11201-048-2.626 cold SG_2-3 12.73 VF,D&C no 1-RB 11201-012-2.626 cold SG 1-4 12.67 VF,TF, D&C no 5-RB 11201-008-2.626 cold SG_1-4 12.63 D&C no 5-RB 11201-005-2.626 cold SG_1-4 12.63 D&C no 5-RB E3-46

Enclosure 3 Risk Quantification Max Weld Break Weld Transported Degradation Failure Compartment ID Side Identifier Fiber to RHR Mechanism  ? A Strainer (ft 3 ) 11201-006-2.626 cold SG_2-3 12.62 D&C no 5-RB 11201-007-2.626 cold SG_2-3 12.61 D&C no 5-RB 11204-023-5.189 hot SG_1-4 12.54 D&C no 20-RB 11204-023-5.189 hot SG_1-4 12.48 D&C no 21-RB 11201-011-2.626 cold Reactor Cavity 12.48 TF,D&C no 6-RB 11201-012-2.626 cold Reactor Cavity 12.47 TF, D&C no 7-RB 11201-029-3.438 cold SG_1-4 12.45 D&C no 2-RB 11204-246-2.626 cold Reactor Cavity 12.44 TF,D&C no 36-RB 11201-005-2.626 cold SG_1-4 12.44 D&C no 9-RB 11201-006-2.626 cold SG_2-3 12.43 D&C no 9-RB 11201-008-2.626 cold SG_1-4 12.42 D&C no 9-RB 11201-007-2.626 cold SG_2-3 12.41 D&C no 9-RB 11204-245-2.626 cold Reactor Cavity 12.37 TF, D&C no 33-RB 11201-030- Pressurizer 5.189 cold 12.37 D&C no 29-RB Compartment 11201-030- Pressurizer 5.189 cold 12.35 TF,D&C no 30-RB Compartment 11201-030- Pressurizer 5.189 cold 12.32 TF,D&C no 34-RB Compartment 11201-030- Pressurizer 5.189 cold 12.32 TF, D&C no 33-RB Compartment 11201-029-3.438 cold SG_1-4 12.30 D&C no 3-RB 11208-007-2.626 cold SG_1-4 12.29 VF,TF, D&C no 6-RB 11201-029-3 3.438 cold SG 1-4 12.26 D&C no A-RB 11201-030- Pressurizer 5.189 cold 12.24 D&C no 28-RB Compartment 11201-029-3.438 cold SG_1-4 12.23 D&C no 8-RB E3-47

Enclosure 3 Risk Quantification Max Weld Break Weld Transported Degradation Failure Compartment ID Side Identifier Fiber to RHR Mechanism  ? A Strainer (ft3 ) 11201-029-3.438 cold SG_1-4 12.21 D&C no 5-RB 11201-029-3.438 cold SG_1-4 12.17 D&C no 6-RB 11201-029-3.438 cold SG_1-4 12.16 D&C no 9-RB 11201-029-3.438 cold SG_1-4 12.15 D&C no 7-RB 11201-048-2 cold SG_2-3 12.14 VF, D&C no 2-RB 11201-030- Pressurizer 5.189 cold 12.13 TF,D&C no 35-RB Compartment 11201-030- Pressurizer 5.189 cold 12.13 TF,D&C no 32-RB Compartment 11201-030- Pressurizer 5.189 cold 12.11 TF,D&C no 31 A-RB Compartment 11201-011-2.125 cold SG_2-3 12.11 D&C no 3-RB 11201-009-2.125 cold SG_1-4 12.11 D&C no 3-RB 11201-012-2.125 cold SG_1-4 12.11 D&C no 3-RB 11201-010-2.125 cold SG_2-3 12.11 D&C no 3-RB 11201-029-3.438 cold SG_1-4 12.08 D&C no 4-RB 11201-030- Pressurizer 3.438 cold 12.04 D&C no 20-RB Compartment 11201-030- Pressurizer 3.438 cold 12.03 D&C no 19-RB Compartment 11201-048-2.626 cold SG_2-3 12.01 VF, D&C no 3-RB 11201-009-2.626 cold SG_1-4 12.00 TF, D&C no 7-RB 11201-010-2.626 cold SG_2-3 11.97 TF,D&C no 5-RB 11204-243-2.626 cold SG 1-4 11.97 TF,D&C no 34-RB 11201-030- Pressurizer 3.438 cold 11.93 D&C no 21-RB Compartment 11204-244-2.626 cold SG_2-3 11.92 TF, D&C no 28-RB 11204-023-5.189 hot SG_1-4 11.91 D&C no 19-RB E3-48

Enclosure 3 Risk Quantification Max Weld Break Weld Transported Degradation Failure Compartment ID Side Identifier Fiber to RHR Mechanism  ? A Strainer (ft 3 ) 11208-007-2.626 cold SG 1-4 11.89 VF,TF, D&C no 5-RB 11208-007-2.626 cold SG 1-4 11.86 VF,TF, D&C no 4-RB 11201-030- Pressurizer 3.438 cold 11.85 D&C no 18-RB Compartment 11201-030- Pressurizer 3.438 cold 11.83 D&C no 17-RB Compartment 11201-030-3.438 cold SG_1-4 11.83 D&C no 12-RB 11201-029- Pressurizer 3.438 cold 11.82 D&C no 25-RB Compartment 11201-030-2 cold SG_1-4 11.81 D&C no 14-RB 11201-030-3.438 cold SG_1-4 11.81 D&C no 13-RB 11201-030-3.438 cold SG_1-4 11.81 D&C no 11-RB 11208-009-2.626 cold SG 1-4 11.79 VF,TF, D&C no 5-RB 11201-030- Pressurizer 3.438 cold 11.78 D&C no 22-RB Compartment 11201-030-3.438 cold SG_1-4 11.78 D&C no 5-RB 11208-009-2.626 cold SG_1-4 11.77 VF,TF, D&C no 4-RB 11201-009-1.689 cold SG_1-4 11.77 D&C no 2-RB 11201-030-3.438 cold SG_1-4 11.77 D&C no 8-RB 11201-011-1.689 cold SG_2-3 11.77 D&C no 2-RB 11201-012-1.689 cold SG_1-4 11.77 D&C no 2-RB 11201-010-1.689 cold SG_2-3 11.77 D&C no 2-RB 11201-029- Pressurizer 3.438 cold 11.76 D&C no 24-RB Compartment 11201-030-3.438 cold SG_1-4 11.76 D&C no 9-RB 11201-030-3.438 cold SG_1-4 11.76 D&C no 4-RB 11201-007-1.689 cold SG_2-3 11.76 D&C no 6-RB E3-49

Enclosure 3 Risk Quantification Max Weld Break Weld Transported Degradation Failure Compartment ID Side Identifier Fiber to RHR Mechanism  ? A Strainer (ft3 ) 11201-005-1.689 cold SG_1-4 11.76 D&C no 6-RB 11208-012-1.689 cold Annulus 11.76 VF,D&C no 3-RB 11201-006-1.689 cold SG_2-3 11.76 D&C no 6-RB 11201-008-1.689 cold SG_1-4 11.76 D&C no 6-RB 11201-030-2 cold SG_1-4 11.75 D&C no 7-RB 11208-012-5 1.689 cold Annulus 11.75 VF,D&C no 8-RB 11201-031-1.689 cold SG_1-4 11.75 D&C no 1-RB 11201-029- Pressurizer 3.438 cold 11.74 D&C no 22-RB Compartment 11201-029-3.438 cold SG_1-4 11.74 D&C no 18-RB 11201-030-3.438 cold SG_1-4 11.74 D&C no 6-RB 11201-011-1.689 cold SG_2-3 11.74 D&C no 9-RB 11201-010-1.689 cold SG_2-3 11.74 D&C no 8-RB 11201-009-1.689 cold SG_1-4 11.74 D&C no 10-RB 11201-012-1.689 cold SG_1-4 11.74 D&C no 10-RB 11201-029-3.438 cold SG_1-4 11.73 D&C no 19-RB 11201-030-3.438 cold SG_1-4 11.73 D&C no 3-RB 11201-042-1.689 cold SG_2-3 11.73 D&C no 1-RB 11201-051-1.689 cold SG_1-4 11.72 D&C no 1-RB 11201-046- - 1.689 cold SG_2-3 11.72 D&C no 1-RB 11208-012-5 1.689 cold Annulus 11.72 VF,D&C no A-RB 11201-042-1.689 cold SG_2-3 11.72 TF,D&C no 2-RB 11201-051-1.689 cold SG_1-4 11.72 TF,D&C no 2-RB E3-50

Enclosure 3 Risk Quantification Max Weld Break Weld Transported Degradation Failure Compartment ID Side Identifier Fiber to RHR Mechanism  ? A Strainer (ft3 ) 11201-046-1.689 cold SG_2-3 11.72 TF,D&C no 2-RB 11208-012-1.689 cold Annulus 11.72 VF, D&C no 4-RB 11201-031-1.689 cold SG_1-4 11.72 TF,D&C no 2-RB 11208-012-1.689 cold Annulus 11.72 VF,D&C no 5-RB 11201-030-2 cold SG_1-4 11.71 D&C no 2-RB 11208-007-2.626 cold SG_1-4 11.71 VF,TF, D&C no 3-RB 11201-029- Pressurizer 3.438 cold 11.70 D&C no 26-RB Compartment 11201-029-3.438 cold SG_1-4 11.69 D&C no 17-RB 11208-009-2.626 cold SG_1-4 11.68 VF,TF, D&C no 3-RB 11201-046-1.689 cold SG_2-3 11.68 TF, D&C no 3-RB 11201-030- Pressurizer 5.189 cold 11.68 TF, D&C no 38-RB Compartment 11201-051-1.689 cold SG_1-4 11.68 TF, D&C no 3-RB 11201-042-1.689 cold SG_2-3 11.68 TF, D&C no 3-RB 11201-029-3.438 cold SG_1-4 11.68 D&C no 12-RB 11201-031-1.689 cold SG_1-4 ., ...-;-"'~ 11.68 TF, D&C no 3-RB 11201-029-3.438 cold SG_1-4 11.68 D&C no 13-RB 11201-060- Pressurizer 2.626 cold 11.67 D&C no 3-RB Compartment 11201-029-3.438 cold SG_1-4 11.67 D&C no 14-RB 11201-029- Pressurizer 3.438 cold 11.66 D&C no 23-RB Compartment 11201-029-3.438 cold SG_1-4 11.66 D&C no 11-RB 11201-029-3.438 cold SG_1-4 11.65 D&C no 16-RB 11201-048-2.626 cold SG_2-3 11.64 VF,TF, D&C no 5-RB E3-51

Enclosure 3 Risk Quantification Max Weld Break Weld Transported Degradation Failure Compartment ID Side Identifier Fiber to RHR Mechanism  ? A Strainer (ft3 ) 11201-048-2.626 cold SG_2-3 11.64 VF,TF, D&C no 4-RB 11201-060- Pressurizer 2.626 cold 11.64 D&C no 5-RB Compartment 11201-029-3.438 cold SG_1-4 11.64 D&C no 15-RB 11201-030- Pressurizer 3.438 cold 11.64 D&C no 24-RB Compartment 11201-030- Pressurizer 3.438 cold 11.64 D&C no 23-RB Compartment 11201-030- Pressurizer 5.189 cold 11.64 TF, D&C no 37-RB Compartment 11201-059- Pressurizer 2.626 cold 11.64 D&C no 14-RB Compartment 11201-060- Pressurizer 2.626 cold 11.63 D&C no 4-RB Compartment 11201-030- Pressurizer 5.189 cold 11.63 TF, D&C no 36-RB Compartment 11201-003-5.189 hot SG_2-3 11.61 D&C no 2-RB 11201-048-2.626 cold SG_2-3 11.61 VF,TF, D&C no 6-RB 11201-059- Pressurizer 2.626 cold 11.60 D&C no 15-RB Compartment 11201-060- Pressurizer 2.626 cold 11.60 D&C no 6-RB Compartment 11201-030- Pressurizer 3.438 cold 11.60 D&C no 25-RB Compartment 11201-030- Pressurizer 3.438 cold 11.59 D&C no 26-RB Compartment 11201-059- Pressurizer 2.626 cold 11.59 D&C no 16-RB Compartment 11201-030- Pressurizer 3.438 cold 11.59 D&C no 27-RB Compartment 11201-051-1.689 cold SG_1-4 11.56 D&C no 4-RB 11201-048-2.626 cold SG_2-3 11.54 VF,TF, D&C no 7-RB 11201-059- Pressurizer 2.626 cold 11.54 D&C no 17-RB Compartment 11204-246-1.338 cold Reactor Cavity 11.54 TF, D&C no 35-RB 11201-030- Pressurizer 3.438 cold 11.53 TF,D&C no 39-RB Compartment E3-52

Enclosure 3 Risk Quantification Max Weld Break Weld Transported Degradation Failure Compartment ID Side Identifier Fiber to RHR Mechanism  ? A Strainer (ft3 ) 11201-030- Pressurizer 3.438 cold 11.52 D&C no 16-RB Compartment 11201-030- Pressurizer 3.438 cold 11.51 D&C no 15-RB Compartment 11201-048-2.626 cold SG 2-3 11.51 VF,TF, D&C no 10-RB 11201-051-1.689 cold SG_1-4 11.50 D&C no 6-RB 11201-048-2.626 cold SG_2-3 11.50 VF,TF, D&C no 8-RB 11201-048-2.626 cold SG_2-3 11.50 VF,TF, D&C no 9-RB 11201-051-1.689 cold SG_1-4 11.49 D&C no 5-RB 11201-029- Pressurizer 3.438 cold 11.49 D&C no 21-RB Compartment 11204-245-1.338 cold Reactor Cavity 11.49 TF,D&C no 32-RB 11201-029- Pressurizer 3.438 cold 11.49 D&C no 20-RB Compartment 11201-030- Pressurizer 3.438 cold 11.47 TF,D&C no 40-RB Compartment 11201-042-1.689 cold SG 2-3 11.46 D&C no 4-RB 11201-030- Pressurizer 3.438 cold 11.45 TF,D&C no 41-RB Compartment 11201-030- Pressurizer 3.438 cold 11.45 TF,D&C no 44-RB Compartment 11201-030- Pressurizer 3.438 cold 11.45 TF,D&C no 42-RB Compartment 11201-030- Pressurizer 3.438 cold 11.45 TF,D&C no 43-RB Compartment 11201-060- Pressurizer 2.626 cold 11.45 D&C no 7-RB Compartment 11201-030- Pressurizer 1.689 cold 11.44 VF,TF, D&C no 31-RB Compartment 11208-012- Pressurizer 1.689 cold 11.43 VF,D&C no 6-RB Compartment 11201-059- Pressurizer 2.626 cold 11.43 D&C no 18-RB Compartment 11201-042-1.689 cold SG_2-3 11.43 D&C no 5-RB 11201-031-1.689 cold SG_1-4 11.43 D&C no 5-RB E3-53

Enclosure 3 Risk Quantification Max Weld Break Weld Transported Degradation Failure Compartment ID Side Identifier Fiber to RHR Mechanism  ? A Strainer (ft 3 ) 11201-031-1.689 cold SG_1-4 11.42 D&C no 6-RB 11204-246-1.338 cold Reactor Cavity 11.42 TF,D&C no 34-RB 11204-243-1.338 cold SG_1-4 11.42 TF,D&C no 33-RB 11204-245-1.338 cold Reactor Cavity 11.42 TF,D&C no 31-RB 11201-042-1.689 cold SG_2-3 11.42 D&C no 6-RB 11201-046-1.689 cold SG_2-3 11.41 D&C no 5-RB 11201-046-1.689 cold SG_2-3 11.41 D&C no 4-RB 11201-031-1.689 cold SG_1-4 11.41 D&C no 4-RB 11201-030-1.16 cold SG_1-4 11.40 D&C no 10-RB 11201-059- Pressurizer 2.626 cold 11.40 D&C no 19-RB Compartment 11201-029-1.16 cold SG_1-4 11.40 D&C no 10-RB 11204-244-1.338 cold SG 2-3 11.40 TF, D&C no 27-RB 11204-246-1.338 cold Reactor Cavity 11.40 TF, D&C no 33-RB 11201-030-1.16 cold SG_1-4 11.40 D&C no 46-RB 11201-029-1.16 cold SG_1-4 11.40 D&C no 27-RB 11201-060- Pressurizer 2.626 cold 11.40 D&C no 8-RB Compartment 11204-245-1.338 cold Reactor Cavity 11.39 TF,D&C no 30-RB 11201-060- Pressurizer 2.626 cold 11.38 D&C no 10-RB Compartment 11204-243-1.338 cold SG_1-4 11.37 TF, D&C no 32-RB 11201-059- Pressurizer 2.626 cold 11.37 D&C no 20-RB Compartment 11201-060- Pressurizer 2.626 cold 11.37 D&C no 9-RB Compartment 11204-243-1.338 cold SG_1-4 11.36 D&C no 30-RB E3-54

Enclosure 3 Risk Quantification Max Weld Break Weld Transported Degradation Failure Compartment ID Side Identifier Fiber to RHR Mechanism  ? A Strainer (ft3 ) 11204-243-1.338 cold 31-RB SG_1-4 11.36 TF,D&C no 11204-245-1.338 cold 29-RB SG_2-3 11.36 D&C no 11208-045-1.338 cold 14 A-RB SG_2-3 11.35 VF, D&C no 11208-045-1.338 cold 14 B-RB SG 2-3 11.35 VF,D&C no 11208-045-1.338 cold 14-RB SG_2-3 11.35 VF,D&C no 11208-043-5 1.338 cold A-RB SG_2-3 11.35 VF,D&C no 11208-043-5 1.338 cold B-RB SG 2-3 11.35 VF,D&C no 11208-024-5 1.338 cold A-RB SG_1-4 11.35 VF,D&C no 11208-024-5 1.338 cold B-RB SG 1-4 11.35 VF,D&C no 11208-047-1.338 cold 14 A-RB SG 1-4 11.35 VF,D&C no 11208-047-1.338 cold 14 B-RB SG_1-4 11.35 VF,D&C no 11204-244-1.338 cold 26-RB SG_2-3 11.34 TF,D&C no 11208-047-1.338 cold 14-RB SG_1-4 11.34 VF,D&C no 11208-024-1.338 cold 5-RB SG 1-4 11.34 VF,D&C no 11208-043-1.338 cold 5-RB SG_2-3 11.34 VF,D&C no 2.626 cold 11201-059- Pressurizer 11.34 D&C no 21-RB Compartment 11208-045-1.338 cold 13 A-RB SG_2-3 11.33 VF,D&C no 11204-244-1.338 cold 25-RB SG 2-3 11.33 TF,D&C no 11208-045-1.338 cold 9-RB SG_2-3 11.33 VF,D&C no 11208-045-1.338 cold 8-RB SG_2-3 11.33 VF,D&C no 11208-045-9 1.338 cold A-RB SG_2-3 11.33 VF, D&C no 11208-045-1.338 cold 16-RB SG_2-3 11.33 VF, D&C no E3-55

Enclosure 3 Risk Quantification Max Weld Break Weld Transported Degradation Failure Compartment ID Side Identifier Fiber to RHR Mechanism  ? A Strainer (ft 3 ) 11208-045-1.338 cold SG_2-3 11.33 VF,D&C no 12-RB 11208-045-1.338 cold SG_2-3 11.33 VF,D&C no 15-RB 11208-024-1.338 cold SG_1-4 11.33 VF, D&C no 3-RB 11208-045-9 1.338 cold SG 2-3 11.32 VF, D&C no B-RB 11208-045-1.338 cold SG_2-3 11.32 VF,D&C no 13-RB 11208-045-1.338 cold SG_2-3 11.32 VF,D&C no 10-RB 11208-045-1.338 cold SG_2-3 11.32 VF,D&C no 11-RB 11208-024-4 1.338 cold SG_1-4 11.32 VF, D&C no B-RB 11204-246-1.338 cold SG_1-4 11.32 TF, D&C no 32-RB 11208-047-9 1.338 cold SG 1-4 11.32 VF, D&C no C-RB 11208-047-9 1.338 cold SG 1-4 11.32 VF, D&C no B-RB 11208-047-1.338 cold SG 1-4 11.32 VF, D&C no 16-RB 11208-024-4 1.338 cold SG_1-4 11.31 VF, D&C no A-RB 11208-024-1.338 cold SG 1-4 11.31 VF,D&C no 7-RB 11208-043-1.338 cold SG_2-3 11.31 VF,D&C no 6-RB 11208-047-9 1.338 cold SG_1-4 11.31 VF,D&C no A-RB 11208-043-3 1.338 cold SG_2-3 11.31 VF,D&C no A-RB 11208-043-3 1.338 cold SG_2-3 11.31 VF,D&C no B-RB 11208-043-1.338 cold SG_2-3 11.31 VF, D&C no 3-RB 11208-043-3 1.338 cold SG_2-3 11.31 VF, D&C no C-RB 11208-043-3 1.338 cold SG_2-3 11.31 VF,D&C no D-RB 11208-047-5 1.338 cold SG_1-4 11.30 VF, D&C no B-RB E3-56

Enclosure 3 Risk Quantification Max Weld Break Weld Transported Degradation Failure Compartment ID Side Identifier Fiber to RHR Mechanism  ? A Strainer (ft3 ) 11208-047-1.338 cold SG_1-4 11.30 VF, D&C no 15-RB 11208-043-4 1.338 cold SG_2-3 11.30 VF, D&C no A-RB 11208-043-1.338 cold SG_2-3 11.30 VF, D&C no 7-RB 11208-024-3 0.5 cold SG_1-4 11.30 VF, D&C no A-RB 11208-024-3 0.5 cold SG_1-4 11.30 VF, D&C no B-RB 11208-024-0.5 cold SG 1-4 11.30 VF, D&C no 4-RB 11208-024-0.5 cold SG 1-4 11.30 VF, D&C no 6-RB 11208-045-0.5 cold SG 2-3 11.30 VF,D&C no 3-RB 11208-045-0.5 cold SG 2-3 11.30 VF, D&C no 4-RB 11208-045-5 0.5 cold SG_2-3 11.30 VF,D&C no A-RB 11208-045-5 0.5 cold SG_2-3 11.30 VF, D&C no B-RB 11208-045-0.5 cold SG_2-3 11.30 VF, D&C no 5-RB 11208-045-0.5 cold SG_2-3 11.30 VF, D&C no 6-RB 11208-045-0.5 cold SG_2-3 11.30 VF, D&C no 7-RB 11208-047-0.5 cold SG_1-4 11.30 VF, D&C no 3-RB 11208-047-5 1.338 cold SG_1-4 11.30 VF, D&C no A-RB 11208-043-4 1.338 cold SG_2-3 11.30 VF, D&C no B-RB 11204-244-0.5 cold SG 2-3 11.30 TF,D&C no 24-RB 11208-024-4 0.5 cold SG_1-4 11.30 VF, D&C no AA-RB 11208-043-0.5 cold SG_2-3 11.30 VF, D&C no 4-RB 11204-025-5.189 hot SG_2-3 11.28 D&C no 24-RB 11201-002-5.189 hot SG_2-3 11.26 D&C no 2-RB E3-57

Enclosure 3 Risk Quantification Max Weld Break Weld Transported Degradation Failure Compartment ID Side Identifier Fiber to RHR Mechanism  ? A Strainer (ft 3 ) 11204-024-5.189 hot SG_2-3 11.23 D&C no 19-RB 11204-024-5.189 hot SG_2-3 11.20 D&C no 14-RB 11204-021-5.189 hot SG_1-4 11.12 D&C no 18-RB 11204-025-5.189 hot SG_2-3 11.09 D&C no 23-RB 11204-025-5.189 hot SG_2-3 11.06 D&C no 20-RB 11204-021-5.189 hot SG_1-4 10.97 D&C no 19-RB 11204-021-5.189 hot SG_1-4 10.89 D&C no 17-RB 11204-024-5.189 hot SG_2-3 10.84 D&C no 18-RB 11204-025-5.189 hot SG_2-3 10.83 D&C no 22-RB 11204-025-5.189 hot SG_2-3 10.82 D&C no 21-RB 11204-024-5.189 hot SG_2-3 10.67 D&C no 17-RB 11204-021-5.189 hot SG_1-4 10.41 D&C no 26-RB 11204-021-5.189 hot SG_1-4 10.41 D&C no 22-RB 11204-024-5.189 hot SG_2-3 10.41 D&C no 16-RB 11204-024-5.189 hot SG_2-3 10.33 D&C no 15-RB 11204-023-5.189 hot SG_1-4 10.27 D&C no 17-RB 11204-021-5.189 hot SG 1-4 10.18 D&C no 25-RB 11204-021-5.189 hot SG_1-4 10.15 D&C no 20-RB 11204-021-5.189 hot SG_1-4 10.15 D&C no 23-RB 11204-023-5.189 hot SG 1-4 10.03 D&C no 18-RB 11204-021-5.189 hot SG 1-4 10.03 D&C no 24-RB 11204-023-5.189 hot SG_1-4 9.93 D&C no 16-RB E3-58

Enclosure 3 Risk Quantification Max Weld Break Weld Transported Degradation Failure Compartment ID Side Identifier Fiber to RHR Mechanism  ? A Strainer (ft3 ) 11204-021-5.189 hot SG_1-4 9.84 D&C no 21-RB 11201-003-2.125 hot SG_2-3 9.15 D&C no 4-RB 11201-004-2.125 hot SG_1-4 9.15 D&C no 5-:RB 11201-002-2.125 hot SG_2-3 9.14 D&C no 4-RB 11201-001-2.125 hot SG_1-4 9.14 D&C no 4-RB 11204-025-4 hot SG_2-3 8.48 D&C no 25-RB 14.2 NARWHAL Uncertainty and Sensitivity The purpose of this section is to describe the sensitivity analysis and uncertainty quantification associated with the GSl-191 phenomenological models evaluated using NARWHAL. 14.2.1 Simplified Risk Estimation Methodology For the purposes of sensitivity analysis and uncertainty quantification, a simplified method was used to estimate LiCDF in NARWHAL. The simplification includes a reduction in the number of equipment configurations explicitly evaluated for each sensitivity, and also directly calculates LiCDF in NARWHAL using the LOCA frequencies and the equipment functional failure probabilities (along with the NARWHAL calculated CFP results). The GSl-191 risk can be reasonably estimated for the purpose of sensitivity analysis without explicitly modeling all six of the configurations listed in Section 6.3. The scenario with a combined failure of one RHR pump and two CS pumps has a very low probability resulting in a negligible impact on the risk quantification. Although it is not necessarily bounded by any of the other cases, it was assumed to have the same conditional failure probability as the scenario with a combined failure of one RHR pump and one CS pump. The scenario with one RHR pump failure is bounded by the scenario with a combined failure of one RHR pump and one CS pump. In addition, the scenario with one CS pump failure is bounded by the scenario with two CS pump failures. Therefore, the functional failure probabilities listed in Section 6.3 were combined for the purpose of sensitivity analysis and uncertainty quantification as shown in Table 3-10. E3-59

Enclosure 3 Risk Quantification Table 3 Combined Functional Failure Probabilities Functional Failure Equipment Configuration Probability No Equipment Failures 91.50% 2 CS Pump Failures 6.57% 1 RHR Pump+ 1 CS Pump Failures 1.92% Total 100% For the VEGP NARWHAL model, the CFPs were reported for each PRA size category and success criterion. To estimate the LlCDF, the CFPs for each PRA success criterion were summed within each PRA size category. Therefore, a single CFP value was calculated for each PRA size category. The following equation was then used to estimate LlCDF. i=N j=X

 .1CDF =   II i=O j=O IEFi

• CFPij
• FF11 Nomenclature:

                = Each PRA size category j       = Each equipment configuration IEF     = Initiating event frequency for each PRA size category CFP     =Conditional failure probability for each PRA size category and each equipment configuration FFP     =Functional failure probability for each equipment configuration Note that initiating event frequency of each PRA size category was defined by the LOCA frequencies from Section 6.4.

If there are no medium or small breaks that fail, the equation can be simplified as shown below: j=X

 .1CDF = IEF    *I

• j=O CF11 FF11 Nomenclature:

j = Each equipment configuration IEF = Initiating event frequency for large LOCAs CFP =Large LOCA conditional failure probability for each equipment configuration FFP = Functional failure probability for each equipment configuration E3-60

Enclosure 3 Risk Quantification Given a LOCA frequency for large breaks of 1.85x1 o-6 yr 1 (see Section 6.4), the functional failure probabilities shown in Table 3-10, and the conditional failure probabilities shown in Table 3-5, LiCDF can be estimated as shown below: LlCDF = 1.85 x 10- 6 * (0.9150

• 0.0118 + 0.0657
• 0.0177 + 0.0192
• 0.0736) = 2.47 x 10-s This value is nearly identical to the LiCDF value that was calculated using the GSl-191 PRA model (2.46x1 o-8 yr 1 excluding the SSBI contribution in Table 3-8). Therefore, this method provides an efficient and accurate LiCDF estimate.

Because the base CDF and LERF values at VEGP are well within the RG 1.174 acceptance guidelines (Reference 1), and LiLERF is more than two orders of magnitude lower than LiCDF (see Table 3-8), the risk sensitivity was evaluated by comparing LiCDF to the LiCDF acceptance guideline. As shown in Enclosure 1, Section 2.2, the risk associated with GSl-191 is considered to be small for a mean LiCDF below 1x10-5 yr 1 and very small for a mean LiCDF below 1 x1 o-6 yr 1 . 14.2.2 Sensitivity Analysis Parametric sensitivity analysis was performed to identify which inputs have the greatest impact on the risk quantification results. The parametric sensitivity analysis includes the process of identifying input variables to evaluate, selecting minimum, nominal, and maximum values for each variable, quantifying risk in terms of LiCDF as a common output that can be compared for each sensitivity, and using the LiCDF results to rank the sensitivity of each input variable. The VEGP NARWHAL model includes numerous inputs that could have been included in the sensitivity analysis. However, some of these input parameters are directly correlated to other parameters (and therefore should not be independently analyzed), some parameters were pre-screened as having an insignificant effect on the results, and some parameters do not require an independent analysis because they would have the same type of effect as other similar parameters that are evaluated. A consistent methodology was used to determine the minimum, nominal, and maximum values for each of the parametric sensitivity inputs. Consistency is important because using a very large range for one parameter and a very small range for another parameter may mask the true sensitivity of the second parameter and indicate that the first parameter has a much greater effect on the results. However, selecting consistent minimum and maximum values is challenging due to practical considerations. For example, debris transport fractions can vary between 0 percent and 100 percent, the initial RWST level may vary between the technical specification minimum limit and the high level alarm, and debris head loss may vary from 0 ft at the low end to an unknown value at the high end. In addition, some parameters are not fixed values and may be determined as a function of time (e.g., pool temperature) or as a correlation based on other calculated parameters (e.g., penetration fraction). The following methodology E3-61

Enclosure 3 Risk Quantification provides an approach for evaluating the various input parameters in a consistent manner.

• The nominal value was defined as the input value used in the NARWHAL base case. As discussed in Section 0, the base case NARWHAL model that was used for sensitivity analysis and uncertainty quantification was equivalent to the model from the NARWHAL CFP calculation with the exception of a smaller break database.
• The minimum and maximum values for each sensitivity input depend on the nominal value and the available information. If the nominal value was conservatively skewed toward the minimum direction, the minimum value used for the parametric sensitivity was 10 percent lower than the nominal value. Similarly, if the nominal value was conservatively skewed toward the maximum direction, the maximum value used for the parametric sensitivity was 10 percent higher than the nominal value.
• For all other cases, the minimum and maximum values were determined by the available information. Design limits were used preferentially if they were available. If a range of values was determined analytically, the minimum or maximum from the range was used if design limits were not available.
• If no information was available for the range of a given input, then the minimum or maximum value was assumed to be +/- 25 percent of the nominal value.

The results of the parametric sensitivity analysis were used to rank each input parameter. This was done using a tornado diagram, which illustrates how sensitive the chosen output metric (LiCDF) is to changing an input variable's value from nominal to maximum (or minimum). The tornado diagram was created by first running NARWHAL with all inputs set at nominal conditions, and recording the output metric. One variable was then changed to its maximum value (with all others held constant), the software was re-run, the output metric was recorded, and the results were compared to the nominal case. This process was repeated with each variable being independently modified to the maximum and minimum values. The output responses were then sorted by magnitude and shown from highest output response (most risk-sensitive parameter) to lowest output response (least risk-sensitive parameter). The minimum and maximum values used in the sensitivity analysis are shown in Table 3-11. The LiCDF results are shown in Table 3-12, and the difference in LiCDF (compared to the NARWHAL base case value of 2.46E-08 yr 1 ) was plotted in the tornado diagram shown in Figure 3-9. Note that based on the methodology described above, some of the parametric sensitivity cases consider input values that are outside the plant operating conditions (e.g., the RWST volume used for the NARWHAL base case corresponds to the technical specification minimum level, so the minimum volume used for the sensitivity is less than the technical specification minimum). Therefore, these sensitivity results are only intended to provide insights into the relative importance of each input parameter. E3-62 __J

Enclosure 3 Risk Quantification Table 3 Maximum and Minimum Parametric Sensitivity Inputs Input Parameter Units Minimum Maximum Input Input Simulation Time minutes 32,400 54,000 Initial RWST Level lbm 5,062,577 6,025,079 RHR Pump Flow Rate gpm 2,775 4,500 CS Pump Flow Rate gpm 1,950 3,374 75% *Design 110%

• Pressure and Temperature Profiles Basis Design Basis Sump pH 7.1 8.1 75%
• 110%
• ZOI Debris Quantity Base Case Base Case Latent Debris Quantity Particulate lbm 51 187 Fiber ft 3 3.75 13.75 2

Miscellaneous Debris Quantity ft 2 55 Submerged Aluminum Surface Area ft2 278.7 383.2 Unsubmerged Aluminum Surface Area ft2 741.3 1,019.3 Debris Head Loss Conventional for Fiber~ 3.1 ft3 ft of HzO 0.47 0.78 Conventional for Fiber> 3.1 ft3 ft of HzO 3.50 6.83 Calcium Phosphate ft of HzO 0.83 2.25 SAS ft of HzO 3.24 6.55 Strainer Debris Limits Fiber ft3 9.927 13.79 Particulate lbm 327.348 454.65 Fire Barrier lbm 26.24 36.45 Calcium Phosphate lbm 4.77 6.63 SAS lbm 8.046 11.18 Containment Accident Pressure psi 3.15 4.375 75%

• 125%
• Strainer Penetration Fractions Correlation Correlation Results Results Containment Spray Duration minutes 120 43,200 Reactor Vessel Hot Leg Break Fine Fiber Limit g/FA 50 125 Reactor Vessel Cold Leg Break Fine Fiber Limit g/FA 11.25 18.75 Geometric LOCA Frequency Values 5th Percentile 95th Percentile E3-63

Enclosure 3 Risk Quantification Table 3 Results of Parametric Sensitivity Analysis

                                                                   .dCDF at Input Parameter                     .dCDF at       Maximum Minimum Input        Input Simulation Time                                     2.46E-08        2.47E-08 Initial RWST Level                                  1.18E-08        2.47E-08 RHR Pump Flow Rate                                  2.21 E-08       5.75E-08 CS Pump Flow Rate                                   2.73E-08        2.25E-08 Pressure and Temperature Profiles                   1.09E-08        2.46E-08 Sump pH                                             1.08E-08        2.56E-08 ZOI Debris Quantity                                 8.54E-09        3.33E-08 Latent Debris Quantity                              2.30E-08        2.48E-08 Miscellaneous Debris Quantity                       2.22E-08        2.48E-08 Submerged Aluminum Surface Area                     2.47E-08        2.47E-08 Unsubmerged Aluminum Surface Area                   2.47E-08        2.47E-08 Debris Head Loss                                    2.47E-08        2.60E-08 Strainer Debris Limits                              3.47E-07        5.32E-08 Containment Accident Pressure                       2.47E-08        2.47E-08 Strainer Penetration Fractions                      2.50E-08        2.98E-08 Containment Spray Duration                          2.85E-08        2.47E-08 Reactor Vessel Hot Leg Break Fine Fiber Limit       1.09E-07        2.47E-08 Reactor Vessel Cold Leg Break Fine Fiber Limit      2.47E-08        2.47E-08 Geometric LOCA Frequency Values                     4.16E-11        6.25E-08 E3-64

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• M*UT14
• MuWN Figure 3 Tornado Diagram Showing Risk Sensitivity Ranking The parameters that have the most significant effect on LiCDF are the in-vessel fiber limit for hot leg breaks, the LOCA frequency values , and pump flow rates . The inputs that have no effect on the LiCDF (within the range analyzed) are the containment accident pressure used for degasification and flashing , the in-vessel fiber limit for cold leg breaks , and the submerged and unsubmerged aluminum surface areas.

Reducing the total fine fiber limit in the reactor vessel for hot leg breaks affects the number of core failures . Note that for the NARWHAL base case, no core fiber limit failures were observed. However, when reducing the total fiber limit to 50 g/FA, some failures were observed . Adjusting the LOCA frequencies to the 5th or 95th percentiles affects the results in two ways-first, it is an input for calculating the CFP values as described in Section 13.0, and second , it is a direct input for calculating LiCDF as described in Section 0. The effect on the CFP values is relatively minor and is driven by the change in frequency as E3-65

Enclosure 3 Risk Quantification a function of break size. The direct effect on b.CDF is driven by the magnitude of the LOCA frequency, which has a significant effect because the 5th and 95th percentiles are in some case orders of magnitude different from the mean LOCA frequencies. The flow rates of the RHR and CS pumps also affect the b.CDF. Generally, a higher RHR flow rate and a lower CS flow rate results in more fiber debris deposited on the RHR strainers. A lower RHR flow rate and higher CS flow rate results in more fiber being deposited on the CS strainers. As previously mentioned, the debris limit failure criterion is the most significant contributor to failure. Therefore, altering inputs that affect debris deposition on the strainer results in a change in the risk results. The strainer penetration fraction resulted in a higher b.CDF for the minimum input and the maximum input when compared to the base case. The maximum penetration fractions increased b.CDF because it resulted in an increased quantity of fine fiber deposition in the reactor vessel. This led to core failures, which were not seen in the NARWHAL base case. The minimum penetration fractions increased b.CDF because it resulted in an increased quantity of fine fiber deposited on the strainers. At each time step, less of the fine fiber that arrives penetrated the strainer, and less fine fiber that was previously on the strainer was shed. Although this does not have a large effect, it led to a few more fiber debris limit failures than were seen in the NARWHAL base case. Additionally, the containment temperature and pressure profiles parameter resulted in a lower b.CDF for both the minimum and maximum values. The minimum temperature and pressure profiles result in a lower b.CDF because the quantity of chemical precipitates that form is reduced, which leads to less debris limit failures. The maximum temperature and pressure profiles result in a negligibly lower b.CDF than when compared to the base case. This is because the higher temperature results in a larger volume of water in the containment pool, which affects the rate at which debris accumulates on the strainer. For a handful of breaks, this variation in the rate of debris accumulation resulted in success instead of failure due to the debris limit failure criterion. 14.2.3 Uncertainty Quantification As described in Enclosure 1, Section 5.0, uncertainty quantification includes parametric uncertainty, model uncertainty, and completeness uncertainty. The parametric and model uncertainties were quantified by running NARWHAL sensitivity cases. Note that the parametric uncertainty evaluation has a different purpose than the parametric sensitivity analysis described in Section 14.2.2. The purpose of the parametric sensitivity analysis was to determine the effect of one-at-a-time changes in various input parameters to understand the independent effect of each parameter on the results. In many cases, the parameter changes went outside the bounds of the realistic plant-specific conditions. However, the purpose of the parametric uncertainty quantification was to quantify the overall uncertainties associated with the input E3-66

Enclosure 3 Risk Quantification parameters. Therefore, the effect of simultaneous variations in multiple input parameters was considered, but none of the inputs were shifted beyond the bounds of realistic plant-specific conditions. The parametric uncertainties were quantified using a series of sensitivities with a bounding set of input parameters with respect to a) strainer failures and b) core failures. In cases where the bounding direction for a given input parameter (e.g., pool volume/level) could not be determined, both the minimum and maximum values were run. Table 3-13 shows the worst case conditions for strainer failures, and Table 3-14 shows the worst case conditions for core failures. Table 3-13 -Worst Case Conditions for Strainer Failure Bounding NARWHAL Base Sensitivity Case Parameter Direction Case Input Input Fiber Insulation Debris Consensus Same as NARWHAL Maximum Quantity (Maximum) Base Case Qualified Coatings Consensus Same as NARWHAL Maximum Debris Quantity (Maximum) Base Case Fire Barrier Debris Consensus Same as NARWHAL Maximum Quantity (Maximum) Base Case Unqualified Coatings Consensus Same as NARWHAL Maximum Debris Quantity (Maximum) Base Case Consensus Same as NARWHAL Latent Debris Quantity Maximum (Maximum) Base Case Miscellaneous Debris Consensus Same as NARWHAL Maximum Quantity (Maximum) Base Case Debris Transport Consensus Same as NARWHAL Maximum Fractions (Maximum) Base Case Minimum or Minimum and Pool Volume/Level Minimum Maximum Maximum Consensus Same as NARWHAL Containment Pressure Minimum (Minimum) Base Case Same as NARWHAL Base Case Minimum or Design Basis (Maximum is Pool Temperature Maximum (Maximum) Conservative Based on Parametric Sensitivity Results) Design (based on comparison of the ECCS Flow Rate Maximum Maximum pump curve and system resistance) CS Flow Rate (assuming Minimum Design Minimum sprays initiate) E3-67 J

Enclosure 3 Risk Quantification Bounding NARWHAL Base Sensitivity Case Parameter Direction Case Input Input Function of Water Function of Water ECCS/CS Switchover Minimum Volume and Flow Volume and Flow Time Rates Rates Hot Leg Switchover Procedural Step Same as NARWHAL N/A Time (Minimum) Base Case Minimum or Minimum and Secure CS Time Midpoint Maximum Maximum Consensus Same as NARWHAL Boil-off Flow Rate N/A (Maximum) Base* Case Maximum for Minimum or Same as NARWHAL pH release, Minimum for Maximum Base Case solubilitv Consensus Same as NARWHAL Head Loss Maximum (Maximum) Base Case Same as NARWHAL Structural Margin Minimum Design (Minimum) Base Case Consensus Same as NARWHAL NPSH Margin Minimum (Minimum) Base Case Pump Void Fraction Consensus Same as NARWHAL Minimum Limit (Minimum) Base Case Consensus Minimum (no Penetration Minimum (Maximum) penetration) Consensus Same as NARWHAL Core Fiber Limit N/A (Minimum) Base Case Maximum (95th LOCA Frequency Maximum Nominal (mean) percentile) Table 3-14 -Worst Case Conditions for Core Failure Bounding NARWHAL Base Sensitivity Case Parameter Direction Case Input Input Fiber Insulation Debris Consensus Same as NARWHAL Maximum Quantity (Maximum) Base Case Qualified Coatings Consensus Same as NARWHAL N/A Debris Quantity (Maximum) Base Case Fire Barrier Debris Consensus Same as NARWHAL N/A Quantity (Maximum) Base Case Unqualified Coatings Consensus Same as NARWHAL N/A Debris Quantity (Maximum) Base Case Consensus Same as NARWHAL Latent Debris Quantity Maximum (Maximum) Base Case Miscellaneous Debris Consensus Same as NARWHAL N/A Quantity (Maximum) Base Case E3-68

Enclosure 3 Risk Quantification Bounding NARWHAL Base Sensitivity Case Parameter Direction Case Input Input Debris Transport Consensus Same as NARWHAL Maximum Fractions (Maximum) Base Case Minimum or Minimum and Pool Volume/Level Minimum Maximum Maximum Consensus Same as NARWHAL Containment Pressure N/A (Minimum) Base Case Same as NARWHAL Base Case Minimum or Design Basis (Maximum is Pool Temperature Maximum (Maximum) Conservative Based on Parametric Sensitivity Results) Design (based on Minimum or comparison of the Minimum and ECCS Flow Rate Maximum pump curve and Maximum s*vstem resistance) CS Flow Rate (assuming Minimum Design Minimum sprays initiate) Function of Water Function of Water ECCS/CS Switchover Minimum Volume and Flow Volume and Flow Time Rates Rates Hot Leg Switchover Procedural Step Maximum Maximum Time (Minimum) Secure CS Time Minimum Midpoint Minimum Consensus Same as NARWHAL Boil-off Flow Rate Maximum (Maximum) Base Case Maximum for Minimum or Same as NARWHAL pH release, Minimum Maximum Base Case for solubility Consensus Same as NARWHAL Head Loss N/A (Maximum) Base Case Same as NARWHAL Structural Margin N/A Design (Minimum) Base Case Consensus Same as NARWHAL NPSH Margin N/A (Minimum) Base Case Pump Void Fraction Consensus Same as NARWHAL N/A Limit (Minimum) Base Case Consensus Same as NARWHAL Penetration Maximum (Maximum) Base Case Consensus Same as NARWHAL Core Fiber Limit Minimum (Minimum) Base Case E3-69

Enclosure 3 Risk Quantification Bounding NARWHAL Base Sensitivity Case Parameter Direction Case Input Input Maximum (951h LOCA Frequency Maximum Nominal (mean) percentile) Because minimum and maximum inputs were considered for both the pool volume and CS duration inputs, a 2x2 matrix of simulations was required for the bounding strainer failure cases. Similarly, because minimum and maximum inputs were considered for both the pool volume and RHR flow rate inputs, a separate 2x2 matrix of simulations was required for the bounding core failure cases. The difference in the bounding strainer and core failure cases compared to the NARWHAL base case are summarized below:

1. Strainer Failure Cases (2x2 Matrix)

a. Water Volume

i. Minimum (NARWHAL Base Case Inputs) ii. Maximum (Approximately 500,000 lbm Additional Water)

b. Maximum RHR Flow Rate (4,500 gpm)

c. Minimum CS Flow Rate (1,950 gpm)

d. CS Duration

i. Minimum (120 minutes) ii. Maximum (43,200 minutes)

e. Minimum Penetration (0 percent)

f. Maximum LOCA Frequency (95 1h Percentile)

2. Core Failure Cases (2x2 Matrix)

a. Water Volume

i. Minimum (NARWHAL Base Case Inputs) ii. Maximum (Approximately 500,000 lbm Additional Water)

b. RHR Flow Rate

i. Minimum (2,775 gpm) ii. Maximum (4,500 gpm)

c. Minimum CS Flow Rate (1,950 gpm)

d. Maximum hot leg switchover (HLSO) Time (563 minutes)

e. Minimum CS Duration (120 minutes)

f. Maximum LOCA Frequency (95 1h Percentile)

Table 3-15 shows the LiCDF for each of the parametric uncertainty cases. Figure 3-10 illustrates the change in LiCDF for each of the parametric uncertainty cases in comparison to the NARWHAL base case. E3-70

Enclosure 3 Risk Quantification

                    - - Resu Itsof Paramet"r1c uneertamty Ta bl e 3 15                                . t Q uan ff 1 1caf ion Change in Sensitivity                                                            ACDF from Description                 ACDF Case                                                                NARWHAL Base Case Strainer Case  1   Min water volume and min CS duration     1.22E-07        9.77E-08 Strainer Case  2   Min water volume and max CS duration     1.22E-07        9.77E-08 Strainer Case  3   Max water volume and min CS duration     1.19E-07        9.43E-08 Strainer Case  4   Max water volume and max CS duration     1.19E-07        9.42E-08 Min water volume and min RHR flow Core Case 1                                                 7.11 E-08       4.65E-08 rate Min water volume and max RHR flow Core Case 2                                                 1.16E-07        9.13E-08 rate Max water volume and min RHR flow Core Case 3                                                 7.23E-08        4.76E-08 rate Max water volume and max RHR flow Core Case 4                                                 1.12E-07        8.77E-08 rate I

E3-71

Enclosure 3 Risk Quantification Parametric Uncertainty sensitivity Results

                                             -    tk.Hmtling lnpuh  -  -   t\d~ C..t'iP 1.00E-06 1
     .... 1.00(-07 c
     ~

UX>E-08 (dS~ 1

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SensltivltyCase Figure 3-1 O - Comparison of Parametric Uncertainty Sensitivity Cases to the NARWHAL Base Case The process described above evaluates the parametric uncertainty in a very conservative manner by analyzing the worst-case combinations of input values. Although the scenario is hypothetically possible , the probability of all of the worst-case conditions occurring simultaneously is extremely unlikely. The overall results of this evaluation show that the parametric uncertainty is low. To meet the guidance in NUREG-1855 (Reference 11 ), model uncertainty must be addressed for any models or approaches for which no consensus exists . As discussed in Enclosure 1, Section 5.0, most of the GSl-191 models used for the VEGP evaluation are consensus models that have been widely used by the industry and accepted by the NRC . However, the following models used for VEGP are not consensus models and therefore were included in the model uncertainty quantification evaluation :

• Break model
• LOCA frequencies E3-72

Enclosure 3 Risk Quantification

• LOCA frequency allocation to individual welds
• CS actuation
• Aluminum metal release equation
• Fiber bed thickness required for chemical head loss
• LBLOCA size range discretization To address the uncertainty in these models, alternative models were evaluated as shown in Table 3-16.

T a bl e 3 16 - Alterna f 1ve M odes . t 1ry M o d e I U nee rtamty I Use d t 0 Q uan ff Model NARWHAL Base Case Sensitivity Case(s) Break model Continuum break model DEGB-only model VEGP PRA frequencies NUREG-1829 arithmetic LOCA frequencies (derived from NUREG-mean frequencies 1829 geometric mean) Hybrid allocation with multiple options (based on weld LOCA frequency allocation Top-down allocation degradation mechanism probability weighting) Multiple options including no Hot leg breaks larger than CS actuation breaks and all breaks larger 15inches than 2 inches WCAP-16530 release Aluminum metal release UNM release equation equation Fiber thickness required for 0.45 inches 0 inches chemical head loss Multiple options with a biased LBLOCA size range (6-15, 15-25, and 25-43.84 allocation of frequencies to discretization inches) smaller break sizes and larger break sizes E3-73

Enclosure 3 Risk Quantification Table 3-17 shows the ~CDF for each of the model uncertainty cases. Figure 3-11 illustrates the change in ~CDF for each of the model uncertainty cases in comparison to the NARWHAL base case.

                         - - Resu Itsof Mo deI Unee rtamcy T a bl e 3 17                           . t Q uan ff 1 1caf ion Change in Model with No                                                          .dCDF from Sensitivity Case           .dCDF Consensus                                                            NARWHAL Base Case Continuum Break Model       DEGB-Only Model                   8.10E-08        5.63E-08 Top-Down LOCA Top-Down LOCA Frequency Frequency Allocation Allocation with NUREG-1829        5.28E-07        5.04E-07 with Values from the Arithmetic Mean Values VEGP PRA Hybrid Allocation Methodology:

Skewed to High Rupture 4.90E-11 -2.46E-08 Probability Welds Methodology to Allocate Hybrid Allocation Methodology: LOCA Frequency to Skewed to High and Medium 3.55E-09 -2.11 E-08 Welds Rupture Probability Welds Hybrid Allocation Methodology: Spread Equally Across all Welds 2.47E-08 O.OOE+OO (top-down) All Breaks >15" 2.42E-08 -4.46E-10 Breaks Activating All Breaks >6" 2.39E-08 -7.61E-10 Containment Sprays All Breaks >2" 2.39E-08 -7.61E-10 No Breaks 2.70E-08 2.39E-09 UNM Aluminum Metal WCAP-16530 Equation 2.57E-08 9.98E-10 Release Equation 0.45-inch Fiber 0-inch Fiber Thickness Required Thickness Required for 6.74E-08 4.27E-08 for Chemical Head Loss Chemical Head Loss Bias 1 (6-10, 10-15, and 15-LBLOCA Size Range 5.13E-08 2.66E-08 43.84 inches) Discretization (6-15, 15-Bias 2 (6-20, 20-27, and 27-25, and 25-43.84 inches) 2.33E-08 -1.36E-09 43.84 inches) E3-74

Enclosure 3 Risk Quantification Model Uncertainty Sensitivity Results 100{ 06 I 00£. 07

  -    UJOIOll I..
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   ~   l(ll{()'l 100£.-10 1 OOE*ll DE<iB onty    NUREG* 1829 I

Sltnwt'd to ~d to Sptf!<id Al Sttih All 8ru*~ >6~ Al)81f!.D.s >Z~ ,.., 8'"Nh WUJl.* 16S30 Otn<t~ &t.b 1 Bi.u l At1th~llt H.ghl\""luflf H'istund cqu.iltyAnon >15" Mf!~n Prob;ab ty MtdH.lm ,a!I Wtkb W('lck llluptllflf P'fot>> ty Wtlth 9.re.;I( Model lOCA LOCA frequency Allouhon toWlkts Aluminum fiber LBUXA. Sil!' lt.ange Met.ii ~l'W.Ut thkkneu l>'~Jf!liZilttOO Rtquwtdfor Chf!miul lfff'Ct~ Hf!.ld Sen.sltlvltyC.11e "'" Figure 3 Comparison of Model Uncertainty Sensitivity Cases to the NARWHAL Base Case The results of this evaluation show that the model uncertainty is low (i.e ., the resulting b.CDF for each of the model uncertainty cases is within Region 3 as defined in RG 1.174). Because all of the cases that were evaluated for model uncertainty and parametric uncertainty resulted in a b.CDF less than 1x1 o-6 , it can be concluded with high confidence that the risk associated with GSl-191 is very low as defined by the acceptance guidelines in RG 1.174 (Reference 1). 14.3 PRA Model Uncertainty and Sensitivity The purpose of this section is to address the impact of PRA modeling epistemic uncertainty on the GSl-191 risk assessment. The baseline internal events and seismic PRA models document assumptions and sources of uncertainty, and these have been reviewed during the model peer reviews. Therefore, the approach taken was to review these PRA models and documentation to identify those items that may be directly relevant to the GSl-191 risk assessment, perform sensitivity analyses where appropriate , and discuss the results with dispositions for the uncertainties. E3-75

Enclosure 3 Risk Quantification 14.3.1 Internal Events PRA Model Uncertainty The epistemic uncertainty analysis approach described below applies to the Internal Events PRA. The baseline Internal Events PRA model uncertainty report was developed based on the guidance in NUREG-1855 (Reference 11 ). As described in NUREG-1855, sources of uncertainty include "parametric" uncertainties, "modeling" uncertainties, and "completeness" (or scope and level of detail) uncertainties. (The epistemic uncertainties unique to the seismic PRA are addressed in a later section.) Parametric uncertainty was addressed as part of the VEGP baseline PRA model aleatory uncertainty analysis. Assumptions are made during the PRA development as a way to address a particular modeling uncertainty because there is not a single definitive approach. The assumptions are defined consistent with the definition provided in NUREG-1855 (Reference 11 ). Plant-specific assumptions made for each of the VEGP Internal Events PRA technical elements are noted in the individual PRA notebooks. These assumptions were collected from each notebook and evaluated to determine if they are related to source of modeling uncertainty, and if so that uncertainty was characterized. In addition, EPRI TR-1016737 (Reference 12) compiled a listing of generic sources of modeling uncertainty for each PRA technical element, which were also considered. Completeness uncertainty addresses scope and level of detail of the PRA model. Uncertainties associated with scope and level of detail are documented in the PRA, but are only considered for their impact on a specific application. From the characterization of potential sources of uncertainty in the baseline Internal Events PRA model and of supplementary issues from EPRI TR-1 016737 (Reference 12), the following items may impact the internal events PRA results. Sensitivity analyses are included to further evaluate these items as a source of uncertainty. Table 3-18 provides a summary of the evaluation assessing the impact of the identified sources of model (epistemic) uncertainty on the GSl-191 risk assessment. For each of the sources of uncertainty, the potential impact on the GSl-191 risk assessment is addressed, either qualitatively or by an appropriate sensitivity case, to determine the impact on the GSl-191 application. E3-76

Table 3 Assessment of VEGP Internal Events PRA Epistemic Uncertainty Impacts Source of Epistemic Related Assumptions Sensitivity Case Disposition Uncertainty High RCS pressure Scenarios with significant Sensitivity cases were The GSl-191 risk impacts the potential for RCP seal leakage, a stuck performed for the assessment demonstrates induced steam generator open pressurizer valve, or a baseline Internal Events that only large LOCAs could tube rupture (SGTR). pressurizer PORV open for PRA by reclassifying the result in debris related feed-and-bleed cooling are identified scenarios as failures. Therefore, the Medium and large LOCA conservatively considered low RCS pressure to possible overestimation of and reactor vessel rupture high RCS pressure determine impact on induced SGTR for high are treated as low RCS scenarios. LERF. pressure scenarios has no pressure scenarios. All impact on the GSl-191 risk other core damage assessment. sequences are considered high pressure sequences where the induced SGTR failure mode is possible. Therefore, the baseline PRA model may overestimate the contribution of induced SGTR to LERF. Certain initiating events The generic industry None The GSl-191 phenomena can be affected by frequency for the LOSP are of concern for initiating seasonal variations (e.g., event developed in events that could generate loss of offsite power NUREG/CR-6890 is debris from insulation (LOSP), loss of service applicable to the VEGP site. materials and coatings inside water (SW), etc.) and The NSCW cooling towers containment, which could baseline PRA does not are not required during cold then be transported to the address seasonal weather months. containment sump and fail variations. the ECCS sump suction strainers during the recirculation phase needed

Enclosure 3 Risk Quantification Source of Epistemic Related Assumptions Sensitivity Case Disposition Uncertainty to maintain core cooling. For VEGP, the initiating events that meet these criteria are LOCAs and SSBI. Therefore, seasonal variations of certain other initiating events have no impact on the GSl-191 risk assessment. The method of calculation Detailed evaluations of The overall modeling Since the VEGP PRA model of human error probabilities HEPs are performed for the uncertainty associated is based on industry (HEPs) for the Human risk significant, pre- and with the general basis consensus modeling Reliability Analysis (HRA) post-initiator human failure for HEPs is addressed approaches for its HEP may introduce uncertainty events (HFEs) using by the standard baseline calculations, and there are based on the particular industry consensus PRA HEP sensitivity no additional HFEs added for methodology applied. methods. The Technique for cases for the internal the GSl-191 risk Human Error Rate events PRA. assessment, this is not Prediction (THERP) method considered a significant is applied for pre-initiator source of epistemic HFEs. The Cause-Based uncertainty and therefore Decision Tree Method has no impact on the GSI-(CBDTM) is used for 191 risk assessment. cognitive errors and THERP for execution errors for post-initiator HFEs. The VEGP PRA medium None A sensitivity was The LOCA frequency values LOCA frequency is based performed for the in NUREG/CR-6928 are in upon data from Internal Ev~nts to turn based on the LOCA NUREG/CR-6928, which is determine the impact of frequency data from an order of magnitude the increased medium NUREG-1829. NUREG-1829 higher than the previous LOCA frequency from data are used to develop E3-78

Enclosure 3 Risk Quantification Source of Epistemic Related Assumptions Sensitivity Case Disposition Uncertainty data used from NUREG/CR-5750. A LOCA frequencies for the NUREG/CR-5750. more than 10% increase GSl-191 risk assessment. in CDF and nearly 9% The GSl-191 risk impact, increase in LERF occurs however, is not sensitive to due to the updated data. the initiating event frequency for medium LOCAs. No medium LOCAs result in sump strainer or core failures due to the effects of debris. Steam generator (SG) tube If SG tube condition A sensitivity analysis The GSl-191 risk condition affects the degrades, the induced was performed with assessment demonstrates probabilities of induced SGTR probability during average vs. pristine SG that only large LOCA could SGTR. The current VEGP secondary side break or tube conditions. CDF result in sump strainer 3-18SG tube condition is anticipated transient without increased by slightly failure. Therefore, the pristine. scram for pressure- or more than 1%, while possible under-estimation of thermal-induced SGTR in LERF nearly tripled. induced SGTR for high the LERF analysis would pressure scenarios has no increase. impact on the GSl-191 risk assessment. The presence of water in The base internal events A sensitivity study was The risk increase for large the reactor cavity at the VEGP Level 2 PRA performed for a wet early release due to GSl-191 time of vessel breach assumes a dry reactor cavity reactor cavity. CDF is nearly three orders of would affect the probability condition. increased by less than magnitude below the RG of containment failures 2%, and LERF 1.174 Region Ill risk (early release due to steam increased by more than acceptance criteria. A 12% explosion and late release 12%. increase in the large early due to base mat melt release frequency would still through). be well below the Region Ill threshold. E3-79

Enclosure 3 Risk Quantification 14.3.2 Seismic PRA Model Uncertainty The ASME/ANS PRA Standard (Reference 13) and RG 1.200 (Reference 14) include a number of requirements related to identification and evaluation of the impact of assumptions and sources of uncertainty on the PRA results. NUREG-1855 (Reference

11) and EPRI TR-1016737 (Reference 12) provide guidance on assessment of uncertainty for applications of a PRA. Sources of uncertainty within the VEGP seismic PRA model are addressed as follows:

• Parametric uncertainty was addressed as part of the VEGP seismic PRA model quantification.
• Modeling uncertainties and associated assumptions specific to the VEGP seismic PRA technical elements are noted in the seismic PRA documentation and were subject to peer review.
• Completeness uncertainty addresses scope and level of detail. Uncertainties associated with scope and level of detail are documented in the seismic PRA. No specific completeness issues were identified in the VEGP seismic PRA peer review.

A summary of potentially important sources of uncertainty in the VEGP seismic PRA is provided in Table 3-19. E3-80

Enclosure 3 Risk Quantification Table 3-19 -Assessment of VEGP Seismic PRA Uncertainty mpacts PRA Summary of Treatment of Sources Potential Impact on Element of Uncertainty Seismic PRA Results Seismic The VEGP seismic PRA peer review team With regard to aleatory and Hazard noted that both the aleatory and epistemic epistemic uncertainties in the uncertainties were addressed by site response analysis, there characterizing the seismic sources. is an abundance of site-specific data from VEGP Units 3 and 4 that reduces epistemic uncertainty to an The review team commented that the site insignificant level. response analysis did not fully evaluate and model aleatory and epistemic The characterization of the uncertainties in the site response analysis. seismic hazard reasonably reflects sources of uncertainty. Seismic The seismic PRA peer review team had no Several sensitivity studies Fragilities comments on sources of uncertainty evaluate the impact of pertaining to fragilities. changes to fragilities on the seismic PRA results as one means of assessing the impact of fragilities uncertainties on the seismic PRA results. No changes to the model were recommended based on these results. Seismic The seismic PRA peer review team The seismic PRA PRA commented that the VEGP seismic PRA quantification report includes Model team relied on the UN CERT code for the a discussion of sources of propagation of the parametric uncertainties model uncertainty, and in the seismic PRA with little explanation or potentially important sources documentation of the meaning of the have been addressed in the uncertainties results. sensitivity analysis. No changes to the model were recommended based on these results. 15.0 References

1. Regulatory Guide 1.174, Revision 2, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," May 2011

2. NEI 04-07 Volume 2, Revision 0, "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'," December 2004 E3-81

Enclosure 3 Risk Quantification

3. NUREG-1829, "Estimating Loss-of-Coolant Accident (LOCA) Frequencies Through the Elicitation Process," April 2008

4. ANSl/ANS-5.1-1979, "American National Standard for Decay Heat Power in Light Water Reactors," August 1979

5. WCAP-17788-P, Revision 0, "Comprehensive Analysis and Test Program for GSl-191 Closure (PA-SEE-1090)," July 2015

6. WCAP-16530-NP-A, "Evaluation of Post-Accident Chemical Effects in Containment Sump Fluids to Support GSl-191," March 2008

7. NARWHAL-SUM-02, Revision 1, "NARWHAL Version 2.1 Software User's Manual,"

September 9, 2016

8. Howe, Kerry J., ET. Al, "Corrosion and solubility in a TSP-buffered chemical environment following a loss of coolant accident: Part 1 - Aluminum," Nuclear Engineering and Design, Volume 292, October 2015: 296-305

9. ML080230038, "NRC Staff Review Guidance Regarding Generic Letter 2004-02 Closure in the Area of Strainer Head Loss and Vortexing," March 2008

10. Regulatory Guide 1.82, Revision 4, "Water Sources for Long-Term Recirculation Cooling Following a Loss-of-Coolant Accident," March 2012

11. NUREG-1855, Revision 1, "Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decisionmaking," March 2017

12. EPRI Report 1016737, "Treatment of Parameter and Model Uncertainty for Probabilistic Risk Assessments," December 2008 13.ASME/ANS RA-Sa-2009, "Addenda to ASME/ANS RA-S-2008 Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications," February 2009

14. Regulatory Guide 1.200, Revision 2, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities,"

March 2009 E3-82

Vogtle Electric Generating Plant - Units 1 & 2 Supplemental Response to NRC Generic Letter 2004-02 Enclosure 4 Defense-in-Depth and Safety Margin

Enclosure 4 Defense-in-Depth and Safety Margin Table of Contents 1.0 Introduction 2.0 Defense-in-Depth 2.1 Evaluation for RG 1.174 Defense-in-Depth Philosophy 2.2 Detecting and Mitigating Adverse Conditions 2.3 Barriers for Release of Radioactivity 2.4 Emergency Plan Actions 3.0 Safety Margin 4.0 References E4-1

Enclosure 4 Defense-in-Depth and Safety Margin 1.0 Introduction For the purpose of this VEGP risk-informed GSl-191 submittal, defense-in-depth (DID) is defined as the response to the question of what happens if the analysis is wrong about a successful end state and it actually turns out to be a failure. DID includes mitigative design features and actions that address protection of the public from radiation due to sequences that go to failure (e.g., containment integrity, emergency plans, operator actions not credited in the GSl-191 evaluation, use of FLEX, etc.). Similarly, safety margin is defined as the response to the question of what aspects of the analysis increase confidence that a declared success is a success. Therefore, safety margin is a combination of built-in conservatisms that increase confidence that scenarios that go to success remain in success (and why some scenarios that are assumed to fail might actually succeed). The DID evaluation shows that there is adequate system capability to provide assurance that public health and safety are protected in the event that a LOCA results in strainer blockage or loss of long-term core cooling due to effects of LOCA-generated debris. It identifies operator actions that can be taken to mitigate the event and describes the robustness of the design for the VEGP containment buildings. The safety margin evaluation identifies many conservatisms throughout the evaluation, which provides high confidence that successful end states are truly successful, and that many end states that are assumed to fail in reality would also be successful. The conclusion of the evaluation is that there is substantial DID and safety margin. 2.0 Defense-in-Depth The evaluation of DID first addresses whether the impact of the proposed licensing basis (LB) change (individually and cumulatively) is consistent with the DID philosophy, as outlined in Regulatory Guide (RG) 1.174 (Reference 1). This section also presents the measures available to VEGP for preventing, detecting, and mitigating conditions that could challenge long-term core cooling due to strainer blockage and inadequate cooling flow to the reactor core. Finally, the evaluation shows if and how the proposed changes affect the barriers for release of radioactivity and emergency plan actions. 2.1 Evaluation for RG 1.174 DID Philosophy VEGP is proposing a licensing basis change to use a risk-informed approach to address the concerns of GSl-191 with respect to maintaining long-term core cooling following a LOCA. An evaluation was performed to determine whether the change meets the DID principles defined in RG 1.174 (Reference 1). As stated in the RG, consistency with the DID philosophy is achieved if the following occurs: i I L______ ~ E4-2

Enclosure 4 Defense-in-Depth and Safety Margin

• A reasonable balance is preserved among prevention of core damage, prevention of containment failure, and consequence mitigation. VEGP has performed various physical and procedural changes, for example, installation of new strainers with increased surface areas and a reduced opening size, increased RWST inventory to sump pool, removal of problematic insulation materials, procedural changes to delay isolation of RHR pumps from RWST, and program controls to ensure the debris load limits are not exceeded. Additional changes are being planned, for example, modifying the height of the RHR strainers and sump recirculation initiation sequence. These changes reduced the risk associated with the effects of LOCA-generated debris. The new risk-informed elements of the analysis showed a very small increase in risk of containment or reactor failures related to GSl-191, as demonstrated by the very small ~CDF and ~LERF per the RG 1.174 criteria (Reference 1). Therefore, the existing balance among prevention of core damage, prevention of containment failure, and consequence mitigation is preserved.
• Over-reliance on programmatic activities as compensatory measures associated with the change in the licensing basis is avoided. The proposed licensing basis change does not adversely impact any of the programmatic activities, such as the in-service inspection (ISi) program, plant personnel training, RCS leakage detection program, or containment cleanliness inspection activities. Therefore, the licensing change will not cause any over-reliance on these activities.
• System redundancy, independence, and diversity are preserved commensurate with the expected frequency, consequences of challenges to the system, and uncertainties (e.g., no risk outliers). As discussed above, the modifications made as part of the proposed licensing basis change do not change the redundancy, independence, and diversity of the ECCS or containment spray system. These systems have been fully analyzed relative to their contribution to nuclear safety through plant-specific PRA. The risk contribution related to GSl-191 due to the proposed licensing basis change has also been evaluated for the full spectrum of LOCA events. As described in Enclosure 3, Section 14.4, the uncertainties in the risk-informed approach were examined. Although the use of alternate models or variations in inputs can in some cases result in higher calculated LiCDF values, all uncertainty quantification cases showed risk results in Region Ill of RG 1.174 (Reference 1), which provides high confidence that there are no risk outliers.
• Defenses against potential common-cause failures are preserved, and the potential for the introduction of new common-cause failure mechanisms is assessed. The potential for new common-cause failure mechanisms has been assessed for the GSl-191 issues. The primary failure mechanism includes clogging of the sump strainers and/or reactor core, which is not a new failure mechanism. The defenses against these clogging mechanisms are not affected by the physical and procedural changes. Additionally, the new risk-informed E4-3

Enclosure 4 Defense-in-Depth and Safety Margin approach does not introduce any new common-cause failures or reduce the current plant defenses against common-cause failures.

• Independence of barriers is not degraded. The three barriers to a radioactive release are the fuel cladding, RCS pressure boundary, and reactor containment building. For the evaluation of a LOCA, the RCS barrier is postulated to be breached. The proposed licensing basis change for the use of a risk-informed approach to evaluate the effects of LOCA-generated debris does not affect the design and analysis requirements for the fuel. Therefore, the fuel barrier independence is not degraded.

The post-LOCA recirculation function is provided by the ECCS located inside the auxiliary building. During the recirculation phase, the RHR pumps take suction from the containment recirculation sumps and supply flow back to the reactor directly and/or through the CCPs and SI pumps. The pumps, system piping and other components on the recirculation flow path serve as the barrier to release. The auxiliary building has a dedicated ventilation system to control airborne radioactivity during emergency conditions and the building is capable of handling recirculating water leakage. The proposed licensing basis change does not alter the design and operating requirements for ECCS or auxiliary building. Analyses have been performed to show that, assuming a single failure that results in the loss of one air cooling train and one CS train, the containment fan coolers and the CS system can remove sufficient thermal energy from the containment atmosphere following a LOCA or MSLB to maintain the peak containment pressure below design values. The licensing basis change does not alter the design or operating requirements of these systems. It is therefore reasonable to conclude that the independence of the barriers is maintained and not degraded by the licensing basis change.

• Defenses against human errors are preserved. The use of the risk-informed methodology in the GSl-191 analysis does not impose any additional operator actions or increase the complexity of existing operator actions. Thus, the defenses that are already in place with respect to human errors are not impacted by the proposed licensing basis change.
• The intent of the plant's design criteria is maintained. The proposed licensing basis change does not alter any of the ECCS acceptance criteria specified in 10 CFR 50.46. Additionally, the proposed change does not affect the design or design requirements of the plant equipment associated with GSl-191. As discussed above, the risk-informed analysis shows that the risk increase due to GSl-191 related failures is very small and meets the RG 1.174 acceptance criteria (Reference 1). Therefore, the intent of the plant's design criteria is maintained.

E4-4

Enclosure 4 Defense-in-Depth arid Safety Margin 2.2 Detecting and Mitigating Adverse Conditions For the purposes of GSl-191 resolution, the primary regulatory objective is specified in 10 CFR 50.46(b)(5) as maintaining long-term core cooling. Adequate DID is maintained by ensuring the capability exists for operators to detect and mitigate adverse conditions due to potential impacts of debris blockage, such as inadequate flow through the strainers and/or through the reactor core. This section evaluates the VEGP DID measures for detecting and mitigating adverse conditions in order to support the VEGP application for a risk-informed approach to resolve GSl-191. Inadequate strainer flow refers to the condition where significant pump cavitation occurs due to inadequate RHR and/or CS pump NPSH margin associated with the high head losses across the sump strainers and debris bed. For VEGP, testing was performed to measure the debris bed head losses using a prototypical strainer configuration and post-LOCA conditions. The effect of debris head loss was conservatively accounted for in the risk-informed analysis. Inadequate reactor core flow refers to the condition where the normal core cooling flow path has become impeded (blocked) and is not allowing sufficient cooling water to reach the core. This condition could result from the formation of a debris bed at the reactor core inlet or at the fuel grid inside the core due to debris that passes through the sump strainers. The effect of debris accumulation in the reactor core was conservatively accounted for in the risk-informed analysis. 2.2.1 Prevention of Strainer Blockage The primary means to delay or prevent strainer blockage is to monitor and reduce the flow through the sump strainers as necessary, and control debris sources inside containment. Specific measures are laid out as follows.

• Various VEGP emergency operating procedures (EOPs) provide the operators with guidance on monitoring sump strainer blockage (e.g., Procedures 19010-C "E-1 Loss of Reactor or Secondary Coolant", 19013-C "ES-1.3 Transfer to Cold Leg Recirculation", and 19111-C "ECA-1.1 Loss of Emergency Coolant Recirculation"). If sump blockage is detected, Procedure 19113-C "ECA-1.3 Recirculation Sump Blockage" provides actions that operators should take to respond to the condition.
• VEGP EOPs incorporated the Bulletin 2003-01 training and procedural guidance to expedite plant cooldown in response to a small break LOCA.
• For small to medium break LOCAs, depletion of the RWST,can be delayed by following Procedure 19012-C "ES-1.2 Post LOCA Cooldown and Depressurization". This procedure provides actions to cool down and depressurize the RCS to reduce the break flow, thereby lowering the injection flow necessary to maintain RCS subcooling and inventory. It is possible to bring the plant to cold shutdown conditions before the RWST is drained to the sump E4-5

Enclosure 4 Defense-in-Depth and Safety Margin recirculation switchover level. Therefore, sump recirculation may not be required and, in that case, sump blockage would not be an issue. 1

• The Technical Specification minimum required RWST volume is 686,000 gallons, and the low-level alarm setting is 696,448 gallons. The RWST level is normally maintained above the low-level alarm setting, and the nominal volume of the tank is 715,000 gallons.

Several measures are in place to control the debris sources inside the VEGP containment buildings.

• Training is provided to personnel accessing containment to raise their awareness of the more stringent containment cleanliness requirements, the potential for sump blockage, and actions being taken to address sump blockage concerns.
• For the Technical Requirements Manual Surveillance Requirement (TRS) 13.5.1.1, VEGP has implemented Procedure 14900-C, "Containment Exit Inspection" in conjunction with 00303-C, "Containment Entry". Per these procedures, prior to entering Mode 4 (Hot Shutdown) from Mode 5, several walk-downs are required to be performed by station management and operations personnel to ensure the containment buildings are free of loose debris. For subsequent entries, inspections of the travel path and work locations are required to ensure the areas are free of loose debris.
• For the Technical Specification Surveillance Requirement (SR) 3.5.2. 7, VEGP has implemented procedures 14903-1/2 both titled "Containment Emergency Sump Inspection", to verify by visual inspection that the suction inlets are not restricted by debris and that the sump strainers are correctly configured according to plant design and show no structural distress or abnormal corrosion.

These procedures also ensure that the protective covers for the TSP baskets and sump strainer are removed. This inspection is required on an 18-month frequency in accordance with the Surveillance Frequency Control Program.

• VEGP Procedure 00309-C is used to control unattended temporary materials in containment. The program includes periodic surveillance and assessment of containment material conditions during Modes 1-4. It imposes strict controls on the types and quantities of materials that may be taken into containment.
• Inspections of the coatings in containment are part of a protective coatings program complying with Regulatory Guide 1.54 (Reference 2) and ANSI N 101.4-1972 (Reference 3), to ensure that coatings do not adversely affect safety-related systems, structures, or components.

2.2.2 Detection of Strainer Blockage During sump recirculation following a LOCA, accumulation of fiber, particulate, and chemical debris on the strainer could cause high flow head losses which may challenge the operation of the RHR and CS pumps. This, in turn, could result in a condition where insufficient cooling is provided for reactor core cooling and/or containment pressure control. When such a condition exists, it is important for the plant operators to be able E4-6

Enclosure 4 Defense-in-Depth and Safety Margin to detect this condition in a timely manner. VEGP maintains a post-accident monitoring instrumentation program, which ensures the capability to monitor plant variables and system status during and following an accident. This program includes those instruments that indicate system status and furnish information regarding the release of radioactive materials, in accordance with Regulatory Guide 1.97 Revision 2 (Reference 4). VEGP has the following methods for detection of sump strainer blockage conditions.

• VEGP has indications in the control room for SI, RHR, and CS pump flows and SI and CS pump discharge pressures. Instrumentation is available to provide the operators with indications of pump cavitation, such as erratic flow or low discharge pressure.
• VEGP has core exit thermocouple (CET) and reactor vessel level indications in the control room to allow monitoring for any potential reduction in core cooling flow due to sump blockage. This indication is also displayed on the computer systems as part of the critical safety system status tree indicators, monitored by the reactor operators and shift technical advisor. The status tree indicators provide changes based on status tree logic to enhance operator recognition of a distress condition developing.

2.2.3 Mitigation of Strainer Blockage Multiple methods are available to mitigate an inadequate recirculation flow condition caused by the accumulation of debris on the sump strainer.

• The VEGP EOPs contain steps to reduce flow through the system up to and including stopping all pumps taking suction from a clogged sump strainer. It has been observed, during strainer head loss testing, that stopping all flow through a debris-laden strainer could dislodge portions of the debris bed from the strainer because the force that holds the debris bed in place was the flow head loss through the debris. This is also an important measure to avoid permanent pump damage that could be caused by the loss of suction condition.
• VEGP Procedure 19111-C "ECA-1.1 Loss of Emergency Coolant Recirculation" minimizes the pumps required depending on plant conditions and directs shutting down all pumps as applicable. Procedure 19113-C "ECA-1.3 Recirculation Sump Blockage" also contains steps to shut down SI pumps and CCPs that piggyback off a potentially cavitating RHR pump during recirculation.
• VEGP Procedure 19113-C "ECA-1.3 Recirculation Sump Blockage" contains steps to initiate makeup to the RWST from, for example, the spent fuel pool.

This would allow realignment of SI and CS pumps to the direct injection flow path from the RWST and provide necessary cooling for an extended period. The operators would establish the minimum flow required for core decay heat removal depending on sub-cooling conditions.

• In response to the Nuclear Regulatory Commission (NRC) Order EA-12-049 (Reference 5), "Mitigation Strategies for Beyond-Design-Basis External Event E4-7

Enclosure 4 Defense-in-Depth and Safety Margin (BDBEE)", VEGP developed diverse and flexible coping strategies (FLEX) to maintain fuel cooling (spent fuel pool and core) and containment integrity. Various modifications have been implemented such that non-emergency equipment can be credited during a BDBEE. For example, the Auxiliary Feedwater System can be used to deliver cooling water from the condensate storage tank (CST) to the steam generators for reactor core cooling. Makeup capabilities were added to refill the CST and Reactor Make-up Water Storage Tank (RMWST), which would serve as suction sources for core cooling. 2.2.4 Prevention of Inadequate Reactor Core Flow The set of actions identified in Section 2.2.1 for reducing or controlling flow through the emergency sump strainers during the recirculation phase can have a similar positive impact on reducing the potential for fuel blockage. Controlling flow to the reactor vessel to maintain fuel coverage and match decay heat has benefits through reduced head loss and delayed onset of any chemical precipitates. The VEGP plant design has simultaneous hot leg and cold leg injection once the RWST is depleted and the RHR and SI pumps have been realigned during the recirculation phase. Initially all of the ECCS pumps would be aligned for cold leg injection. At 7.5 hours after the initiating event, the switchover to simultaneous hot/cold leg injection would be made. For this configuration, the RHR and SI pumps provide cooling water through the hot leg while the CCP continues injecting coolant through the cold leg. It is expected that, with most of the flow traveling through the hot leg, the motive force that holds the debris at the core inlet would be removed and the flow from the hot legs would travel down the heated core to the inlet, which could dislodge the debris bed at the core inlet. 2.2.5 Detection of Inadequate Reactor Core Flow Multiple methods exist for detection of a core blockage condition as manifested by an inadequate RCS inventory or inadequate RCS and core heat removal conditions. The primary methods for detection include core exit thermocouple (GET) temperature indication and reactor water level, as monitored by the reactor vessel level instrumentation system (RVLIS). An additional method for detection of a core blockage condition includes monitoring of containment radiation levels.

• Core exit temperature behavior is the primary indicator of adequate core cooling.

If cold leg recirculation has been established with flow maintained into the RCS, core exit temperature should be stable or slowly lowering during accident recovery. Increasing core exit temperatures while injection flow is maintained, regardless of reactor vessel water level behavior, could be an indication of insufficient core flow. In this regard, VEGP's functional restoration procedure would attempt to establish injection flow of clean water from the RWST. CETs E4-8

Enclosure 4 Defense-in-Depth and Safety Margin are monitored during EOP usage as well as for status tree functional restoration entries and the safety parameter display system (SPDS).

• Reactor vessel water level is also monitored and a decreasing water level could indicate a lower core region flow blockage. VEGP employs the RVLIS to provide instrumentation for the detection of inadequate core cooling. The RVLIS utilizes two sets of differential pressure cells to measure reactor vessel level continuously. The measurement provides an approximate indication of the relative void content or density of the circulating fluid.
• Increasing radiation levels are indicated by alarms in the control room with specific procedural steps in both alarm response procedures and EOPs for addressing the condition. Radiation monitor indication in the auxiliary building may be indication of a LOCA outside containment or provide initial entry conditions due to increasing radiation levels. Abnormal containment radiation could be an indication of fission product barrier degradation, which is monitored by the control room. Due to the sensitivity of the monitors and the low alarm set points, identification of degrading core conditions is expected well before a significant release of radioactivity to containment occurs.

2.2.6 Mitigation of Inadequate Reactor Core Flow Multiple methods are available to mitigate an inadequate reactor core flow condition, as laid out in Procedures 19221-1/2 "FR-C.1 Response to Inadequate Core Cooling" and 19222-1/2 "FR-C.2 Response to Degraded Core Cooling". Upon identification of an inadequate RCS inventory or an inadequate core heat removal condition, the EOPs direct the operators to take actions to restore cooling flow to the RCS including:

• Reestablish SI flow to the RCS
• Reduce RCS pressure by performing rapid secondary depressurization
• Restart RCPs and open pressurizer PORVs These actions are to be performed sequentially. Success, as indicated by improved core cooling and increasing vessel inventory, is evaluated prior to performing the next action in the sequence. Re-initiation of high pressure SI may be, depending on the cause of inadequate core cooling, the most effective method to recover the core and restore adequate core cooling. If some form of high-pressure injection cannot be established or is ineffective in restoring adequate core cooling, the operator takes actions to reduce the RCS pressure in order for the SI accumulators and low-head pumps to inject. Analyses have shown that a rapid secondary depressurization is the most effective means for achieving this objective. If secondary depressurization is not possible, or primary to secondary heat transfer is significantly degraded, and at least one idle SG is available, the operator can start the RCP(s) associated with the available idle SG(s). The RCPs will provide forced two-phase flow through the core and temporarily improve core cooling until some form of makeup flow to the RCS can be established.

E4-9

Enclosure 4 Defense-in-Depth and Safety Margin VEGP has also implemented procedures per the severe accident management guidelines (SAMG) which provide the operator with actions to protect fission product boundaries and return the plant to a controlled stable condition when the emergency operating procedures are no longer effective in controlling the casualty. Entry into the SAMG procedures is directed by the emergency operating procedures when certain conditions are met. Some of the operator actions outlined in the SAMG procedures can help maintain reactor core flow, for example, injection into SGs and RCS, depressurization of RCS, makeup to RWST, realignment to injection from RWST, and flooding the containment. Cooling can also be provided to the reactor core using the flow paths established by the FLEX strategy or by reinitiating injection through a refilled RWST, as discussed in Section 2.2.3. If it is determined that the inadequate core cooling condition is caused by clogged sump strainers, the actions discussed in Section 2.2.3 can also be taken to reestablish cooling flow through the strainers. 2.3 Barriers for Release of Radioactivity The purpose of this section is to demonstrate that there are additional defense in depth measures to protect the current barriers for release of radioactivity. The three barriers are the fuel cladding, the RCS boundary, and the reactor containment building. Each of these barriers is addressed in the subsections below. 2.3.1 Fuel Cladding Following a LOCA, the ECCS provides both the initial phase of accident mitigation and long-term cooling to the fuel cladding barrier. For the initial phase of accident mitigation, the proposed licensing basis change for the use of a risk-informed approach to evaluate the effects of debris does not alter the fuel cladding limits, or previous analysis and testing programs that demonstrate the acceptability of ECCS. The primary goal of the VEGP SAMG procedures is to protect fission product boundaries and mitigate any ongoing fission product releases in the event that conditions warrant entry into the SAMGs. Some of the operator actions outlined in the SAMG procedures can help maintain reactor core flow and integrity of the fuel cladding, for example, injection into SGs and RCS, depressurization of RCS, makeup to RWST, realignment to injection from RWST and flooding the containment. 2.3.2 RCS Pressure Boundary The integrity of the RCS pressure boundary is assumed to be compromised for the GSl-191 sump performance evaluation. However, the proposed licensing basis change does not modify the previous analyses or testing programs that demonstrate the integrity of the RCS. Additional measures are in place to prevent and detect pipe breaks, as discussed below. E4-10

Enclosure 4 Defense-in-Depth and Safety Margin

• The inservice inspection (ISi) program provides rules for the examination and repair of piping and other RCS components, and plays an important role in the prevention of pipe breaks. The integrity of the Class 1 welds, piping, and components are maintained at a high level of reliability through the ASME Section XI inspection program (Reference 6). VEGP ISi procedures also ensure that inspections are performed in accordance with the schedule requirements of the code.
• RCS overpressure protection is provided by the pressurizer safety valves, steam generator safety valves, and the reactor protection system and associated equipment. Combinations of these systems ensure compliance with the overpressure requirements of the ASME Boiler and Pressure Vessel Code, Section Ill, Paragraphs NB-7300 and NC-7300, for pressurized water reactor systems (Reference 7).
• The leak detection program at VEGP is capable of early identification of RCS leakage in accordance with RG 1.45 (Reference 8) to provide time for appropriate operator action to identify and address RCS leakage. The effectiveness of this program is not reduced by the proposed licensing basis change to the risk-informed approach for GSl-191.
• Some of the operator actions outlined in the VEGP SAMG procedures can help maintain integrity of the RCS when directed by the emergency operating procedures. Such actions include injection into SGs and RCS, depressurization of RCS, makeup to RWST, realignment to injection from RWST and flooding containment.

2.3.3 Reactor Containment Integrity The VEGP containment buildings are designed such that for all break sizes, up to and including a double-ended guillotine break of an RCS pipe or secondary system pipe, the containment peak pressure is below the design pressure with adequate margin. This has been demonstrated by previous analyses based on conservative assumptions (e.g., minimum heat removal and maximum containment pressure). The analyses also considered the worst single active failure affecting the operation of the ECCS, CSS, and containment fan coolers during the injection phase, and the worst active or passive single failure during the recirculation phase. For primary system breaks, loss of offsite power is also assumed. The analyses showed that the containment fan coolers, in conjunction with the CS system, can remove sufficient thermal energy and decay heat from the containment atmosphere following a LOCA or MSLB to maintain the containment pressure below design values. Therefore, the containment buildings remain a low leakage barrier against the release of fission products for the duration of the postulated LOCAs. The evaluation of post-LOCA debris effects using a risk-informed approach is not part of the analyses that demonstrate containment integrity. The proposed licensing basis change does not affect the methodology, acceptance criteria, or conclusion of the existing analysis. Therefore, the reactor containment integrity is not affected. E4-11

                                                                                         ,j

Enclosure 4 Defense-in-Depth and Safety Margin Additionally, some of the operator actions outlined in the VEGP SAMG procedures can help maintain integrity of the containment when directed by the emergency operating procedures. Such actions include control of containment pressure and hydrogen concentration. 2.4 Emergency Plan Actions The proposed change to the licensing basis to use the methodology of a risk-informed approach does not involve any changes to the emergency plan. There is no change to the strategies for preventing core damage and containment failure, or for consequence mitigation. The use of the risk-informed approach does not impose any additional operator actions or complexity. Implementation of the proposed change would not result in any changes to the response requirements for emergency response personnel during an accident. 3.0 Safety Margin There are numerous conservatisms used throughout the risk-informed GSl-191 evaluation for VEGP. However, not all of these conservatisms were classified as safety margin. Some conservatisms were included to provide future operating margin (i.e., margin added to the current plant conditions to allow for future changes, and flexibility in conducting maintenance or inspections). The key distinction between safety margin and operating margin is that safety margin cannot be reduced without approval from the NRG (Reference 1), whereas operating margin can be modified if necessary based on plant changes. Table 4-1 describes the safety margins included in the risk-informed GSl-191 evaluation. As noted in this table, there are many conservatisms throughout the evaluation, which provide high confidence that successful end states are truly successful, and that many end states that are assumed to fail in reality would also be successful. Note that in several places, the effect of conservatism on the model is described as over-predicting or under-predicting a specific physical phenomena or failure. These terms are generically used to refer to either a change in the actual value predicted by the model or an increase in margin. For example, flashing is not predicted to occur. However, because of conservatism in specific portions of the model, the potential for flashing is over-predicted (i.e., there is more real margin to prevent flashing than is predicted by the model). E4-12 J

Enclosure 4 Defense-in-Depth and Safety Margin Table 4 Description of Safety Margin Conservatism Credited as Topic Realistic Conditions Impact on Evaluation Safety Margin Scenario All frequency associated with Smaller breaks on main steam Overall likelihood of failure is Frequency secondary side breaks is and feedwater piping are much over-predicted for secondary allocated to DEGBs more likely than DEGBs, and side breaks would generate significantly lower debris quantities Scenario Random pump failures are Random pump failures can Failures at the start of the Frequency assumed to occur at switchover occur at the beginning of the event would delay switchover to recirculation event, at the start of to recirculation, failures later recirculation or any time during in the event would result in the event distribution of debris across more strainers Thermal- Initial containment pressure is at Containment pressure would NPSH margin is under-Hydraulics technical specification (TS) be above TS minimum predicted and degasification minimum of -0.3 psig and flashing are over-predicted Thermal- No credit taken for containment The post-LOCA containment . NPSH margin is under-Hydraulics accident pressure in NPSH pressure would be significantly predicted and degasification calculations and minimal credit higher than the saturation and flashing are over-taken for degasification and pressure predicted flashing calculations E4-13

Enclosure 4 Defense-in-Depth and Safety Margin Conservatism Credited as Topic Realistic Conditions Impact on Evaluation Safety Margin Thermal- Design basis containment and Containment and sump Chemical release, precipitate Hydraulics sump temperature profiles used temperature profiles would be quantities, degasification, and for all break sizes significantly lower for smaller flashing are over-predicted, break sizes and aluminum solubility, corrected strainer head loss, and NPSH margin are under-predicted (all impacts are conservative with the exception of under-predicted aluminum solubility and corrected strainer head loss; however, the sensitivity analysis described in Enclosure 3, Section 14.3 showed that the conservative effects outweigh the non-conservative effects) Debris With the exception of shadowing Full offset of pipe DEGBs Quantity of debris generated Generation by concrete walls, no credit was (especially on the primary loop inside the ZOI is over-taken for structures or restraints piping) would be significantly predicted that would limit the quantity of limited due to physical debris generated within a break restraints; also, insulation and ZOI qualified coatings would not be completely destroyed within a given ZOI due to the shielding effects of equipment and other structures Debris 100% failure of unqualified Some types of unqualified Particulate debris quantity on Generation coatings for all breaks coatings may have a relatively strainers is over-predicted low failure fraction E4-14

Enclosure 4 Defense-in-Depth and Safety Margin Conservatism Credited as Topic Realistic Conditions Impact on Evaluation Safety Margin Debris Unqualified epoxy fails as 100% Epoxy coatings are likely to fail Unqualified coatings debris Generation particulate in a range of sizes (including transport and particulate both particulate and chips) debris quantity on strainers are over-predicted Debris Unqualified coatings fail at the Unqualified coatings would fail Unqualified coatings that fail Generation start of the accident gradually and may not fail until in upper containment after much later in the event sprays are secured would not transport Chemical Maximum pH for chemical Consistent time-dependent pH Precipitate quantity is over-Effects release and minimum pH for profile resulting in lower predicted and precipitates solubility release and/or increased would form later than solubility predicted Chemical No aluminum remains in solution Some breaks would never Aluminum precipitate quantity Effects after the solubility limit has been exceed the solubility limit, and and strainer head loss are reached or 24 hours (whichever breaks that do exceed the over-predicted comes first) solubility limit would still have some aluminum in solution Chemical All insulation debris is assumed In reality, a large fraction of the Aluminum and calcium Effects to be in the sump for the chemical debris would be captured in release from insulation are release calculation upper containment, and the over-predicted, resulting in an release of chemicals would be over-prediction of aluminum significantly reduced for breaks and calcium precipitate where containment sprays are quantities not initiated E4-15

Enclosure 4 Defense-in-Depth and Safety Margin Conservatism Credited as Topic Realistic Conditions Impact on Evaluation Safety Margin Debris Fine debris has a high A condensate washdown of The quantity of fine debris Transport condensate washdown fraction 1% is a realistic estimate washed down to lower (10%) when sprays are not (Reference 10). containment (and initiated subsequently transported to the strainers and core) is over-predicted for breaks that do not initiate containment sprays Debris Fine debris has a high spray Some fine debris would be The quantity of fine debris Transport washdown fraction (100%) when blown to locations shielded washed down to lower sprays are initiated from containment sprays and containment (and would be retained in these subsequently transported to locations for the duration of the the strainers and core) is event over-predicted for breaks that initiate containment sprays Debris Fine debris has a high Some fine debris would settle The quantity of fine debris Transport recirculation transport fraction and be retained in stagnant transported to the strainers (100%) for all breaks regions of the recirculation and core is over-predicted pool (especially for cases where fewer pumps are operating) Debris Small and large pieces of Sustained movement of a The quantity of small and Transport fiberglass transport at the piece of debris all the way to large piece debris transported incipient tumbling velocity for the the strainer would require a to the strainers is over-respective debris sizes (note that somewhat higher fluid velocity, predicted the incipient tumbling velocity is particularly in cases where defined as the minimum fluid large debris quantities velocity at which an individual (including a mixture of sizes) piece would begin to move would result in agglomeration (Reference 11 )) of the debris on the containment floor E4-16

Enclosure 4 Defense-in-Depth and Safety Margin Conservatism Credited as Topic Realistic Conditions Impact on Evaluation Safety Margin Debris Small and large pieces of Based on 30-day erosion test The quantity of fines Transport fiberglass debris have a high results, the erosion fraction for generated and subsequently containment pool erosion fraction small pieces of fiberglass transported to the strainers (10%) would be somewhat less than and core is over-predicted 10% and the erosion fraction for large pieces of fiberglass would be less than small pieces Strainer/Pump A strainer is assumed to fail any In many cases, one type of The breaks that fail the Failures time the accumulated debris debris (e.g., calcium - strainer acceptance criteria quantities exceed the tested phosphate) exceeds the tested are over-predicted debris quantities quantity while other types of debris (e.g., sodium aluminum silicate) are significantly below the tested quantity; also, most breaks have available margin to accommodate higher head losses Strainer/Pump Miscellaneous debris (e.g., tags It is likely that a large portion of The strainer surface area is Failures or labels) all transports to the the miscellaneous debris under-predicted, and strainer strainers prior to any other debris would not transport to the head loss and debris limit and reduces the effective strainer strainers, and any failures are over-predicted area miscellaneous debris that does transport would tend to arrive along with or after other debris E4-17

Enclosure 4 Defense-in-Depth and Safety Margin Conservatism Credited as Topic Realistic Conditions Impact on Evaluation Safety Margin Strainer/Pump Debris head loss was Head loss would increase The strainer head loss for Failures conservatively calculated using a gradually as debris both conventional and rule-based approach (i.e., if the accumulates and most breaks chemical debris is over-accumulation of a given debris would not accumulate enough predicted type exceeds a certain threshold, debris to reach the head a bounding head loss is losses that were applied automatically applied) Strainer/Pump Strainer head loss testing was Because flow preferentially The strainer head loss for Failures conservatively performed using a passes through the lower both conventional and strainer module with fewer disks disks, it is likely that a larger chemical debris is over-and scaled up to the full height quantity of debris could predicted strainers based on the area ratios accumulate on the full height strainers than predicted using a simple area ratio Strainer/Pump Calcium phosphate head loss Based on observations from The strainer head loss is Failures was applied for all breaks that autoclave tests described in over-predicted for calcium generate and transport a Enclosure 2 Section 3.o.2.9 phosphate debris sufficient quantity of fiber debris and other tests representative of VEGP conditions, calcium phosphate precipitation is either unlikely to occur or the actual precipitates would have a negligible effect on head loss Strainer/Pump The chemical head loss was The head loss associated with The strainer head loss is Failures extrapolated to 30 days and the the full chemical debris load is over-predicted early in the extrapolation constant was not likely to continue event applied 450 minutes after the increasing over 30 days, and start of the event even if it did, additional NPSH margin would be available later in the event as the pool temperature drops E4-18

Enclosure 4 Defense-in-Depth and Safety Margin Conservatism Credited as Topic Realistic Conditions Impact on Evaluation Safety Margin Strainer/Pump Strainer failure is assumed in all It is likely that the strainer Strainer structural failures are Failures cases where the head loss meets could withstand higher head over-predicted or exceeds the structural margin losses than predicted, and of the strainer even if a structural failure occurs, it may not result in a complete loss of functionality Strainer/Pump All gas voids formed by Due to the relatively low Gas void fractions at the Failures degasification were assumed to Froude number, gas voids are pumps are over-predicted transport to the pumps likely to accumulate in the strainer and vent back to the pool when the buoyancy of the accumulated air exceeds the strainer head loss Strainer/Pump Pump NPSH required was Small gas void fractions would Pump NPSH required is over-Failures adjusted for gas voids based on likely have a much smaller predicted and Pump NPSH very conservative guidance effect on NPSH required margin is under-predicted (Reference 13) when gas voids are present Core Failures The fiber penetration correlation The penetration of fiberglass Fiber penetration (and ignores effects of fiber and fines would be reduced by the subsequent accumulation on particulate interactions and accumulation of particulate the core) is over-predicted accumulation of pieces of and fiberglass pieces on the fiberglass strainer Core Failures The WCAP-16793-N P The peak cladding The peak cladding methodology for evaluating peak temperature and total temperature and total cladding temperature and total deposition thickness would be deposition thickness are over-deposition thickness due to lower predicted accumulation of debris on the fuel rods is conservative (Reference 14) E4-19 I_

Enclosure 4 Defense-in-Depth and Safety Margin Conservatism Credited as Topic Realistic Conditions Impact on Evaluation Safety Margin Core Failures Maximum boil-off flow rate with Debris transport to core inlet Debris accumulation on core additional 20% margin used to would be proportional to boil- inlet is over-predicted for cold calculate debris accumulation on off flow rate and actual boil-off leg breaks core inlet for cold leg breaks flow rate is likely to be lower Core Failures Fiber limits associated with core It is likely that significantly Core failures due to the blockage and boron precipitation larger quantities of debris accumulation of fiber debris are based on bounding tests and could accumulate in the core are over-predicted. analyses without resulting in core damage E4-20

Enclosure 4 Defense-in-Depth and Safety Margin For information, Table 4-2 shows the operating margin included in the analysis for various types of debris and exposed aluminum surface areas.

                               - - Descr1p" fion of O1peraf In!

T a bl e 4 2 M argm Actual Item Value Used Operating Margin Value Epoxy Unqualified 2700.6 lbm 2,729 lbm 28.4 lbm Coatings Alkyd Unqualified 30.6 lbm 591bm 28.4 lbm 0.753 ft 3 Coatings IOZ Unqualified 27.6 lbm 56 lbm 28.4 lbm Coatinos Latent Debris 60 lbm 2001bm 1401bm Miscellaneous 4 ft2 50 ft2 46 ft2 Debris Unsubmerged 741.3 ft 2 926.6 ft 2 185.3 ft2 Aluminum Metal Submerged 278.7 ft2 348.4 ft2 69.7 ft2 Aluminum Metal E4-21

4.0 References

1. Regulatory Guide 1.174, Revision 2, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," May 2011

2. Regulatory Guide 1.54, Revision 2, "Service Level I, II, and Ill Protective Coatings Applied to Nuclear Power Plants," October 2010

3. ANSI N101.4-1972, "Quality Assurance for Protective Coatings Applied to Nuclear Facilities," American National Standards Institute, Washington, DC

4. Regulatory Guide 1.97, Revision 2, "Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions during and Following an Accident," December 1980

5. EA-12-049, "Issuance of Order to Modify Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events," March 12, 2012

6. ASME Boiler and Pressure Vessel Code, Section XI, "Rules for lnservice Inspection of Nuclear Power Plant Components," 2010 Edition, July 1, 2010

7. ASME, Boiler and Pressure Vessel Code, Section Ill, "Rules for Construction of Nuclear Power Plant Components," 2010 Edition, July 1, 2010

8. Regulatory Guide 1.45, Revision 1, "Guidance on Monitoring and Responding to Reactor Coolant System Leakage," May 2008

9. Not used

10. NUREG/CR-7172, "Knowledge Base Report on Emergency Core Cooling Sump Performance in Operating Light Water Reactors," January 2014

11. NUREG/CR-6772, "GSl-191: Separate-Effects Characterization of Debris Transport in Water," August 2002 12.Not Used

13. Regulatory Guide 1.82, Revision 4, "Water Sources for Long-Term Recirculation Cooling Following a Loss-of-Coolant Accident," March 2012 14.WCAP-16793-NP-A, "Evaluation of Long-Term Cooling Considering Particulate, Fibrous and Chemical Debris in the Recirculating Fluid", Revision 2, July 2013.

E4-22

CAW-17~4565 AFFIDAVIT COMMONWEALTH OF PENNSYLVANIA: SS COUNTY OF BUTLER:

• I, James A. Gresham, am aµthorized to ex\:)cute tliis Affidavit onbehalfofWestinghouse Electric .
• Company LLC ("Westinghouse~') and. declare that the averments of fact set forth in this. Affidavit are true and correct to the best of my knowledge, information, arid belief.

Executed on: 1/l~f 1

                                                      *f J~e~ A. Gresham, Manager
                                                    .//      . .. .      .. .     .

Regulatory Compliance J

3 CAW-17-4565 (1) I am Manager, Regulatory Compliance, Westinghouse Electric Company LLC ("Westinghouse"), and as such, I have been specifically delegated the function of reviewing the proprietary information sought to be withheld from public disclosure in connection with nuclear power plant licensing and rule making proceedings, and am authorized to apply for its withholding on behalf of Westinghouse. (2) I am making this Affidavit in conformance with the provisions of 10 CFR Section 2.390 of the Nuclear Regulatory Commission's ("Commission's") regulations and in conjunction with the Westinghouse Application for Withholding Proprietary Information from Public Disclosure accompanying this Affidavit. (3) I have personal knowledge of the criteria and procedures utilized by Westinghouse in designating information as a trade secret, privileged or as confidential commercial or financial information. (4) Pursuant to the provisions of paragraph (b)(4) of Section 2.390 of the Commission's regulations, the following is furnished for consideration by the Commission in determining whether the information sought to be withheld from public disclosure should be withheld. (i) The information sought to be withheld from public disclosure is owned and has been held in confidence by Westinghouse. (ii) The information is of a type customarily held in confidence by Westinghouse and not customarily disclosed to the public. Westinghouse has a rational basis for determining the types of information customarily held in confidence by it and, in that connection, utilizes a system to determine when and whether to hold certain types of information in confidence. The application of that system and the substance of that system constitute Westinghouse policy and provide the rational basis required. Under that system, information is held in confidence if it falls in one or more of several types, the release of which might result in the loss of an existing or potential competitive advantage, as follows: (a) The information reveals the distinguishing aspects of a process (or component, structure, tool, method, etc.) where prevention of its use by any of L

4 CAW-17-4565 Westinghouse's competitors without license from Westinghouse constitutes a competitive economic advantage over other companies. (b) It consists of supporting data, including test data, relative to a process (or component, structure, tool, method, etc.), the application of which data secures a competitive economic advantage, e.g., by optimization or improved marketability. (c) Its use by a competitor would reduce his expenditure ofresources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing a similar product. ( d) It reveals cost or price information, production capacities, budget levels, or commercial strategies of Westinghouse, its customers or suppliers. (e) It reveals aspects of past, present, or future Westinghouse or customer funded development plans and programs of potential commercial value to Westinghouse. (f) It contains patentable ideas, for which patent protection may be desirable. (iii) There are sound policy reasons behind the Westinghouse system which include the following: (a) The use of such information by Westinghouse gives Westinghouse a competitive advantage over its competitors. It is, therefore, withheld from disclosure to protect the Westinghouse competitive position. (b) It is information that is marketable in many ways. The extent to which such information is available to competitors diminishes the Westinghouse ability to sell products and services involving the use of the information. (c) Use by our competitor would put Westinghouse at a competitive disadvantage by reducing his expenditure of resources at our expense.

5 CAW-17-4565 (d) Each component of proprietary information pertinent to a particular competitive advantage is potentially as valuable as the total competitive advantage. If competitors acquire components of proprietary information, any one component may be the key to the entire puzzle, thereby depriving Westinghouse of a competitive advantage. (e) Unrestricted disclosure would jeopardize the position of prominence of Westinghouse in the world market, and thereby give a market advantage to the competition of those countries. (f) The Westinghouse capacity to invest corporate assets in research and development depends upon the success in obtaining and maintaining a competitive advantage. (iv) The information is being transmitted to the Commission in confidence and, under the provisions of 10 CFR Section 2.390, is to be received in confidence by the Commission. (v) The information sought to be protected is not available in public sources or available information has not been previously employed in the same original manner or method to the best of our knowledge and belief. (vi) The proprietary information sought to be withheld in this submittal is that which is appropriately marked in "Vogtle Electric Generating Plant Supplemental Response to NRC Generic Letter 2004-02, Enclosure 2, 'Supplemental Response to NRC Generic Letter 2004-02' "(Proprietary), for submittal to the Commission, being transmitted by Letter GP-19572. The proprietary information as submitted by Westinghouse is that associated with resolution of and response to NRC Generic Letter 2004-02 and may be used only for that purpose. (a) This information is part of that which will enable Westinghouse to provide commercial support for resolution of and response to NRC Generic Letter 2004-02.

6 CAW-17-4565 (b) Further this information has substantial commercial value as follows: (i) Westinghouse plans to sell the use of similar information to its customers for the purpose of providing support for resolution of and response to NRC Generic Letter 2004-02. (ii) Westinghouse can sell support and defense of industry guidelines and acceptance criteria for plant-specific applications. (iii) The information requested to be withheld reveals the distinguishing aspects ofa methodology which was developed by Westinghouse. Public disclosure of this proprietary information is likely to cause substantial harm to the competitive position of Westinghouse because it would enhance the ability of competitors to provide similar technical evaluation justifications and licensing defense services for commercial power reactors without commensurate expenses. Also, public disclosure of the information would enable others to use the information to meet NRC requirements for licensing documentation without purchasing the right to use the information. The development of the technology described in part by the information is the result of applying the results of many years of experience in an intensive Westinghouse effort and the expenditure of a considerable sum of money. In order for competitors of Westinghouse to duplicate this information, similar technical programs would have to be performed and a significant manpower effort, having the requisite talent and experience, would have to be expended. Further the deponent sayeth not.

Proprietary Information Notice Transmitted herewith are proprietary and non-proprietary versions of a document, furnished to the NRC associated with resolution of and response to NRC Generic Letter 2004-02 and may be used only for that purpose. In order to conform to the requirements of 10 CFR 2.390 of the Commission's regulations concerning the protection of proprietary information so submitted to the NRC, the information which is proprietary in the proprietary versions is contained within brackets, and where the proprietary information has been deleted in the non-proprietary versions, only the brackets remain (the information that was contained within the brackets in the proprietary versions having been deleted). The justification for claiming the information so designated as proprietary is indicated in both versions by means oflower case letters (a) through (f) located as a superscript immediately following the brackets enclosing each item of information being identified as proprietary or in the margin opposite such information. These lower case letters refer to the types of information Westinghouse customarily holds in confidence identified in Sections (4)(ii)(a) through (4)(ii)(f) of the Affidavit accompanying this transmittal pursuant to 10 CFR 2.390(b)(l). Copyright Notice The reports transmitted herewith each bear a Westinghouse copyright notice. The NRC is permitted to make the number of copies of the information contained in these reports which are necessary for its internal use in connection with generic and plant-specific reviews and approvals as well as the issuance, denial, amendment, transfer, renewal, modification, suspension, revocation, or violation of a license, permit, order, or regulation subject to the requirements of 10 CFR 2.390 regarding restrictions on public disclosure to the extent such information has been identified as proprietary by Westinghouse, copyright protection notwithstanding. With respect to the non-proprietary versions of these reports, the NRC is permitted to make the number of copies beyond those necessary for its internal use which are necessary in order to have one copy available for public viewing in the appropriate docket files in the public document room in Washington, DC and in local public document rooms as may be required by NRC regulations if the number of copies submitted is insufficient for this purpose. Copies made by the NRC must include the copyright notice in all instances and the proprietary notice ifthe original was identified as proprietary. L_

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Vogtle Electric Generating Plant Supplemental Response to NRC Generic Letter 2004-02 Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Attachment 3 ALION Proprietary Information with Affidavit E5:A3-1

 .ALION SCIEN'C£ AND lECHNOLOGY ALION Science & Technology AFFIDAVIT We, Andy Roudenko, Project Manager and Martin Rozboril, Jr. Assistant Vice President Division Manager (AVPDM) state as follows:

(1) We, Andy Roudenko, Project Manager, and Martin Rozboril, Jr. AVPDM, Nuclear Services, ALION Science & Technology ("Alion") and have been delegated the function of reviewing the information described in paragraph (2) which is sought to be withheld, and have been authorized to apply for its withholding. (2) The infonnation sought to be withheld is contained in all revisions of ALION Science & Technology report "Erosion Testing of Small Pieces of Low Density Fiberglass Debris-Test Report," ALION-REP-ALION-1006-04, with the latest revision to date, Rev. 1, dated November 17, 20011. Information from this report was used to support analysis of post-LOCA debris transport in work designed to address GSI-191, Assessment of Debris Accumulation on PWR Sump Performance, issues at Southern Nuclear Operating Company, Vogtle Units 1 and 2. Specifically, the following Sections and Figures are to be withheld, on that basis that these unique attribute of the testing approach, test results and conclusions:

• Background
• Figure 1.1.1
• Figure 2.1.2
• Figure 2.1.3
• Figure 2.1.5
• Figure 2.1.6
• Figure 2.1.9
• Test Results, including Figures and Tables
• Data Analysis, including Figures and Tables
• Conclusions
• Appendices (3) In making this application for withholding of proprietary information of which it is the owner, Alion relies upon the exemption from disclosure set forth in the Freedom of Information Act ("FOIA"), 5 USC Sec. 552(b)(4), and the Trade Secrets Act, 18 USC Sec. 1905, and NRC regulations 10 CFR 9.17(a)(4), and 2.390(a)(4) for "trade secrets" (Exemption 4). The material for which exemption from disclosure is here sought also qualify under the narrower definition of "trade secret", within the meanings assigned to those terms for purposes of FOIA Exemption 4 in, respectively, Critical Mass Energy Project v. Nuclear Regulatory Commission, 975F2d87 l (DC Cir. 1992), and Public Citizen Health Research Group v. FDA, 704F2d1280 (DC Cir. 1983).

Page I of3 MAS Affidavit

AL I 0 N SCIENCE AtiD TECHNOLOGY (4) Some examples of categories of information which fit into the definition of proprietary information are:

a. Information that discloses a process, method, or apparatus, including supporting data and analyses, where prevention of its use by Alion's competitors without license from Alion constitutes a competitive economic advantage over other companies

b. Information which, if used by a competitor, would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing of a similar product;

c. Information which reveals aspects of past, present, or future Alion customer-funded development plans and programs, resulting in potential products to Alion;

d. Information which discloses patentable subject matter for which it may be desirable to obtain patent protection.

The information sought to be withheld is considered to be proprietary for the reasons set forth in paragraphs (4) a, and (4) b, above. (5) To address 10 CFR 2.390 (b) (4), the information sought to be withheld is being submitted to NRC in confidence. The information is of a sort customarily held in confidence by Alion, and is in fact so held. The information sought to be withheld has, to the best of my knowledge and belief, consistently been held in confidence by Alion, no public disclosure has been made, and it is not available in public sources. All disclosures to third parties including any required transmittals to NRC, have been made, or must be made, pursuant to regulatory provisions or proprietary agreements which provide for maintenance of the information in confidence. Its initial designation as proprietary information, and the subsequent steps taken to prevent its unauthorized disclosure, are as set forth in paragraphs (6) and (7) following. (6) Initial approval of proprietary treatment of a document is made by the manager of the originating component, the person most likely to be acquainted with the value and sensitivity of the information in relation to industry knowledge. Access to such documents within Alion is limited on a "need to know" basis. (7) The procedure for approval of external release of such a document typically requires review by the staff manager, project manager, principal scientist or other equivalent authority, by the manager of the cognizant marketing function (or their delegate), and by the Legal Operation, for technical content, competitive effect, and determination of the accuracy of the proprietary designation. Disclosures outside Alion are limited to regulatory bodies, customers, and potential customers, and their agents, suppliers, and licensees, and others with a legitimate need for the information, and then only in accordance with appropriate regulatory provisions or proprietary agreements. (8) The document identified in paragraph (2), above, is classified as proprietary because it contains "know-how" and "unique data" developed by Alion within our research and Page 2 of3 MAS Affidavit

AL I 0 N SCIENCE. AND TECHNOl.OGY development programs. The development of this document, supporting methods and data constitutes a major Al ion asset in this current market. (9) Public disclosure of the information sought to be withheld is likely to cause substantial harm to Al ion's competitive position and foreclose or reduce the availability of profit-making opportunities. The information is part of Alion's comprehensive BWR/PWR GSI-191 analysis base, and its commercial value extends beyond the original development cost. The value of the technology base goes beyond the extensive physical database and experimental methodology and includes development of the expertise to determine and apply the appropriate evaluation process. The research, development, engineering, analytical and experimental costs comprise a substantial investment of time and money by Alion. The precise value of the expertise to devise an evaluation process and apply the correct analytical methodology is difficult to quantify, but it clearly is substantial. Alion's competitive advantage will be lost if its competitors are able to use the results of the Alion experience to normalize or verify their own process or if they are able to claim an equivalent understanding by demonstrating that they can arrive at the same or similar conclusions. The value of this information to Alion would be lost if the information were disclosed to the public. Making such information available to competitors without their having been required to undertake a similar expenditure of resources would unfairly provide competitors with a windfall, and deprive Alion of the opportunity to exercise its competitive advantage to seek an adequate return on its large investment in developing these very valuable analytical tools. I declare under penalty of perjury that the foregoing affidavit and the matters stated therein are true and correct to the best of my knowledge, information, and belief. Executed on this 20th day of April 2017. Martin rif Digitally signed by Andy Roudenko Rozboril, Jr.

               ~I DN: cn=Andy Roudenko, o=Alion A- r K II t,~ Scie~ce an.d_T.echnology, ou=Nuclear f ~Services D1v1s1on,                                                       2017.04.20
                            ."       e1fuail=aroudenko@a1ionscience.com,
                          ;' r   _,.7=uS -'-:~                                                             15:06:17 -06'00'
                    .~ j?            Date: 2017.04.20 12:09:50 -07'00' Andy Roudenko                                                                     Martin Rozboril, Jr.

Project Manager Assistant Vice President ALION Science & Technology Division Manager, Nuclear Services ALION Science & Technology Page 3 of3 MAS Affidavit

Vogtle Electric Generating Plant - Units 1 & 2 Supplemental Response to NRC Generic Letter 2004-02 Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary)

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Table of Contents 1.0 Overall Compliance 2.0 General Description of and Schedule for Corrective Actions 3.0 Specific Information Regarding Methodology for Demonstrating Compliance 3.a Break Selection 3.b Debris Generation/Zone of Influence 3.c Debris Characteristics 3.d Latent Debris 3.e Debris Transport 3.f Head Loss and Vortexing 3.g Net Positive Suction Head 3.h Coating Evaluation 3.i Debris Source Term 3.j Screen Modification Package 3.k Sump Structural Analysis 3.1 Upstream Effects 3.m Downstream Effects - Components and System 3.n Downstream Effects - Fuel and Vessel 3.o Chemical Effects 3.p Licensing Basis 4.0 NRC Requests for Additional Information 5.0 References Attachments ES:A 1 General Electric Hitachi (GEH) Proprietary Information Affidavit E5:A2 Westinghouse Electric Corporation (WEC) Proprietary Information Affidavit E5:A3 ALION Proprietary Information Affidavit E5-1

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) 1.0 Overall Compliance: Provide information requested in GL 2004-02 Requested Information Item 2(a) regarding compliance with regulations. GL2004-02 Requested Information Item 2(a) Confirmation that the ECCS and CSS recirculation functions under debris loading conditions are or will be in compliance with regulatory requirements listed in the Applicable Regulatory Requirements section of this GL. This submittal should address the configuration of the plant that will exist once all modifications required for regulatory compliance have been made and this licensing basis has been updated to reflect the results of the analysis described above. Response to 1.0: This submittal proposes a risk-informed methodology for determining the design requirements to address the effects of loss-of-coolant accident (LOCA)-generated debris on emergency core cooling system (ECCS) and containment spray system (CSS) recirculation functions. The risk-informed analysis covers a full spectrum of postulated LOCAs, including double-ended guillotine breaks (DEGBs), for all pipe sizes up to and including the design-basis accident (OBA) LOCA, to provide assurance that the most severe postulated LOCAs are evaluated. Vogtle Electric Generating Plant (VEGP) conservatively relegates to failure the individual breaks that can generate and transport debris that are not bounded by VEGP analyzed limits. The results of the evaluation in Enclosure 3 show that the risk from the failures related to LOCA-generated debris is "very small" as the risk falls in Region Ill of RG 1.174. The methodology includes conservatisms in the plant-specific testing and in the assumption that all breaks that exceed the tested debris quantities are relegated to failure. Conservatisms in the VEGP approach are discussed in Enclosure 4, Defense-in-Depth and Safety Margin. Southern Nuclear Operating Company (SNC) is utilizing a risk-informed approach to the effects of LOCA debris for VEGP. The risk-informed approach replaces the existing deterministic approach described in the VEGP licensing basis and consequently requires an amendment to the VEGP Units 1 and 2 operating licenses to incorporate the revised methodology per the requirements of Title 10 of the Code of Federal Regulations (CFR) Section 50.59 (10 CFR 50.59). The proposed amendment to the operating license will be described in the future license amendment request (LAR). Exemptions to the overall requirements associated with 10 CFR 50.46(a)(1 ), GDC 35, GDC 38, and GDC 41 are required due to the change in methodology. The requests for exemption will be provided in the future LAR. In addition, SNC proposes to amend the VEGP Unit 1 and Unit 2 operating licenses to revise the Technical Specifications (TSs) for the ECCS and CSS. The proposed TS changes detailed in the future LAR will align the TSs with the risk-informed E5-2

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) methodology change. The licensing discussion is continued in the Response to 3.p of this enclosure. In the resolution of Generic Safety Issue (GSl)-191, "Assessment of Debris Accumulation on Pressurized Water Reactor (PWR) Sump Performance," VEGP implemented (or will implement) the following changes:

• The refueling water storage tank (RWST) High level was increased and the Low-Low level (initiation of semi-automatic switchover to recirculation) decreased to provide increased submergence of the sump strainers while maintaining adequate net positive suction head (NPSH) for the ECCS and containment spray (CS) pumps; allowing sufficient time for completion of operator actions for switchover to recirculation.
• To improve existing margins until all corrective actions can be implemented, VEGP installed larger sump strainers that increased the available screen area for each of the residual heat removal (RHR) strainers and the CS strainers. The hole diameters of the strainer perforated plates were reduced to lessen the potential for debris passing through the strainer and causing plugging and/or wear of the downstream ECCS and CS piping and equipment, and reactor vessel. In addition, Min-K insulation was removed from the containment bioshield area.
• Orifices were installed in the intermediate- and high-head ECCS lines, and the associated throttle valves were adjusted to operate at a minimum internal clearance greater than the size of debris that could pass through the strainers.

The opening size was increased to ensure that adequate clearance in the valves will prevent debris from causing excessive wear or plugging.

• Procedural and program controls are in place to ensure materials used in the containments will not result in an increase of the debris loading beyond the analyzed values. This includes controls for containment coatings, labels, and insulation.
• Extensive analysis has been performed in accordance with Nuclear Energy Institute (NEI) 04-07 guidance (Reference 2), the associated United States Nuclear Regulatory Commission (NRC) safety evaluation (SE) (Reference 3),

and other industry documents reviewed by the NRC. With few exceptions, VEGP has followed this guidance. Technical justification is available and provided for the few cases where other approaches were utilized.

• The emergency operating procedures are being revised to delay operator action to isolate the RHR pumps from the RWST. This ensures that water level in the RWST is drawn down to the Empty alarm level for all scenarios (it previously only reached the Empty alarm level for scenarios that actuated CS) and prevents most scenarios from resulting in partially submerged strainers.

E5-3

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary)

• To ensure full submergence for an increased number of postulated break scenarios, the Uhit 1 and Unit 2 RHR strainers require a reduction in height. The Unit 1 and Unit 2 design packages have been prepared.

Correspondence Background The following discussion contains correspondences issued by or submitted to the NRG prior to December 31, 2007, on the subject of GSl-191. The title of each letter is provided in the reference section of this enclosure. The NRG issued Bulletin 2003-01 on June 9, 2003 (Reference 5), asking for a 60-day response providing a description of any interim compensatory measures that have been implemented, or that will be implemented, to reduce the risk which may be associated with potentially degraded or nonconforming EGGS and GSS recirculation functions until an evaluation to determine compliance is complete. SNG provided the 60-day response in a letter dated August 7, 2003 (Reference 6). Supplemental letters dated October 29, 2004 (Reference 7), and July 22, 2005 (Reference 8), were provided by SNG in response to requests for additional information. The NRG issued Generic Letter (GL) 2004-02 on September 13, 2004 (Reference 1), requesting an initial 90-day response, a 12-month response, and for the guidance of the GL to be met by December 31, 2007. In December 2004, NEI issued NEI 04-07 (Reference 2) providing an evaluation methodology for the industry. The NRG provided the associated SE (Reference 3) on December 6, 2004. The NRG had already issued RG 1.82 Revision 3 (Reference 25) in November 2003. SNG provided the initial response for VEGP in a letter dated February 25, 2005 (Reference 10). SNG provided a follow-up response on August 31, 2005 (Reference 11 ), providing more details on how SNG would meet the GL 2004-02 requirements. The NRG issued a request for additional information on February 9, 2006 (Reference 12), with a 60-day response time. NEI worked with the NRG and recognized that much of the information needed to address the RAls would not be available until ongoing testing activities were completed. The NRG-issued letter dated March 28, 2006 (Reference 13), identified that the Request for Additional Information (RAI) answers could be provided as part of the supplemental response by the end of December 2007. An NRG letter dated January 4, 2007 (Reference 18), provided clarification that even if a licensee had an extension for modifications past 2007, the supplemental response was still due by December 31, 2007.

• SNG submitted an extension request in a letter dated June 22, 2006 (Reference 14),

for modification/installation of the Unit 1 EGGS flow orifices and for chemical effects testing. In a teleconference on June 30, 2006, with the NRG staff reviewer of the E5-4

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) June 22, 2006, extension request, SNC was asked to provide an update of on-going activities and a clarification as to what activities are driving the extension request. SNC provided the requested information in a response dated July 28, 2006 (Reference 15), which also requested an extension from December 31, 2007, to the spring 2008 outage. This extension request was approved in an NRC letter dated September 7, 2006 (Reference 16).

• The NRC issued a letter dated August 15, 2007, containing the content guide for the GL 2004-02 supplemental response due in December 2007. Additional information was provided by the NRC in a letter dated September 27, 2007, for chemical effects, protective coatings, and head loss testing. A revision to the content guide was issued by the NRC in a letter dated November 21, 2007 (Reference 108). The due date for the supplemental response was extended by an NRC letter dated November 30, 2007 to allow the supplemental response to be submitted by February 29, 2008.

An NRC letter dated November 8, 2007, provided guidance for requesting plant-specific extensions. Additional information was also provided in an NRC letter dated November 13, 2007, on how GSl-191 would be closed and how the closure would be documented for each site. SNC submitted a letter dated December 7, 2007 (Reference 19), requesting an extension for submittal of chemical effects testing results, downstream effects - components and systems, and downstream effects - fuel and vessel until June 30, 2008. This request was approved in an NRC letter dated December 19, 2007 (Reference 20). E5-5

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) 2.0 General Description of and Schedule for Correction Actions: Provide a general description of actions taken or planned, and dates for each. For actions planned beyond December 31, 2007, reference approved extension requests or explain how regulatory requirements will be met as per Requested Information Item 2(b). (Note: All requests for extension should be submitted to the NRC as soon as the need becomes clear, preferably no later than October 1, 2007.) GL 2004-02 Requested Information Item 2(b) A general description and implementation schedule for all corrective actions, including any plant modifications that you identify while responding to this generic letter. Efforts to implement the identified actions should be initiated no later than the first refueling outage starting after April 1, 2006. All actions should be completed by December 31, 2007. Provide justification for not implementing the identified actions during the first refueling outage starting after April 1, 2006. If corrective actions will not be completed by December 31, 2007, describe how the regulatory requirements discussed in the Applicable Regulatory Requirements section will be met until the corrective actions are completed. Response to 2.0: SNC has performed analysis to determine the susceptibility of the ECCS and CSS recirculation functions for VEGP to the adverse effects of post-accident debris blockage and operation with debris-laden fluids. These analyses conform, to the greatest extent practicable, to the NEI 04-07 methodology (Reference 2) as approved by the NRC SE dated December 6, 2004 (Reference 3). As of April 24, 2017, SNC has completed the following GL 2004-02 (Reference 1) actions, analyses, and modifications:

• Replaced Unit 1 and Unit 2 containment emergency sump screens during refueling outage 1R13 (Fall 2006), and refueling outage 2R12 (Spring 2007),

respectively

• Installed ECCS flow orifices in the intermediate and high-head ECCS lines that allow the ECCS throttle valves to be opened greater than the maximum expected strainer bypass debris size while maintaining the capability to ensure ECCS flow balance, mitigating downstream effects (2008)
• Completed inspection of containment per NEI 02-01 (Reference 41 ), "Condition Assessment Guidelines: Debris Sources Inside PWR Containment"
• Performed latent debris sampling and characterization ,
• Participated in the PWR Owners Group (PWROG) program to evaluate downstream effects related to in-vessel long-term cooling with results documented in WCAP-16793-NP-A (Reference 22)
• Removed Min-K insulation in the original zone of influence (ZOI) analyzed for GL 2004-02 from VEGP's containments based on preliminary head loss testing (without chemical effects) which determined that the removal of Min-K insulation resulted in a significant reduction in head loss across a debris-laden strainer E5-6

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary)

• Implemented programmatic and procedural changes to maintain acceptable configuration and to protect the newly established design and licensing basis
• Developed containment 30 CAD model of VEGP Unit 1 containment to include pipe welds, for both VEGP containments because Units 1 and 2 are virtually identical (CAD model of VEGP Unit 1 containment was used to determine a correlation between the containment pool volume and containment pool level)
• Completed detailed laser scans of the VEGP containments, which provide measurements for contingency insulation replacement for Units 1 and 2 (laser scans of both units were completed before January 1, 2013)
• Developed detailed debris generation and debris transport analyses and a computational fluid dynamics (CFO) model
• Developed a hydraulic model of the ECCS
• Performed detailed CS and RHR NPSH analysis
• Performed water level analysis
• Modified probabilistic risk assessment (PRA) to include strainer and core blockage events
• Quantified chemical precipitants using WCAP-16530 (with refinements)
• Performed chemical effects testing
• Completed RELAPS-30/MELCOR modeling similar to South Texas Project (STP) model (however, the results are not used as input for the base case analysis)
• Performed strainer head loss and fiber debris penetration testing
• Participated in the PWROG Comprehensive Analysis and Test Program for GSl-191 Closure
• Performed downstream wear and blockage analysis to WCAP-16406-P-A, Revision 1 (Reference 21)
• Performed detailed structural analysis of strainers
• Assembled base case final inputs for quantifying the conditional failure probabilities related to GSl-191 using the software package Nuclear Accident Risk-Weighted Analysis (NARWHAL) (see Enclosure 3, Section 13.1 for general description of the software).
• Completed NARWHAL sensitivity analyses
• Integrated NARWHAL results into VEGP PRA model to determine changes in core damage frequency (LiCDF) and changes in large early release frequency (LiLERF)
• Revised operating procedures to ensure that the RHR strainers are completely submerged for an increased number of postulated LOCA scenarios (operator action to isolate the RWST from the RHR pumps was delayed until the Empty level is reached to ensure sufficient injection of RWST water for breaks that activate CS)

The following risk-informed resolution path activities are planned by SNC to address GL 2004-02 and support closure of GSl-191 for VEGP.

• SNC is planning to modify the VEGP Unit 1 and Unit 2 RHR sump strainers to reduce the overall height by removing the top two strainer disks per stack from each of the RHR strainer assemblies.

E5-7

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary)

• SNC will submit an LAR for a risk-informed resolution to GL 2004-02 for VEGP within six months after receipt of the SE for WCAP-17788-P.
• SNC will submit any necessary revisions to the supplemental response to support closure of GL 2004-02 for VEGP Units 1 and 2 within six months after receipt of the SE for WCAP-17788-P.

Correspondence Background The following discussion contains correspondences issued by or submitted to the NRC beyond December 31, 2007, on the subject of GSl-191. The correspondences document VEGP's compliance with regulatory requirements per Requested Information Item 2(b) and include reference to approved extension requests. The title of each letter is provided in the reference section of this enclosure. SNC letter dated February 28, 2008, NL-07-1777 (Reference 95), provided SNC's supplemental response to GL 2004-02 for VEGP. SNC letter dated May 21, 2008, NL-08-0670 (Reference 96), provided a revised transmittal of SNC's supplemental response to GL 2004-02 for VEGP based on NRC's questions regarding the proprietary nature of information provided in SNC letter dated February 28, 2008 (Reference 95). SNC letter dated May 22, 2008, NL-08-0818 (Reference 97), requested an extension for the final response to GL 2004-02 for the completion of WCAP-16406-P-A and WCAP-16793-NP-A evaluations, and chemical effects testing and evaluation of test results. An extension was granted to SNC by the NRC to July 31, 2008, in a letter dated May 29, 2008, as stated in NL-08-1155 (Reference 98). SNC letters dated July 31, 2008, NL-08-1155 (Reference 98) and NL-08-1195 (Reference 99), provided the downstream effects results for components and in-vessel analyses and requested an extension for the GL 2004-02 supplemental response for chemical effects, respectively. SNC letter dated August 22, 2008, NL-08-1228 (Reference 100), provided the GL 2004-02 response for chemical effects. The letter also contained a revised answer to question 3.g.15 originally submitted in SNC letter dated May 21, 2008 (Reference 96). NRC letter dated September 17, 2008, NL-08-1497 (Reference 101), provided RAls from a partial review of prior SNC responses to GL 2004-02 pertaining to the reliance on results from testing at the VUEZ facility by Alien Science and Technology. As discussed with Mr. Jared S. Wermiel, Deputy Director of the Division for Engineering and Safety Systems, in a telephone call with SNC on September 11, 2008, the NRC identified several critical issues with the test protocol used in the testing at VUEZ. The NRC staff has stated that based on their review of information provided by Alien on the VUEZ testing, it is highly unlikely that SNC's ES-8

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) reliance on the VUEZ testing, performed to date to demonstrate strainer adequacy, will provide an adequate technical basis to resolve GL 2004-02. SNC letter dated November 7, 2008, NL-08-1583 (Reference 102), requested an extension, in accordance with SECY-06-0078, for completion of chemical effects testing and closeout activities for GL 2004-02 as a result of NRC's concern that SNC's reliance on the VUEZ performed testing would not provide an adequate technical basis to resolve GL 2004-02 for VEGP. NL-08-1583 also noted that the RAls issued by the staff on September 17, 2009 were from a partial review of SNC's responses to GL 2004-02 and did not represent a comprehensive set of RAls. In addition, SNC submitted milestone dates supporting a closeout of GSl-191 and a final response to the staff by November 20, 2009, predicated on a reasonable submittal of SNC test-for-success protocol with NRG review and comment cycle, and resolution of any issues associated with pending revision ofWCAP-16793-NP. NRG letter dated December 2, 2008, NL-08-1829 (Reference 87), provided RAls for prior SNC supplemental responses, letters dated February 28, 2008; May 21, 2008; July 31, 2008; and August 22, 2008 (References 95, 96, 98, and 100, respectively). The NRG requested RAI responses within 90 days. SNC letter dated February 10, 2009, NL-09-0159 (Reference 103), notified the NRG that a single response to the RAls issued by the NRG letter dated December 2, 2008, would be submitted once the new testing and analysis discussed in the November 7, 2008, letter was completed. SNC also notified the NRG that the planned completion date for the GL 2004-02 response was November 20, 2009, excluding issues related to WCAP-16793-NP. SNC letter dated November 19, 2009, NL-09-1839 (Reference 104), stated that the schedule for completion of GSl-191 activities for VEGP is contingent upon resolution of generic issues and their impact to the remaining 26 RAls for VEGP. Of the 29 RAls issued December 2, 2008, SNC discussed 26 satisfactorily with the staff during public telecoms between SNC and the staff on August 13, 2009, and October 13, 2009. The three remaining RAls concerned the following generic issues: Nukon ZOI, fibrous insulation erosion, and in-vessel downstream effects. During the telecom between SNC and the staff on October 13, 2009, the above generic issues were discussed as to how the outcome of said issues would impact the 26 RAls for VEGP. It was agreed to by the staff in this telecom that a written response to the resolved RAls was not required by November 20, 2009. NEI letter to NRG dated May 4, 2012 (Reference 91 ), highlighted the current industry status and recommended actions for closure of GSl-191 based on licensees providing a docketed submittal to the NRG by December 31, 2012, outlining a GSl-191 resolution path and schedule pursuant to the Commission direction in Staff Requirements Memorandum (SRM) SECY-10-0113 (Reference 90). E5-9

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) An NEI letter to the NRC dated November 15, 2012 (Reference 92), and a subsequent NRC review of the schedule (Reference 93) recommended that licensees delay submittal of GSl-191 resolution path and schedule until January 31, 2013. The letter also recommended an alternative option to submit 30 days following placement of both the Commission's response to SECY-12-0093 (Reference 70) and the NRC SE on WCAP-16793-NP (Reference 94) into the public record. The Commission approved the staff's recommendation in SRM-SECY-12-0093 (Reference 70) dated December 14, 2012, to allow licensees the flexibility to choose any of the three options discussed in the paper to resolve GSl-191. Further, the Commission encouraged the staff to remain open to staggering licensee submittals and the associated NRC reviews to accommodate the availability of staff and licensee resources. The SE forWCAP-16793-NP (Reference 94) was made publicly available by the NRC on April 16, 2013. SNC Letter dated May 16, 2013, NL-13-0953 (Reference 105), transmitted the VEGP Units 1 and 2 resolution path forward and schedule for resolution, summary of actions completed for GL 2004-02, and defense-in-depth and mitigation measures, using the industry template developed by NEI. VEGP provided a basis for continued operation in the interim while the industry and NRC collaborated on how to proceed towards resolution of the issue. VEGP is following the "STP Piloted Risk-Informed Approach for GSl-191," as submitted by the following South Texas Project Nuclear Operating Company (STPNOC) letters to the NRC (Reference 105). STPNOC letter to the NRC dated November 13, 2013, NOC-AE-13003043 (Reference 44), submitted Supplement 1 to the STP pilot risk-informed approach for GSl-191. STPNOC letter to the NRC dated August 20, 2015, NOC-AE-15003241 (Reference 45), submitted Supplement 2 to the STP pilot submittal. STPNOC letter to the NRC dated October 20, 2016, NOC-AE-16003401 (Reference 109), submitted Supplement 3 to the Revised STP pilot submittal. E5-10

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) 3.0 Specific Information Regarding Methodology for Demonstrating Compliance:

a. Break Selection The objective of the break selection process is to identify the break size and location that present the greatest challenge to post-accident sump performance.

1. Describe and provide the basis for the break selection criteria used in the evaluation.

Response to 3.a.1: The VEGP debris generation calculation followed the methodology of NEI 04-07 and associated NRG SE (References 2 and 3, respectively), with the exception that it analyzed a full range of breaks, not just the worst-case breaks as suggested by NEI 04-07. The purpose of the calculation is to obtain debris quantities for the range of possible break scenarios. The calculation evaluated debris generation quantities for breaks on every inservice inspection (ISi) weld within the Class 1 pressure boundary. Both DEGBs and partial breaks were considered. All break sizes analyzed are assumed to fall into one of three high-level categories: small-break LOCA (SBLOCA) - a break smaller than 2 inches, medium-break LOCA (MBLOCA)

   - a break greater than or equal to 2 inches and less than 6 inches, or large-break LOCA (LBLOCA) - a break greater than or equal to 6 inches with the largest break being a DEGB of the 31-inch crossover leg.

In the debris generation calculation, a three-dimensional CAD model of VEGP Unit 1 containment building was updated to work with ENERCON's Break Analysis Debris Generator (BADGER) software. Note that the Unit 1 containment was used to represent both containments because the VEGP units are virtually identical. BADGER was used to place ZOls representing possible breaks on every ISi weld identified in containment. See Figure 3.a.1-1 for weld locations and Table 3-8 in Enclosure 3 for a complete list of welds inside the first isolation valve. In Figure 3.a.1-1, welds labeled as "In" are on the RCS side of the first isolation valve and welds labeled as "Out" are downstream of the first isolation valve. As discussed in Enclosure 1 Section 3.2, non-pipe LOCAs were not explicitly analyzed. E5-11

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Figure 3.a.1 Weld Locations where Postulated LOCAs Occur It should be noted that, while DEGBs on main loop piping are typically bounding with regard to the volume of debris generated, small partial breaks are much more likely to occur. A partial break is any break smaller than a DEGB (i .e., sidewall break) . Partial breaks are described by equivalent break size and are represented by a hemispherical ZOI with a radius proportional to the equivalent break size (see Figure 3.a.1-2). E5-12

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Figure 3.a.1 Single Partial Break Zone of Influence

2. State whether secondary line breaks were considered in the evaluation (e.g. , main steam and feedwater lines) and briefly explain why or why not.

Response to 3.a.2: Although the probability is low, a secondary side break (SSB) inside containment could require ECCS recirculation in a feed and bleed scenario. Therefore , secondary side breaks from the steam generator feedwater lines and main steam lines were analyzed . Because secondary side breaks occur at a lower pressure and temperature than the primary side breaks , the ZOI size corresponding to the insulation destruction pressure would be smaller. The appropriate ZOI sizes were calculated based on the ANSI jet methodology described in Appendix I of NEI 04-07 Volume 2 (Reference 3) . The main steam and feedwater pressures, temperatures , and calculated ZOI sizes are presented in Table 3.a .2-1. Breaks were postulated in increments of least every 5 ft along each of the main steam and feedwater pipes. All secondary side breaks were assumed to be DEGBs. Only Nukon insulation was considered for the secondary-side breaks because there is no fire barrier within the vicinity of the main steam and feedwater lines , and the coatings quantities would be bounded by the primary side breaks. E5-13

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Table 3.a.2 Secondary-Side Line ZOI Summary Main Steam Lines Feedwater Lines Pipe Inner Pipe Inner Po (psia) To Diameter Po (psia) To Diameter (oF) (oF) (inch) (inch-) 985 545 24.0 1150 445 12.8 Mass Flux (lbm*ft2 *s*1 ) CT Mass Flux (lbm*ft*2 *s*1 ) CT 1,978 1.25 19,238 1.65 Insulation I Insulation I Destruction Pressure ZOI Radius (D) Destruction Pressure ZOI Radius (D) (psia) (psia) CoatinQs 140 3.0 Coatings I 40 2.8 lnteram / 10.2 7.9 lnteram I 10.2 7.2 Nukon I 6 10.5 Nukon / 6 11.3 Note: Cr is the thrust coefficient

3. Discuss the basis for reaching the conclusion that the break size(s) and locations chosen present the greatest challenge to post-accident sump performance.

Response to 3.a.3: Debris generation quantities were evaluated for breaks on every ISi weld within the Class 1 pressure boundary. Both DEGBs and partial breaks were considered. The welds are sufficiently close, with sufficient overlap in the ZOls to provide confidence that the debris load that' presents the greatest challenge to post-accident sump performance has been captured. The total quantity of each type of debris generated within a particular ZOI is unique for every break scenario. Therefore, the bounding break-specific debris loads contained in the BADGER database were used on a break-specific basis for the analysis. The results of the debris generation calculation are presented below .. When reading the tables in this section it should be noted that the individual quantities for fines, small pieces, large pieces, and intact blankets do not necessarily add up to the total fiber quantity within the ZOI because the minimum, maximum, and average values for each size do not necessarily come from the same break. All average values are based on an equal probability of all breaks and do not consider differences in weld-specific break frequencies or the lower frequencies associated with larger break sizes. These results reflect the following conservatisms:

• All debris sources within the reactor cavity were assumed to be available for destruction by all breaks within the reactor cavity, despite the likelihood that ZOls would be restricted by structures and restraints.

E5-14

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary)

• All qualified coatings on steel and concrete were analyzed as having the worst-case coating system for each surface.
• Main loop breaks in the steam generator (SG) compartments were grouped by loop and truncated collectively in a way that could result in conservative amounts of debris generated for some breaks.

Debris Generated by DEGBs Table 3.a.3-1 shows the location of the worst-case break for each debris type. Table 3.a.3-2, Table 3.a.3-3, and Figures 3.a.3-1 through 3.a.3-8 show the minimum, average, and maximum debris quantities by debris type for DEGBs and partial breaks upstream of the first isolation valve at VEGP Units 1 and 2. Table 3.a.3 Location of Maximum Debris by Debris Type Debris Type Worst-Case Location Amount 3 Nukon (ft ) SG Compartment 1/4 Side 2229.2 lnteram E-50 Series (lbm) SG Compartment 1/4 Side 59.8* IOZ (lbm) SG Compartment 2/3 Side 65.3 Epoxy (lbm) Reactor Cavity 220.4

     *The limiting quantity of lnteram E-50 is generated by a partial break. This is possible because partial breaks are analyzed as being centered on the edge of the pipe, whereas DEGBs are centered on the axis of the pipe (see Figure 3.a.1-2). Because of this, partial breaks can extend further than DEGBs up to the outside radius of the pipe.

E5-15

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Table 3.a.3 Debris Generated by DEGBs Upstream of First Isolation Valve Debris Quantity Generated Small Breaks Medium Breaks Large Breaks Debris Type Debris Size ( < 2" ( 2"- 6") ( > 6" Min Avg Max Min Avg Max Min Avg Max Fines (Individual 0.0 0.1 0.8 0.0 2.9 12.2 1.9 56.0 289.3 Fibers) Small Pieces 0.0 0.4 2.8 0.0 9.4 40.8 6.4 187.1 999.5 (< 6" a side) Large Nukon (ff) Pieces 0.0 0.3 1.5 0.0 6.6 25.6 2.8 106.1 549.6 (> 6" a side) Intact (Covered) 0.0 0.3 1.6 0.0 7.1 27.7 3.1 114.6 594.0 Blankets All Debris 0.0 1.2 6.6 0.1 26.1 104.1 14.6 464.0 2229.2 Within ZOI Fire Barrier Fiber 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.4 12.1 Debris (lbm) 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.8 28.3 IOZ Qualified 0.0 <0.1 0.1 0.0 <0.1 1.3 0.0 9.9 65.3 Coatings (lbm) Particulate Epoxy Qualified 0.0 <0.1 0.3 0.0 0.4 4.9 0.0 44.8 220.4 Coatings (lbm) E5-16

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Range of Debris with Average by Pipe Size, DEGB 10" TO 16" (116 WELDS) 15 165 649 1 -- - - - - - - '

    ~

Vi Cll 6" T08"(198WELDS) 5 46 127

   .~

c.. 2.S" T04" (158WELDS) Q 9 64

               < 2" (354 WELDS)  Q    1     7 Debris Generated (ft 3 )

Figure 3.a.3 Range of Nukon Debris Generated by DEGBs Upstream of First Isolation Valve E5-17

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Range of Debris with Average by Pipe Size, DEGB MAIN LOOP (56 WELDS) 0 4 1--~--~~~~~~~~~~~~~~~~----~..._ ___ 40 10" TO 16" ( 116 WELDS) IJ 0

!l iii CIJ 6" TO 8" (198 WELDS) 0 0 0 Q.

ii: 2.5" TO 4" (158 WELDS) 0 0 0

              < 2" (354 WELDS)   0      0 0 Debris Generated (lb)

Figure 3.a.3 Range of Fire Barrier Debris Generated by DEGBs Upstream of First Isolation Valve Note that Figure 3.a.3-2 shows the total amount of fi re barrier destroyed (fiber plus particulate) . E5-18

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Range of Debris with Average by Pipe Size, DEGB MAINLOOP(S6WELDS) 37 143 220 r--~--__;;...-~~~~~~~~~~.--~~~~~:------ 10" TO 16" (116 WELDS) 0 3 21

   .~

VI QJ 6"T08" (198WELDS) Q 1 6 c.. ii: 2.S" T04 " (158WELDS) Q 0 1

               < 2" (354 WELDS)  Q     0    0 Debris Generated (lb)

Figure 3.a.3 Range of Epoxy Coatings Debris Generated by DEGBs Upstream of First Isolation Valve ES-19

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Range of Debris with Average by Pipe Size, DEGB

    ~

iii QI 6" TO 8" (198 WELDS) 0 0 2 a. ii: 2.S" T04 " (158WELDS) Q 0 1

             < 2" (3S4 WELDS) 0  0  0 Debris Generated (lb)

Figure 3.a.3 Range of IOZ Coatings Debris Generated by DEGBs Upstream of First Isolation Valve E5-20

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Table 3.a.3 Debris Generated by Partial Breaks Upstream of First Isolation Valve Debris Quantity Generated Small Breaks Medium Breaks Large Breaks Debris Type Debris Size ( < 2" ( 2" - 6") ( > 6") Min Avg Max Min Avg Max Min Avg Max Fines (Individual 0.0 <0.1 0.2 0.0 1.0 8.9 0.0 38.4 223.7 Fibers) Small Pieces 0.0 <0.1 0.8 0.0 3.0 28.5 0.2 128.6 794.1 (< 6" a side) Large Pieces Nukon (ft3) 0.0 <0.1 0.3 0.0 2.3 24.4 0.0 72 .0 438.4 (> 6" a side) Intact (Covered) 0.0 <0.1 0.3 0.0 2.5 26.4 0.0 77 .8 473.7 Blankets All Debris 0.0 0.1 1.6 0.0 8.8 79.5 0.2 317.0 1783.0 Within ZOI lnteram E-50 Fiber 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.2 17.9 Series (lbm) 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.4 41 .9 IOZ Qualified 0.0 <0.1 0.1 0.0 <0.1 2.1 0.0 7.4 40.9 Coatings (lbm) Particulate Epoxy Qualified 0.0 <0.1 0.4 0.0 0.2 3.9 0.0 30.7 149.1 Coatings (lbm) E5-21

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Range of Debris with Average by Break Size 31* ~6 74_ 161 30.5"

                -                                        --                                           154 30*
                -                                     _9-7                                          1492 29.s*
                -                                   67_                                          144 29*
                -                               6.18                                                              m1:783 28.5"
                -                                                                                                -7 28"   -                              8_                               I

_68 21.s*

                 -                      4'                                                                  _64 27*
                -         r

_489 1594 2&*

                 -        I                                                     l 25"
                 -        I                                                     I

_404 24* 86 131 QI N 23" .., f)4: n1 v; 22* Ir -128 I

   .¥     21*                 _:t_                                        '10l3 Ill          -

QI

    ~

20"

                 -                                                   9-14 cc     19*                                                 -2 ,..

1s* 5' 55 17" 5'

16. 613 ~

15" 14" _] 12* 10*

                -           7

_;~1 8' 6' 4' < 10 5 2* .-0 2 21 0.5"" 0 0 2 Debris Generated (ft3 ) Figure 3.a.3 Range of Nukon Debris Generated by Partial Breaks Upstream of First Isolation Valve E5-22

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Range of Debris with Average by Break Size 31" .. 30.5" **3 30" ~ I 29.5" 29" ..... 28.5" I .... I I I

!l 28" I

                                                           *9 v;
  ..:.:: 27.5" I        I                                      60 !

nl cc

    ....C1I   27" I        I                                    ~

26" n I I

• u 25" I I ..,

24" ~o .., I I 34 23* I I 32 22* I I 22 21* 00 1 I I Debris Generated (lb) Figure 3.a.3 Range of Fire Barrier Debris Generated by Partial Breaks Upstream of First Isolation Valve Note that Figure 3.a.3-6 shows the total amount of fire barrier destroyed (fiber plus particulate) . Note that no fire barrier material is destroyed by partial breaks smaller than 21 inches. E5-23

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Range of Debris with Average by Break Size 31" 14 30.5' 1 30'

                 --                                             --                                                 139 29.5' f"-ll !                                        --                                                _34 29'
                                                              -                                            l28 28.5'                                                                                         ~"23 28' 27.5"
                 --                             I

_'1.1

                                                           -                              1'1.2 27" 26'
                                                        --                            _-08
                                                    '1                            99 25" 24'                                                              _7 Ql N

23"

                 -                                                    n Vi      22"    -                      -                          70
  .:.:. 21"
                                      -                       6 I'll cc Ql    20" 19'

_3 58 18" - 8 17' 16' 9:. 15' _5 14"

                 -        --          _2 12"
                 -      -    2. !'8 10"  ~-.

8"  !:o:-J. 7 6" c()I 0 4 4" =o 0 2 2" 0 0 0 0.5" 0 0 0 Debris Generated (lb) Figure 3.a.3 Range of Epoxy Coatings Debris Generated by Partial Breaks Upstream of First Isolation Valve E5-24

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Range of Debris with Average by Break Size 31* 2S 41 30.S" 30" 8 29.5" 29" s 2s.s

• 3 I

2s* 32 27.S" 1 I 27" I 26. 2s*

                     -         I   --                                      _6

_8 I - 24" s QI 23" 2' N Vi 22" 23

       ~

Ill 21*

                     -                                         ll QI co 20*

19*

                     -                              18

_'() ~ 1s

• 16 17" §:1 16" 4 i s* 13 14" 2 12*

10* 8" "=O 6" r=o-o 2 4" lrO 0 1 2" 0 0 0 0.5 0 0 0 Debris Generated (lb) Figure 3.a.3 Range of IOZ Coatings Debris Generated by Partial Breaks Upstream of First Isolation Valve

b. Debris Generation/Zone of Influence (excluding coatings)

The objective of the debris generation/ZOI process is to determine , for each postulated break location : (1) the zone within which the break jet forces would be sufficient to damage materials and create debris ; (2) the amount of debris generated by the break jet forces .

1. Describe the methodology used to determine the ZOls for generating debris .

Identify which debris analyses used approved methodology default values . For debris with ZOls not defined in the guidance report/SE , or if using other tha n default values , discuss method(s) used to determine ZOI and the basis for each . Response to 3.b.1: For DEGBs , the ZOI is defined as a spherical volume about the break in which the jet pressure is higher than the destruction/damage pressure for a certain type of insulation , coatings , or other materials impacted by the break jet. E5-25

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) In a PWR reactor containment building, the worst-case pipe break would be a DEGB. In a DEGB, jets of water and steam would blow in opposite directions from the severed pipe. One or both jets could impact obstacles and be reflected in different directions. To take into account the double jets and potential jet reflections, NEI 04-07 (Reference 2) proposes using a spherical ZOI centered at the break location to determine the quantity of debris that could be generated by a given line break. For any break smaller than a DEGB (i.e., a partial break), NEI 04-07, Volume 2 (Reference 3) suggests a hemispherical ZOI centered at the edge of the pipe. Because these types of breaks could occur anywhere along the circumference of the pipe, the partial breaks were analyzed using hemispheres at eight different angles that are 45 degrees apart from each other around the pipe. Since different insulation types have different destruction pressures, different ZOls must be determined for each type of insulation. Table 3.b.1-1 shows the primary side break equivalent ZOI radii divided by the break diameter (LID) for each representative material in the VEGP Units 1 and 2 containment buildings. See Table 3.a.2-1 for ZOI sizes for SSBs. Table 3.b.1 Primary Side Break ZOI Radii for VEGP Insulation Types Destruction Pressure ZOI Radius/Break Diameter Insulation Type losi) (LID) Unjacketed Nukon 6 17.0* Qualified Coatings Unknown 4.0** Fire Barrier Material Unknown 11.7***

• NRC SE for NEI 04-07 (Reference 3)
  • "Revised Guidance Regarding Coatings Zone of Influence for Review of Final Licensee Responses to Generic Letter 2004-02" ADAMS # ML100960495
    • The destruction pressure of the lnteram E-50 series fire barrier material at VEGP is unknown. However, its ZOI size was assumed to be 11.70 based on comparison to the robustness of Temp-Mat.

In some cases, if the ZOI for a particular material is very large (i.e., it has a low destruction pressure or is located on a large pipe); the radius of the sphere may extend beyond robust barriers located near the break. Robust barriers consist of structures, such as concrete walls that are impervious to jet flow and prevent further expansion of the jet. Insulation in the shadow of large robust barriers can be assumed to remain intact to a certain extent (Reference 3, Section 3.4.2.3). Due to the compartmentalization of containment in VEGP Units 1 and 2, the insulation on the opposite side of the compartment walls can be assumed to remain intact. However, the SG compartments share an opening where a break jet could extend, so this was accounted for by including destruction of some of the insulation in these areas. All ZOls were truncated to account for robust barriers and compartment openings per NEI 04-07 Volume 2 (Reference 3). ES-26

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Volumetric debris quantities were determined by measuring the interference between a ZOI and its corresponding debris source. This was done within the CAD model. No insulation debris would be generated outside of the ZOls (Reference 2). This practice is considered acceptable by the NRC as stated in the SE for NEI 04-07 (Reference 3, Section 3.4.3.2).

2. Provide destruction ZOls and the basis for the ZOls for each applicable debris constituent.

Response to 3.b.2: See the Response to 3.b.1.

3. Identify if destruction testing was conducted to determine ZOls. If such testing has not been previously submitted to the NRC for review or information, describe the test procedure and results with reference to the test report(s).

Response to 3.b.3: VEGP applied the ZOI refinement discussed in NEI 04-07 Volume 2 (Reference 3, Section 4.2.2.1.1 ), which allows the use of debris-specific spherical ZOls. No new destruction testing was used to determine the ZOls listed above.

4. Provide the quantity of each debris type generated for each break location evaluated. If more than four break locations were evaluated, provide data only for the four most limiting locations.

Response to 3.b.4: Using the ZOls listed in this section, the breaks selected in the Response to 3.a.1, and the size distribution provided in the Response to 3.c.1 of this enclosure, quantities of generated debris for each break case were calculated for each type of insulation. The quantities of debris generated for the four most limiting break cases are listed below in Table 3.b.4-1 as determined in the debris generation calculation. The quantities of debris generated for the four most limiting break cases that do not fail any of the acceptance criteria in the NARWHAL evaluation (see Enclosure 3 Section 13.0) are listed below in Table 3.b.4-2. See Table 3.h.5-1 in the Response to 3.h.5 for the quantity of qualified and unqualified coatings for the four most limiting break locations. See Response to 3.d.3 for the quantity of latent debris. E5-27

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprie~ary) Table 3.b.4-1: The Four Overall Worst-Case Breaks Break Location 11201-004-6-RB 11201-001-5-RB 11201-001-3-RB 11201-004-4-RB Break Size 29" 29" 29" 29" Break Type DEGB DEGB DEGB DEGB Fine 289.3 287.5 280.9 276.0 Nukon Small 999.5 991.4 962.9 938.6 (ft3) Large 451.6 454.0 460.1 473.8 Intact 487.9 490.5 497.1 511.9 Fire Barrier Fine 0.0 2.4 2.4 0.0 (ft3) Small 0.0 2.4 2.4 0.0 Fire Barrier Particulate 0.0 26.9 26.8 0.0 (lbm) Table 3.b.4-2: The Four Worst-Case Breaks that Do Not Fail Any Acceptance Criteria for the Single Train Failure Equipment Configuration Break Location 11201-004-4-RB 11201-001-3-RB 11201-003-5-RB 11201-002-5-RB Break Size 20" 23" 19" 16" Break Type Partial Partial Partial Partial Fine 48.4 47.2 50.7 52.3 Nukon Small 151.2 160.5 186.5 168.0 (ft3) Large 122.1 80.9 46.6 118.6 Intact 132.0 87.4 50.3 128.1 Fire Barrier Fine 0.0 0.0 0.0 0.0 (ft3) Small 0.0 0.0 0.0 0.0 Fire Barrier Particulate 0.0 0.0 0.0 0.0 (lbm)

5. Provide total surface area of all signs, placards, tags, tape, and similar miscellaneous materials in containment.

Response to 3.b.5: Labels, tags, stickers, placards, and other miscellaneous or foreign materials were evaluated via walkdown. As with latent debris, a foreign material walkdown was only performed for Unit 1. However, based on the similarity between units discussed previously, Unit 1 data is considered applicable for Unit 2. The amount of foreign materials found during the walkdown was 2.0 ft2 . However, for conservatism, a total surface area of 4.0 ft2 was assumed in the VEGP debris generation calculation, and 50 ft2 was used in the NARWHAL conditional failure probability (CFP) calculation. E5-28

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary)

c. Debris Characteristics The objective of the debris characteristics determination process is to establish a conservative debris characteristics profile for use in determining the transportability of debris and its contribution to head loss.

1. Provide the assumed size distribution for each type of debris.

Response to 3.c.1: A summary of the material properties of the debris types found within containment are listed in Table 3.c.1-1 below. Table 3.c.1 Debris Material Properties Characteristic Density Debris Distribution Size (lbm/ft3) (µm) 2.4 (bulk) Nukon See section below 7 159 (fiber) Fiber (30% of total mass): 54.3 (bulk) 1.5 (fiber) 50% Fines 175 (fiber) Fire Barrier 50% Smalls Particulate (70% of total mass): 151 (particulate) 1O (particulate) 100% Particulate Qualified 208 (IOZ) 100% Particulate 10 Coatings 107 (Epoxy) 208 (IOZ) Unqualified and Degraded 100% Particulate 109 (Epoxy) 10 Coatings 122 (Alkyd) Nukon Low-Density Fiberglass Insulation The bulk density of Nukon is 2.4 pounds mass per cubic foot (lbm/ft3), and the individual fiber density is 159 lbm/ft3, per NEI 04-07 (Reference 2). The characteristic diameter of the individual fibers is 7 micrometers (µm). A baseline analysis of Nukon includes a size distribution with two categories: 60 percent small fines, and 40 percent large pieces, per NEI 04-07 (Reference 2). The debris generation calculation uses a four-category size distribution based on the guidance in NEI 04-07 Volume 2 (Reference 3). This guidance provides an approach for determining a size distribution for low-density fiberglass using the air jet impact test (AJIT) data, with conservatism added due to the potentially higher level of destruction from a two-phase jet. Within the 170 ZOI, the size distribution varies based on the distance of the insulation from the break (i.e., insulation debris E5-29

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprieftary) generated near the break location consists of more small pieces than insulation debris generated near the edge of the ZOI). Consequently, the following equations were developed to determine the fraction of fines (individual fibers), small pieces (less than 6 inches), large pieces (greater than 6 inches), and intact blankets as a function of the average distance between the break point and the centroid of the affected debris measured in units of pipe diameters (C). (OD H 4D) = 0.2 Ffines(C) (4D H 15D) = -0.01364

• C + 0.2546 (15D H 17D) = -0.025
• c + 0.425 (OD H 4D) = 0.8 Fsmalls (C) ( 4D H 15D) = -0.0682
• C + 1.0724 (15D H 17D) = -0.025
• c + 0.425 (OD H 4D) = 0 Frarge(C) ( 4D H 15D) = 0.0393
• C - 0.157 (15D H 17D) = -0.215
• c + 3.655 (OD H 4D) = 0 FintactCC) (4D H 15D) = 0.0425
• C - 0.170 (15D H 17D) = 0.265
• c - 3.505

2. Provide bulk densities (i.e., including voids between the fibers/particles) and material densities (i.e., the density of the microscopic fibers/particles themselves) for fibrous and particulate debris.

Response to 3.c.2: See the Response to 3.c.1 for the material and bulk densities of the various types of debris.

3. Provide assumed specific surface areas for fibrous and particulate debris.

Response to 3.c.3: Specific surface areas could be calculated for each debris type based on the characteristic diameter described in the Response to 3.c.1. However, specific surface areas were not calculated or used for the Vogtle head loss evaluation.

4. Provide the technical basis for any debris characterization assumptions that deviate from NRG-approved guidance.

Response to 3.c.4: The lnteram E-50 Series fire barrier material was assumed to be comprised of 70 percent particulate and 30 percent fiber by weight. These values fall within the ranges given in the E-50 Material Safety Data Sheet (MSDS). It was assumed that E5-30

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) the fiber constituent of 3M E-50 would fail as 50 percent fines and 50 percent small pieces sized between % inch and 4 inches. This engineering judgment is based on observations from exploratory testing with a 1500 psi pressure washer (the same type used for NEI fiber preparation). This assumption is conservative because the observations indicate that it is a very robust material and is less likely to break up into fines than low density fiberglass (LDFG).

d. Latent Debris The objective of the latent debris evaluation process is to provide a reasonable approximation of the amount and types of latent debris existing within the containment and its potential impact on sump screen head loss.

1. Provide the methodology used to estimate the quantity and composition of latent debris.

Response to 3.d.1: The evaluation for latent debris at VEGP was performed in a manner consistent with the NRC NEI 04-07 SE approved methodology (Reference 3, Section 3.5.2.3). The total source term was determined through the collection of debris samples from multiple locations throughout the containment. Conservatism was added by sampling those areas that exhibited unusually large concentrations of dirt and dust. In addition to dirt and dust, foreign materials and other debris sources were surveyed and documented including lint, paint chips, fibers, pieces of paper (shredded or intact), plastic, tape, adhesive labels, and fines or shards of thermal insulation, fireproof barrier, or other materials that are already present in the containment prior to a postulated break in a high-energy line inside containment. Vertical, horizontal, and equipment surfaces were sampled for dirt and dust by wiping with muslin cloth. Sample areas were chosen by cognizant engineering personnel with the intent to produce bounding results. The containment was divided into categories from which a minimum of three samples were taken. Prior to collecting samples, the containment was surveyed through a series of walkdowns to locate desirable sample locations.

2. Provide the basis for assumptions used in the evaluation.

Response to 3.d.2: See Response to 3.d.3. E5-31

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary)

3. Provide results of the latent debris evaluation, including amount of latent debris types and physical data for latent debris as requested for other debris under c.

above. Response to 3.d.3: Latent debris includes dirt, dust, lint, paint chips, fines, and shards of loose thermal insulation fibers that could potentially transport to the sump strainers during recirculation. Latent debris can be introduced into containment several ways, including by deterioration of items such as insulation and coatings and by personnel tracking in particulate and fibers from outside containment. The quantity of latent debris is calculated in the debris generation calculation. A walkdown of VEGP Unit 1 was performed to measure quantities of latent debris, and the total quantity was calculated based on those samples. The total amount of latent debris calculated based on walkdown data was 60 lbm, but 200 lbm is assumed in the debris generation calculation. This conservatively bounds the 60 lbm of actual latent debris with ample operating margin. Table 3.d.3-1 lists the assumed latent fiber and particulate constituents and their material characteristics. Latent debris is assumed to consist of 15 percent fiber and 85 percent particulate by mass, per the NRC NEI 04-07 SE (Reference 3, page 50). Based on NEI 04-07 Volume 2 (Reference 3, Sections 3.5.2.3, 3.7.2.3.2.3), the size and density of latent particulate were assumed to be 17.3 µm and 168.6 lbm/ft3, respectively. Additionally, the bulk density and microscopic density of latent fiber were assumed to be 2.4 lbm/ft3 and 93.6 lbm/ft3, respectively. Latent fiber is assumed to have a characteristic size of 5.5 µm. This is reasonably conservative, as it is the smallest fiber diameter listed in Table 3-2 of the general reference for low-density fiberglass found in NEI 04-07 (Reference 2). Table 3.d.3 Latent Fiber and Particulate Constituents Latent Bulk Microscopic Characteristic Debris Density Density Size (lbm) (lbm/ft3 ) (lbm/ft3 ) (µm) Particulate (85%) 170 - 168.6 17.3 Fiber (15%) 30 2.4 93.6 5.5 Total 200

4. Provide amount of sacrificial strainer surface area allotted to miscellaneous latent debris.

Response to 3.d.4: There is no sacrificial strainer area allotted to miscellaneous latent debris in addition to that documented in the Response to 3.b.5. E5-32

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary)

e. Debris Transport The objective of the debris transport evaluation process is to estimate the fraction of debris that would be transported from debris sources within containment to the sump suction strainers.

1. Describe the methodology used to analyze debris transport during blowdown, washdown, pool-fill-up, and recirculation phases of an accident.

Response to 3.e.1: The methodology used in the transport analysis is based on the NEI 04-07 guidance and the associated NRC SE (Reference 3), as well as the refined methodologies suggested by the SE in Appendices Ill, IV, and VI (Reference 3). The specific effect of each of four modes of transport was analyzed in the debris transport calculations for each type of debris generated. These modes of transport are:

• Slowdown Transport - the vertical and horizontal transport of debris to all areas of containment by the break jet
• Washdown Transport - the vertical (downward) transport of debris by the containment sprays, break flow, and condensation
• Pool Fill-Up Transport - the transport of debris by break and containment spray flows from the RWST to regions that may be active or inactive during recirculation
• Recirculation Transport - the horizontal transport of debris from the active portions of the recirculation pool to the sump screens by the flow through the ECCS The logic tree approach was applied for each type of debris determined from the debris generation calculation. The logic tree shown in Figure 3.e.1-1 is slightly different from the baseline. This departure was made to account for certain non-conservative assumptions identified by the NRC SE (Reference 3), including the transport of large pieces, erosion of small and large pieces, the potential for washdown debris to enter the pool after inactive areas have been filled, and the direct transport of debris to the sump screens during pool fill-up.

E5-33

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Debris Size. Blo.wdo'lm Wash down PoolfUI CFD Recilcufation Erosion IFraction of Debris Transport Transpcrt Tra.'ISpOrt I Transport I at Sump Retained on Structures up pH iranspon Coot:U-.ment I WasJi,:.dDown Sed:mer.t 1ransport

                         """'                                     ru.*uve r-oot I    Sedimeru:

Lovrer Sump Screens Contai-n:nent lrocll,,.,r<<>1 I Erodes to Fines Retained on StrucWres Remains intact UPP"' 1ransport Contafr.ment I I Erodes to Fines Wash.E-d Down Se<f:ment I Remains intact P.ieo:s uanspon Actve r<Xn I I Erodes to Fines Sediment I Debris Remains intact Generaitio."11 LOl't'S'  ::iump .:-creens Conta1unent Inactive-Pool I Erodes to Flnie-s Retained on I Struch.Jras Hema1ns mtact upp..- 1ransport Conb:mnent I J Erodes to Fine; Washed Dm*.n I Sediment I Remarns mtact Large P:ie-ces I transport Active Pool I I Erodes to Fines SedJnent I Remains intact

                                    ~  ...

Coota:runent

                                                               ~ump.........,..,,.ens lnacti.\.'ePool Rebinedon Strucb.Jres UPP"'                                                  lransport Coobimnent                                           II Washed DD""*YTl Sed:ment I    1ranspm L.3111" Piereswith Actvo Pocl I

Jackef.ng  ;:,eOiment Intact Lower SumpS~ns ContainmE!nt lnactivePoc:4 Figure 3.e.1-1: Generic Debris Transport Logic Tree The basic methodology for the VEGP transport analysis is summarized below.

1. The CAD model was provided as input to determine break locations and sizes.

2. The debris generation calculation was provided as input into the calculation for debris types and sizes.

3. Potential upstream blockage points were qualitatively addressed.

E5-34

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary)

4. The fraction of debris blown into upper containment and lower containment for each compartment was determined based on the volumes of upper and lower containment.

5. The fraction of debris washed down by containment spray flow was determined along with the locations where the debris would be washed down.

6. The quantity of debris transported to inactive areas or directly to the sump strainers was calculated based on the volume of the inactive and sump cavities proportional to the water volume at the time these cavities are filled.

7. The location of each type/size of debris at the beginning of recirculation was determined based on the break location.

8. A CFO model was developed in Flow-30 to simulate the flow patterns that would develop during recirculation.

9. A graphical determination of the transport fraction of each type of debris was made using the velocity and turbulent kinetic energy (TKE) profiles from the CFO model output, along with the determined initial distribution of debris.

10. The initial recirculation transport fractions from the CFO analysis were gathered to determine the final recirculation transport fractions for input into the logic trees.

11. The quantity of debris that could experience erosion due to the break flow or spray flow was determined.

12. The overall transport fraction for each type/size of debris was determined by combining each of the previous steps into logic trees.

Potential Upstream Blockage Points Potential upstream blockage points were qualitatively addressed in the debris transport calculation. It was determined that there are not any upstream blockage points in the VEGP containment building. Upstream effects are discussed in the Response to 3.1. i I I I E5-35

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) CFO Model of Containment Recirculation Pool A diagram showing the significant parts of the CFO model is shown in Figure 3.e .1-

2. The sump mass sink and the various direct and runoff spray regions are highlighted .

Accumulator Inj ection Modeled Spray line Modeled Mass Drainage Through Source Loop 2 Annulus Modeled Modeled Spray Drainage Falling Pressurizer Surge Through Steam Line Modeled Generator Mass Source Compartm nts Loop 4 Modeled Sinks Mass Source CS Sump Strainer Module Mass Sinks Figure 3.e.1-2: Significant Features in CFO Model The key CFO modeling attributes/considerations included the following : Computational Mesh A rectangular mesh was defined in the CFO model that was fine enough to resolve important features , but not so fine that the simulation would take excessively long to run. A mesh spacing of 5 inch by 5 inch was used in the x and y directions and 3-inch to 4-inch mesh space was used in the z direction . ES-36

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Modeling of Containment Spray Flows For CFO cases with CS activated, various plan and section drawings, as well as the containment building CAD model, were considered. Spray water would drain to the pool through many pathways. Some of these pathways include the steam generator enclosures, the various openings in the operating deck, the annulus through the various open sections of grating, and the refueling canal drains. The sprays were introduced near the surface of the pool. Modeling of Break Flow The water falling from the postulated break would introduce momentum into the containment pool that influences the flow dynamics. This break stream momentum was accounted for by introducing the break flow to the pool at the velocity that a freefalling object would have if it fell the vertical distance from the location of the break to the surface of the pool. Modeling of the Sump Strainers Each sump strainer in VEGP consists of four columns of stacked disks with a solid plate on top. In the CFO model, each strainer was modeled with a plate above it to prevent flow from entering through the top of the strainers. The sump strainers were modeled as having flow across their surfaces proportional to the areas of the strainers. A negative flow rate was set for the sump mass sink, which tells the CFO model to draw the specified amount of water from the pool over the entire exposed surface area of the mass sink obstacle. Turbulence Modeling Several different turbulence-modeling approaches can be selected for a Flow-30 calculation. The approaches (ranging from least to most sophisticated) are:

• Prandtl mixing length
• Turbulent energy model
• Two-equation k-E model
• Renormalized group theory (RNG) model
• Large eddy simulation model The RNG turbulence model was determined to be the most appropriate for this CFO analysis. The RNG model has a large spectrum of length scales that would likely exist in a containment pool during emergency recirculation. The RNG approach applies statistical methods in a derivation of the averaged equations for turbulence quantities (such as TKE and its dissipation rate). RNG-based turbulence schemes rely less on empirical constants while setting a framework for the derivation of a range of models at different scales.

E5-37

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Steady-State Metrics The CFO models were started from a stagnant state at a defined pool depth and run long enough for steady-state conditions to develop. A plot of mean kinetic energy was used to determine when steady-state conditions were reached. Checks were also made of the velocity and turbulent energy patterns in the pool to verify that steady-state conditions were reached. Debris Transport Metrics The metrics for predicting debris transport during recirculation are the TKE necessary to keep debris suspended, and the flow velocity necessary to tumble sunken debris along the floor or lift it over a curb. Debris transport metrics have been derived or adopted from data. The metrics utilized in the VEGP transport analysis originate from the sources below.

• NUREG/CR-6772 Tables 3.1 and 3.2 (Reference 37)
• NUREG/CR-6808 Figure 5.2 and Tables 5-1 and 5-3 (Reference 39)

Graphical Determination of Debris Transport Fractions for Recirculation The following steps were taken to determine what percentage of a particular type of debris could be expected to transport through the containment pool to the emergency sump screens. Detailed explanations of each bullet are provided in the paragraphs below.

• Colored contour velocity and TKE maps were generated from the Flow-30 results in the form of bitmap files indicating regions of the pool through which a particular type of debris could be expected to transport.
• The bitmap images were overlaid on the initial debris distribution plots and imported into AutoCAD with the appropriate scaling factor to convert the length scale of the color maps to feet.
• Closed polylines were drawn around the contiguous areas where velocity and TKE were high enough that debris cou.ld be carried in suspension or tumbled along the floor to the sump strainers for uniformly distributed debris.
• The areas within the closed polylines were determined using an AutoCAD querying feature.
• The combined area within the polylines was compared to the initial debris distribution area.
• The percentage of a particular debris type that would transport to the sump strainers was determined based on the above comparison.

Plots showing the TKE and the velocity magnitude in the pool were generated for each case to determine areas where specific types of debris would be transported. The limits on the plots were set according to the minimum TKE or velocity metrics necessary to move each type of debris. The overlying yellow areas represent ES-38

Enclosure 5 Suppl*emental Response to NRC Generic Letter 2004-02 (Non-Proprietary) regions where the debris would be suspended, and the red areas represent regions where the debris would be tumbled along the floor (see Figure 3.e.1-4). The yellow TKE portion of the plots is a three-dimensional representation of the TKE. Since the TKE is a three-dimensional representation, the plots do not show the TKE at any specific elevation. Rather, any debris that is shown to be present in this yellow area will transport, regardless of the elevation of TKE in the pool. The velocity portion of the plots represents the velocity magnitude just above the floor level (1.5 inches), where tumbling of sunken debris could occur. Directional flow vectors were also included in the plots to determine whether debris in certain areas would be transported to the sump strainers or transported to less active regions of the pool where it could settle to the floor (blue regions). The following figures and discussion are presented as an example of how the transport analysis was performed for a generic small debris type. This same approach was used for other debris types analyzed at VEGP. As shown in Figure 3.e.1-3, the small debris (depicted by green shading) was initially assumed to be uniformly distributed between the break location and the sump strainers. The break location in this scenario is a break in the annulus on the pressurizer surge line. E5-39

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Figure 3.e.1-3: Distribution of Small Debris in Lower Containment Figure 3.e.1-4 shows that the turbulence (yellow regions) and the velocity (red regions) in the pool (blue regions) are high enough to transport the generic small debris to the sump strainers during recirculation . The initial distribution area (Figure 3.e.1-3) was overlaid on top of the plot showing tumbling velocity , TKE , and flow vectors (Figure 3.e.1-4) to determine the recirculation transport fraction (Figure 3.e.1-5). ES-40

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Figure 3.e.1-4: TKE and Velocity with Limits Set at Suspension/ Tumbling of Small Generic Debris E5-41

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Figure 3.e.1-5: Floor Area where Small Generic Debris Would Transport to the Sump Strainers (hatched area) This same analysis was applied for each type of debris at VEGP . Recirculation-pool transport fractions were identified for each debris type associated with the location of its initial distribution . This includes a recirculation transport fraction for debris blown to lower containment, debris washed down inside the secondary shield wall , and debris washed down into the annulus. Erosion Discussion Due to the turbulence in the recirculation pool and the force of break and spray flow, Nukon debris may erode into smaller pieces , making transport of this debris to the strainer more likely. Results of the Drywell Debris Transport Study indicate that debris exposed to containment sprays above the recirculation pool undergo an erosion fraction of less than 1%. Therefore , a 1% erosion fraction for debris held up on gratings and other miscellaneous structures at Vogtle was used . E5-42

Enclosure 5 Supplemental Response to NRC G'eneric Letter 2004-02 (Non-Proprietary) Erosion Test Discussion To estimate erosion that would occur in the recirculation pool at Vogtle, generic 30-day erosion testing was performed. ((

                           ))1 Input Parameters The flow conditions used for the testing were based on prototypical plant flow conditions. The target velocity selected for the erosion testing was ((
                    )) 1 based on the tumbling velocity required to transport small pieces of LDFG (Reference 37). Since small pieces of LDFG would transport at higher velocities, the non-transporting small pieces of LDFG on a containment pool floor would be exposed to a velocity less than ((              ))1. Typically, in regions where the velocity is lower than 0.12 ft/s, the pool is relatively quiescent, and the turbulence levels are very low. Based on a review of the average turbulence levels for various plants in the quiescent "non-transport" regions, a target turbulence of ((
    ))1 was selected for the erosion testing.

To prevent potential contamination of the samples from the minerals in tap water, deionized (DI) water was selected for the erosion testing. In prototypical plant conditions, the containment pool water is borated and buffered. Based on observations during chemical effects testing, chemical precipitates tend to accumulate on exposed fiberglass. This effect can mask the actual erosion. However, by using pure water, this phenomenon was eliminated in the erosion testing. Erosion Test Durations Table 3.e.1-1 shows the test matrix for the erosion testing. The length of the pre-test was selected based on the longest period of time that any one filter was installed during the primary test and post-test, or 5 days. The length of the primary test was based on a full 30-day mission time. The length of the post-test was based on the filter measurements during the primary test as well as the initial filter measurements during the post-test, or 5 days. Since the primary test measurements showed that the majority of erosion occurs within the first 10 days, a 10-day test was determined to be an adequate length of time to accomplish the purposes of the post-test. This test was run for a total of 13 days. 1 Alien trade secret ES-43

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Table 3.e.1-1 Erosion Test Durations Test Duration Description Quantify weight change of a filter under test conditions Pre-Test 5 days without anv fiben:1lass in the flume. Measure overall weight loss of fiber clumps after 30 days exposure to flow. Also, determine time dependent Primary Test 30 days erosion curve by measuring weight change of filters throughout. Determine repeatability of primary test, and confirm that there are no unknown long-term phenomena that Post-Test 13 days resulted in a non-conservative weight gain of the fiber samples durinQ the later staQes of the primarv test.

2. Provide the technical basis for assumptions and methods used in the analysis that deviate from the approved guidance.

Response to 3.e.2: The methodology used in the transport analysis is based on and does not deviate from the NRC approved NEI 04-07 guidance and the associated NRC SE (Reference 3) for refined analyses, as well as the refined methodologies suggested by the SE in Appendices Ill, IV, and VI.

3. Identify any computational fluid dynamics codes used to compute debris transport fractions during recirculation and summarize the methodology, modeling assumptions, and results.

Response to 3.e.3: To assist in the determination of recirculation transport fractions, several CFO simulations were performed using Flow-30, a commercially available software package. Seven breaks were investigated that included single- and two-train recirculation with sprays both on and off to ensure a conservative representation of the post-LOCA containment-sump flow velocities. Breaks were also evaluated inside and outside the secondary shield wall to determine which scenario(s) would maximize debris transport. The simulation results include a series of contour plots of velocity and TKE. These results have been combined with settling and tumbling velocities from the GSl-191 literature to determine the recirculation transport fractions for all debris types present in the VEGP containment building. See Response to 3.e.1 for additional discussion of the CFO results. E5-44

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary)

4. Provide a summary of, and supporting basis for, any credit taken for debris interceptors.

Response to 3.e.4: No credit was taken for debris interceptors.

5. State whether fine debris was assumed to settle and provide basis for any settling credited.

Response to 3.e.5: No credit was taken for settling of fine debris.

6. Provide the calculated debris transport fractions and the total quantities of each type of debris transported to the strainers.

Response to 3.e.6: The following debris transport fractions listed in Table 3.e.6-1 through Table 3.e.6-14 are inputs to the NARWHAL CFP calculation. Note that these fractions result in 'the bounding quantity of debris transported to the strainer. The debris transport quantities are provided in Tables 3.e.6-15 and 3.e.6-16. E5-45

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Slowdown Transport Table 3.e.6-1 shows the bounding blowdown transport fractions as a function of break location and debris type. Table 3.e.6-1: Slowdown Transport Fractions Transport Fraction To Upper To Lower Remaining Break Location Debris Type Containment Containment in (UC) (LC) Compartment Fines (all) 80% 20% 0% Steam Generator Small Nukon & 39% 61% 0% Compartments Fire Barrier Large Nukon 0% 100% 0% Fines (all) 80% 20% 0% Small Nukon & Reactor Cavity 39% 61% 0% Fire Barrier Large Nukon 0% 100% 0% Fines (all) 80% 20% 0% Pressurizer Small Nukon & 69% 9% 22% Compartment Fire Barrier Large Nukon 0% 0% 100% Fines (all) 80% 20% 0% Annulus - Small Nukon & Pressurizer Surge 35% 18% 47% Fire Barrier Line Lan:ie Nukon 0% 25% 75% Fines (all) 80% 20% 0% Annulus - Small Nukon & Accumulator 17% 83% 0% Fire Barrier Injection Line Larqe Nukon 0% 100% 0% E5-46

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Washdown Transport Table 3.e.6-2 shows the bounding washdown transport fractions as a function of containment spray activation and debris type. Note that these transport fractions do not depend on the location of the break. Table 3.e.6-2: Washdown Transport Fractions Transport Fraction Sprays Debris Type Washed Down Washed Down Washed Down Initiated? in Annulus Inside SSW RFC Drains Fines (all) 53% 37% 10% Small Nukon & Yes 43% 37% 10% Fire Barrier Large Nukon 0% 0% 10% Nukon Fines/ 10% No Latent Debris All Others 0% 0% 0% Pool-Fill Transport Table 3.e.6-3 shows the bounding pool fill transport fractions as a function of debris type. Table 3.e.6-3: Pool fill Transport Fractions Pool Fill Transport Fraction Debris Type Elevator Cavity Each ECCS Sump Fines (all) 2% 0.75% Small Nukon & Fire Barrier 2% 0.75% Large Nukon 2% 0.75% Unqualified Coatings 0% 0% Recirculation Transport For the recirculation transport fractions, seven different cases were evaluated in the debris transport calculation. These cases are listed below:

• Case 1: LBLOCA in SG Compartment Loop 4, Sprays not Activated, Two Trains Operational
• Case 2: LBLOCA in Pressurizer Surge Line in Annulus, Sprays not Activated, Two Trains Operational
• Case 3: LBLOCA in Accumulator Injection Line in Annulus, Sprays not Activated, Two Trains Operational
• Case 4: LBLOCA in SG Compartment Loop 4, Sprays not Activated, One Train Operational E5-47

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary)

• Case 5: LBLOCA in SG Compartment Loop 4, Sprays not Activated, Two Trains Operational, High Water Level *
• Case 6: LBLOCA in SG Compartment Loop 2, Sprays Activated, Two Trains Operational
• Case 7: LBLOCA in SG Compartment Loop 4, Sprays Activated, Two Trains Operational The bounding recirculation transport fractions for fine debris are shown in Table 3.e.6-4.

Table 3.e.6-4: Recirculation Transport Fractions for Fine Debris Washed Inside Washed Washed Break Case Sump Secondary Shield In Down Recirculation Wall Annulus RFC RHRA 50% NA NA NA Case 1 RHR B 50% NA NA NA RHRA 50% NA NA NA Case 2 RHR B 50% NA NA NA RHRA 50% NA NA NA Case 3 RHR B 50% NA NA NA Case4 RHR B 100% NA NA NA RHRA 50% NA NA NA Case 5 RHR B 50% NA NA NA CSA 21% 21% 21% 21% RHRA 29% 29% 29% 29% Case 6 CS B 21% 21% 21% 21% RHR B 29% 29% 29% 29% CSA 21% 21% 21% 21% RHRA 29% 29% 29% 29% Case 7 CS B 21% 21% 21% 21% RHR B 29% 29% 29% 29% E5-48

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) The bounding recirculation transport fractions for small fiber debris are shown in Table 3.e.6-5. Table 3.e.6-5: Recirculation Transport Fractions for Small Fiber Debris Washed Inside Break Washed In Washed Case Sump Secondary Shield Recirculation Annulus Down RFC Wall RHRA 0% NA NA NA Case 1 RHR B 0% NA NA NA RHRA 9% NA NA NA Case 2 RHR B 33% NA NA NA RHRA 0% NA NA NA Case 3 RHR B 0% NA NA NA Case 4 RHRB 0% NA NA NA RHRA 0% NA NA NA Case 5 RHR B 0% NA NA NA CSA 24% 23% 28% 0% RHRA 0% 0% 9% 0% Case 6 CS B 8% 8% 19% 100% RHR B 9% 10% 7% 0% CSA 20% 17% 25% 0% RHRA 0% 0% 10% 0% Case 7 CS B 8% 9% 6% 100% RHR B 26% 25% 20% 0% E5-49

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) The bounding recirculation transport fractions for large fiber debris are shown in Table 3.e.6-6. Table 3.e.6-6: Recirculation Transport Fractions for Large Fiber Debris Washed Inside Washed Break Washed Case Sump Secondary Shield In Recirculation Down RFC Wall Annulus RHRA 0% NA NA NA Case 1 RHR B 0% NA NA NA RHRA 4% NA NA NA Case 2 RHR B 20% NA NA NA RHRA 0% NA NA NA Case 3 RHR B 0% NA NA NA Case4 RHR B 0% NA NA NA RHRA 0% NA NA NA Case 5 RHR B 0% NA NA NA CSA 0% NA NA NA RHRA 0% NA NA NA Case 6 CS B 0% NA NA NA RHR B 0% NA NA NA CSA 0% NA NA NA RHRA 0% NA NA NA Case 7 CS B 0% NA NA NA RHR B 0% NA NA NA E5-50

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Overall Debris Transport Transport logic trees were developed for each size and type of debris generated. These trees were used to determine the total fraction of debris that would reach the sump strainers in each of the postulated cases. The bounding overall transport fractions are presented in Table 3.e.6-7 through Table 3.e.6-14. Note that the near annulus breaks represent breaks that are within close proximity to the strainers in the annulus, and that the far annulus breaks represent breaks that are far away from the strainer in the annulus. The values below are slightly different than what is calculated in NARWHAL. This is because the total transport fractions are entered in NARWHAL and the time-dependent fiber accumulation on the strainers is calculated based on the flow split. Additionally, some fiber penetrates the strainers and accumulates on the core, which also contributes to this slight difference. Table 3.e.6-7: Overall Transport Fractions for an SG Compartment/ Reactor Cavity Break, Two Trains Operational, CS On Debris Type CSA RHRA CSB RHRB Total Nukon Individual Fibers 21% 29% 21% 29% 100% Nukon & Fire Barrier Small Pieces 20% 5% 13% 25% 63% Nukon Large Pieces 3% 4% 3% 4% 14% Nukon Intact Pieces 0% 0% 0% 0% 0% Fire Barrier Fines & Particulate 21% 29% 21% 29% 100% Qualified Coatings (IOZ, Epoxv) 21% 29% 21% 29% 100% Unqualified Epoxy Coatings Particulate 21% 29% 21% 29% 100% Unqualified IOZ CoatinQs Particulate 21% 29% 21% 29% 100% Unqualified Alkyd Coatings Particulate 21% 29% 21% 29% 100% Latent Dirt/Dust Particulate & Fiber 21% 29% 21% 29% 100% Table 3.e.6-8: Overall Transport Fractions for an SG Compartment/ Reactor Cavity Break, One Train Operational, CS Off Debris Type CSA RHRA CSB RHRB Total Nukon Individual Fibers NA NA NA 27% 27% Nukon & Fire Barrier Small Pieces NA NA NA 6% 6% Nukon Large Pieces NA NA NA 10% 10% Nukon Intact Pieces NA NA NA 0% 0% Fire Barrier Fines & Particulate NA NA NA 27% 27% Qualified Coatings (IOZ, Epoxy) NA NA NA 27% 27% Unqualified Epoxy Coatings Particulate NA NA NA 47% 47% Unqualified IOZ Coatings Particulate NA NA NA 60% 60% Unqualified Alkyd Coatings Particulate NA NA NA 100% 100% Latent Dirt/Dust Particulate & Fiber NA NA NA 31% 31% E5-51

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Table 3.e.6-9: Overall Transport Fractions for an SG Compartment/ Reactor Cavity Break, Two Trains Operational, CS Off Debris Type CSA RHRA CSB RHRB Total Nukon Individual Fibers NA 14% NA 14% 28% Nukon & Fire Barrier Small Pieces NA 3% NA 3% 6% Nukon Large Pieces NA 6% NA 6% 12% Nukon Intact Pieces NA 0% NA 0% 0% Fire Barrier Fines & Particulate NA 14% NA 14% 28% Qualified Coatings (IOZ, Epoxy) NA 14% NA 14% 28% Unqualified Epoxy Coatings Particulate NA 23% NA 23% 46% Unqualified IOZ Coatings Particulate NA 30% NA 30% 60% Unqualified Alkyd Coatings Particulate NA 50% NA 50% 100% Latent Dirt/Dust Particulate & Fiber NA 16% NA 16% 32% Table 3.e.6-10: Overall Transport Fractions for a Pressurizer Compartment Break, Two Trains Operational, CS Off Debris Type CSA RHRA CSB RHRB Total Nukon Individual Fibers NA 14% NA 14% 28% Nukon & Fire Barrier Small Pieces NA 1% NA 3% 4% Nukon Large Pieces NA 0% NA 0% 0% Nukon Intact Pieces NA 0% NA 0% 0% Fire Barrier Fines & Particulate NA 14% NA 14% 28% Qualified Coatings (IOZ, Epoxy) NA 14% NA 14% 28% Unqualified Epoxy Coatings Particulate NA 23% NA 23% 46% Unqualified IOZ Coatings Particulate NA 30% NA 30% 60% Unqualified Alkyd Coatings Particulate NA 50% NA 50% 100% Latent Dirt/Dust Particulate & Fiber NA 16% NA 16% 32% E5-52

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Table 3.e.6-11: Overall Transport Fractions for a Near Annulus Break, Two Trains Operational, CS Off Debris Type CSA RHRA CS B RHRB Total Nukon Individual Fibers NA 14% NA 14% 28% Nukon & Fire Barrier Small Pieces NA 2% NA 6% 8% Nukon LarQe Pieces NA 2% NA 6% 8% Nukon Intact Pieces NA 0% NA 0% 0% Fire Barrier Fines & Particulate NA 14% NA 14% 28% Qualified Coatings (IOZ, Epoxy) NA 14% NA 14% 28% Unqualified Epoxy Coatings Particulate NA 23% NA 23% 46% Unqualified IOZ Coatings Particulate NA 30% NA 30% 60% Unqualified Alkyd CoatinQs Particulate NA 50% NA 50% 100% Latent Dirt/Dust Particulate & Fiber NA 16% NA 16% 32% Table 3.e.6-12: Overall Transport Fractions for a Far Annulus Break, Two Trains Operational, CS Off Debris Type CSA RHRA CS B RHRB Total Nukon Individual Fibers NA 14% NA 14% 28% Nukon & Fire Barrier Small Pieces NA 5% NA 5% 10% Nukon LarQe Pieces NA 6% NA 6% 12% Nukon Intact Pieces NA 0% NA 0% 0% Fire Barrier Fines & Particulate NA 14% NA 14% 28% Qualified Coatings (IOZ, Epoxy) NA 14% NA 14% 28% Unqualified Epoxy Coatings Particulate NA 23% NA 23% 46% Unqualified IOZ Coatings Particulate NA 30% NA 30% 60% Unqualified Alkyd Coatings Particulate NA 50% NA 50% 100% Latent Dirt/Dust Particulate & Fiber NA 16% NA 16% 32% E5-53

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Table 3.e.6-13: Overall Transport Fractions for a Near Annulus Break, Two Trains Operational, CS On Debris Type CSA RHRA CSB RHRB Total Nukon Individual Fibers 21% 29% 21% 29% 100% Nukon & Fire Barrier Small Pieces 21% 3% 10% 27% 61% Nukon Large Pieces 3% 4% 3% 4% 14% Nukon Intact Pieces 0% 0% 0% 0% 0% Fire Barrier Fines & Particulate 21% 29% 21% 29% 100% Qualified Coatings (IOZ, Epoxy) 21% 29% 21% 29% 100% Unqualified Epoxy Coatings Particulate 21% 29% 21% 29% 100% Unqualified IOZ Coatings Particulate 21% 29% 21% 29% 100% Unqualified Alkyd Coatings Particulate 21% 29% 21% 29% 100% Latent Dirt/Dust Particulate & Fiber 21% 29% 21% 29% 100% Table 3.e.6-14: Overall Transport Fractions for a Far Annulus Break, Two Trains Operational, CS On Debris Type CSA RHRA cs 8 RHRB Total Nukon Individual Fibers 21% 29% 21% 29% 100% Nukon & Fire Barrier Small Pieces 20% 5% 13% 25% 63% Nukon Large Pieces 3% 4% 3% 4% 14% Nukon Intact Pieces 0% 0% 0% 0% 0% Fire Barrier Fines & Particulate 21% 29% 21% 29% 100% Qualified Coatings (IOZ, Epoxy) 21% 29% 21% 29% 100% Unqualified Epoxy Coatings Particulate 21% 29% 21% 29% 100% Unqualified IOZ Coatings Particulate 21% 29% 21% 29% 100% Unqualified Alkyd Coatings Particulate 21% 29% 21% 29% 100% Latent Dirt/Dust Particulate & Fiber 21% 29% 21% 29% 100% E5-54

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Transported Debris Quantities The transported debris quantities for the most limiting break cases identified in Tables 3.b.4-1 and 3.b.4-2 are shown below and were derived using the debris transport fractions provided in this section for a single train failure case. The debris transport quantities for the four most limiting break cases are listed below in Table 3.e.6-15 as determined in the NARWHAL CFP calculation. The quantities of debris transported for the four most limiting break cases that do not fail any of the strainer or core acceptance criteria are listed in Table 3.e.6-16. Note that the fiber quantity includes fines, small pieces, large pieces, intact pieces, and latent fiber debris. To calculate the transported quantities of debris presented in the following tables, the blowdown, washdown, pool-fill, and recirculation data (Table 3.e.6-1 through Table 3.e.6-6) are input into NARWHAL. However, the NARWHAL CFP calculation takes into account certain factors that the transport calculation does not consider in order to calculate the time-dependent arrival of debris on the strainer. For example, it takes into account various factors such as the RHR strainer switching over to recirculation before the CS strainer, and the flow split between the strainers. Therefore, the calculation of debris transported to the strainer in the NARWHAL CFP calculation is not a straightforward one. All breaks listed in the tables below occur on the hot leg and activate containment sprays since the break size for each break is greater than 15". The recirculation transport fractions for Case 7 (LBLOCA in SG Compartment Loop 4, Sprays Activated, Two Trains Operational) from the transport calculation were conservatively input into NARWHAL for the single train failure with containment sprays activated. This was done since there was not a CFO case in the transport calculation that examined one train failure with containment sprays activated, and it is conservative since the turbulence and the velocities in the pool during recirculation for two train operation with containment sprays activated are maximized. E5-55 I ,

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Table 3.e.6-15: Transported Debris for the Four Overall Worst-Case Breaks Break Location 11201-004-6-RB 11201-001-5-RB 11201-001-3-RB 11201-004-4-RB Break Size 29" 29" 29" 29" Break Type DEGB DEGB DEGB DEGB Fiber (ft3) RHR 639.8 635.4 617.9 605.8 cs 279.5 277.6 272.7 267.4 Particulate RHR 2, 166.5 2, 166.3 2,159.7 2,156.1 (lbm) cs 935.0 934.9 941.5 940.0 Calcium Phosphate RHR 73.5 73.5 73.5 73.5 (lbm) cs 42.3 42.2 42.1 42.1 Sodium Aluminum RHR 86.2 86.1 85.8 85.8 Silicate (lbm) cs 0.5 0.5 0.5 0.5 Fire Barrier Particulate RHR 0.0 20.3 20.2 0.0 (lbm) cs 0.0 8.8 8.8 0.0 Fire Barrier RHR 0.0 1.0 1.0 0.0 Fiber (ft3) cs 0.0 0.4 0.4 0.0 ES-56

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Table 3.e.6-16: Transported Debris for the Four Worst-Case Breaks that Do Not Fail the Acceptance Criteria Break Location 11201-004-4-RB 11201-001-3-RB 11201-003-5-RB 11201-002-5-RB Break Size 20" 23" 19" 16" Break Type Partial Partial Partial Partial Fiber (ft3 ) RHR 107.2 107.1 107.6 107.5 cs 46.9 46.7 46.4 46.6 Particulate RHR 2,103.5 2, 106.3 2, 109.9 2, 108.4 (lb) cs 917.0 918.2 910.5 909.9 Calcium Phosphate RHR 46.4 38.3 34.0 47.8 (lbm) cs 16.4 14.1 12.8 16.8 Sodium Aluminum RHR 54.3 52.5 51.5 54.6 Silicate (lbm) cs 0.3 0.3 0.3 0.3 Fire Barrier Particulate RHR 0.0 0.0 0.0 0.0 (lbm) cs 0.0 0.0 0.0 0.0 Fire Barrier RHR 0.0 0.0 0.0 0.0 Fiber (ft3 ) cs 0.0 0.0 0.0 0.0 ES-57 I I

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary)

f. Head Loss and Vortexing The objectives of the head loss and vortexing evaluations are to calculate head loss across the sump strainer and to evaluate the susceptibility of the strainer to vortex formation.

1. Provide a schematic diagram of the emergency core cooling system (ECCS) and containment spray systems (CSS).

Response to 3.f.1: See Figure 3.f.1-1 and Figure 3.f.1-2 for ECCS and CSS schematics, respectively. E5-58

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) IRC OA:C I I RWST CL1 CL2 CL3 cu CL1 IRC I ORC HUA I I-A 1 UO CL2 I IH111~ CL3 v c cu T 1111 TO R EGEN HX Figure 3.f.1-1 Emergency Core Cooling System Schematic ES-59

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) CO TAl NWENT SPRAY NO ZZLES C S PUMP A Fl

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~-*-*-*-*-*-*-*-*-*-*-*J I o;;; * - * - * - * - * - * - * - * - * - *** -. '002A _j SGD3A I -- * -

• Figure 3.f.1-2 Containment Spray System Schematic ES-60

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary)

2. Provide the minimum submergence of the strainer under small-break loss-of-coolant accident (SBLOCA) and large-break loss-of-coolant accident (LBLOCA) conditions.

Response to 3.f.2: The sump strainers are fully submerged during recirculation in all cases except reactor nozzle breaks (see Table 3.g.1-3). The highest elevation of the RHR strainer disk is 53-1 /4 inches (or 4.438 ft) above the containment floor (see Figure 3.f.2-1 ). The submergence of the highest elevation point of the RHR strainer is conservatively taken to be its minimum submergence. The height of the 16-disk RHR strainer bounds (i.e., is greater than) the height of the 14-disk CSS strainer, and the minimum submergence of the RHR strainers is always less than the minimum submergence of the CS strainers. 4.438 f1 I lotl *

• loot Ittl *
• It ol Figure 3.f.2-1: Side View of 16-Disk RHR Strainer The RHR and CS sump strainers are fully submerged under an LBLOCA that is not a reactor nozzle break. As shown in the Response to 3.g.1, the minimum LBLOCA water level during recirculation for a break that is not a reactor nozzle break is 4 .536 ft, and the minimum submergence of the RHR strainer is 0.098 ft.

The RHR sump strainers are also fully submerged for an LBLOCA at a reactor nozzle when CS is not activated . As shown in the Response to 3.g.1, the minimum water level during recirculation for this case is 4.977 ft, and the minimum submergence of the RHR strainer is 0.539 ft. The RHR and CS sump strainers are fully submerged under an SBLOCA. As shown in the Response to 3.g .1, the minimum SBLOCA water level during recirculation is 5.186 ft. Therefore, the minimum submergence of the RHR strainer is 0.748 ft. The RHR and CS sump strainers are not fully submerged for an LBLOCA caused by a reactor nozzle break that actuates CS . As shown in the Response to 3.g.1 , the E5-61

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) minimum water level for this case (3.054 ft) occurs at the start of recirculation. This pool level is 1.384 ft below the top of the RHR strainer. It should be noted that the water level increases to 4. 788 ft when the sump recirculation switchover is complete. This corresponds to a strainer submergence of 0.35 ft. According to Regulatory Guide (RG) 1.82 (Reference 107), the total strainer head loss of a partially submerged strainer should be less than half the submerged height of the strainer. This ensures the average hydrostatic head of the submerged portion of the strainer will be greater than the head loss through the debris bed. This requirement was used when evaluating partially-submerged strainers for VEGP.

3. Provide a summary of the methodology, assumptions, and results of the vortexing evaluation. Provide bases for key assumptions.

Response to 3.f.3: Summary of Vortex Tests In 2009, vortex testing was performed on a prototypical strainer module to observe the size, shape, and location of vortices that may develop as both the flow rate through the strainer and the submergence of the strainer module were varied. The vortex tests were performed during the head loss test described in the Response to 3.f.4. Both clean screen and debris laden vortex tests were performed. See Figure 3.f.4-1 for the layout of the test strainer and test tank. Two vortex tests were conducted at clean strainer conditions, as summarized below.

• The first clean strainer vortex test was started at a submergence level of 3.625 inches and an average approach velocity of 0.0258 ft/s. No vortexing was observed. The average approach velocity was then increased to 0.0355 ft/s. No vortexing was observed. The water level was then reduced to 1.825 inches below the top of the strainer. Again, no vortexing was observed.
• A second clean strainer vortex test was started with a strainer submergence of 4.175 inches and an approach velocity of 0.0306 ft/s. This approach velocity was maintained throughout the duration of this test. The water level in the tank was reduced to just below the top of the strainer. Some pump noise was audible and a small surface swirl was visible in the front right corner of the tank but no persistent vortices were observed.

Debris laden vortex tests were performed at the end of the thin bed and full-load head loss tests after adding all conventional and chemical debris. The test loop setup was the same as that used for the clean screen vortex test and the approach velocity for all debris laden vortex tests was 0.0136 ft/s. It should be noted that, during the course of the thin bed and full-load tests, the tank water level was ES-62

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) maintained at 3.675 +/- 0.5 inches above the strainer, and no appreciable vortices were visually observed.

• Water level was slowly reduced at the end of the thin bed test. Air ingestion was not observed until the water level was 0.25 inches below the top of the strainer.
• For the first full debris load test, when the water level was reduced to approximately 3 inches above the strainer, air-entrainment vortices were observed. The vortices became persistent when the water level reached 2.25 inches above the strainer.
• At the end of the second full debris load test, when the water level was reduced to approximately 3 inches above the strainer, air-entrainment vortices were observed. The vortices were not persistent until the water level reached 1.5 inches below the top of the strainer.

Vortexing of Plant Strainers For reference in the discussion below, the average approach velocity of the 16-disk RHR strainers in the plant is 0.0122 ft/s, and the average approach velocity for the 14-disk CS strainers in the plant is 0.0098 ft/s. These approach velocities are calculated using the strainer flow rates and surface areas shown in Table 3.f.3-1. These approach velocities are well bounded by that used during the clean strainer (0.0355 ft/s) and debris-laden (0.0136 ft/s) vortex tests. T a bl e 3 .. - f 3 1 Pl an t St ram . er A verage A~pproac hVe Ioc1"f1es RHR Strainer CS Strainer Flow Rate (gpm) 3,700 2,600 2 Surface Area (ft ) 677.6 590 Approach Velocity (ft/s) 0.0122 0.0098 Table 3.g.1-3 presents the minimum strainer submergences for different breaks at different times following the accident. As shown in the table, for the following four cases, the strainer submergence is greater than 3 inches after the start of sump recirculation. The debris-laden vortex test showed that, even with all debris loaded to the strainer, no vortices were observed for submergences greater than 3 inches. It is reasonable to conclude that vortexing will not occur for these four cases because at the start of recirculation 3 , the strainer is expected to be clear of debris.

• SBLOCA without CS

       * . MBLOCA without CS 3

The start of recirculation for the breaks that actuate containment spray is the time when the RWST level reaches the Low-Low level alarm and the sump suction valves for the RHR pumps open. For the breaks that do not actuate containment spray, start of recirculation is when the switchover of the RHR pump suctions from the RWST to sump is completed at the Empty level alarm. ES-63

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary)

• LBLOCA without CS
• Reactor nozzle break without CS For an LBLOCA with CS, the minimum strainer submergence is 0.098 ft (or 1.2 inches) at the start of recirculation when the strainer is still clean. Since the clean strainer vortex showed no vortices even for a partially submerged strainer, it is concluded that vortexing is not a concern for this case at the start of recirculation.

Table 3.g.1-3 shows that, for an LBLOCA with CS , the minimum strainer submergence increases to 1.803 ft when the switchover to sump recirculation is completed, which occurs approximately 20 minutes after start of recirculation. Since this submergence is much higher than the 3-inch limit identified by the debris-laden vortex test and only small amount of debris is expected to be transported to the strainer during the period of time it would take to reach a submergence of 3 inches, vortexing will not be an issue at debris-laden conditions for an LBLOCA with CS. Lastly, for LBLOCA reactor nozzle breaks with CS, Table 3.g.1-3 shows that the strainer is partially submerged at the start of recirculation. After that, the minimum strainer submergence increases over time and is equal to 0.35 ft (or 4.2 inches) when switchover to sump recirculation is completed, which is approximately 20 minutes after the start of recirculation. Similar to the discussion presented above for the LBLOCA with CS, the reactor nozzle break with CS is also bounded by the debris-laden vortex test with respect to formation of vortices. Based on the discussion above, vortexing is not a concern for any of the analyzed break scenarios.

4. Provide a summary of methodology, assumptions, and results of prototypical head loss testing for the strainer, including chemical effects. Provide bases for key assumptions.

Response to 3.f.4: Head loss tests were performed to measure the head losses caused by conventional debris (fiber and particulate) and chemical precipitate debris generated and transported to the sump strainers following a LOCA. The test program used a test strainer, debris quantities, and flow rates that were prototypical to VEGP. Different I . test cases were performed with the thin bed and full debris load protocols, following the 2008 NRC Staff Review Guidance (Reference 111 ). The results of the head loss tests provided a matrix of head loss data for various combinations of conventional and chemical debris loads. This matrix was used in the NARWHAL CFP calculation to determine the debris head loss for the debris load associated with each postulated break. ES-64

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Test Setup The test strainer assembly consists of seven stacked disks that are duplicates of the disks in the plant strainer. The top surface of the top disk and bottom surface of the bottom disk are solid steel rather than perforated plate . This results in a total of six disks contributing to the effective surface area of the test strainer. The test strainer was placed in a corner of a 6 ft tall by 6 ft wide by 10 ft long test tank on top of a horizontal plenum that simulated the plenum configuration present in the plant. The gaps between the test strainer and the surrounding walls of the test tank simulated the configuration of the plant strainer. See Figure 3.f.4-1 for an illustration of the test strainer and tank. Figure 3.f.4-1: Isometric of Head Loss Test Strainer Assembly inside Test Tank A schematic piping diagram of the test loop is provided in Figure 3.f.4-2 below. The test loop had a recirculation pump that took suction from the plenum underneath the test strainer and returned the water back into the test tank . The return flow exit into the tank was located such that the turbulence from the flow did not affect the debris bed on the test strainer. A flow element was used to measure the flow rate through the loop . Flow control valves and heating and cooling loops were used to control the test flow rate and water temperature . The test water was maintained at temperatures of at least 80 degrees F throughout the tests. ES-65

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) T11nk H atcr

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2500 gpm Pump Figure 3.f.4-2: Piping Diagram of Head Loss Test Loop Test Parameters and Scaling The test strainer replicates all hydraulic dimensions of the plant strainer except for the number of strainer disks and the number of gaps between disks . The test debris quantities and test flow rate were scaled from plant values based on the ratio between the numbers of gaps between disks of the test strainer to that of the plant strainer. This is analogous to scaling the debris loads and flow rate based on the ratio of the test strainer surface area to the plant strainer surface area . The surface area of the test strainer was calculated to be 65 .57 ft2 . This surface area of the test strainer was scaled from the RHR strainer surface area based on the number of disks. The post-modification RHR strainer at the plant consists of four stacks of 15.5 disks (the top surface of the top disk for each stack is solid steel , so the top disk counted as ~ a disk) . The test strainer has a similar configuration as the plant strainer except the test strainer has only seven disks with the top surface of the top disk and bottom surface of the bottom disk being solid steel. Therefore, the active surface area of the test strainer is equivalent to six disks . This simple scaling method is reasonable because the test strainer disks and spacer rings were fabricated to the same dimensions as the disks and spacer rings installed in the plant. ES-66

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) To scale debris loads from the test to the plant, the debris load is multiplied by the ratio of the plant strainer area to the test strainer area (shown above). The 16-disk RHR strainers have a surface area of 677.6 ft2, and the 14-disk CS strainers have a surface area of 590 ft2 (see Table 3.f.3-1 ). This scaling is used when determining the strainer debris limits at the plant scale in Table 3.f.5-1. During VEGP head loss testing, a nominal test flow rate of 400 gpm was used. Therefore, using the test strainer surface area shown above, the average approach velocity of the test strainer was 0.0136 ft/s, which bounds the approach velocities of the current plant RHR and CS strainers, as shown in Table 3.f.3-1. In the NARWHAL CFP calculation, the measured strainer head losses from the 2009 testing were corrected from the testing conditions (e.g., strainer approach velocity and water temperature) to plant conditions of interest using the flow sweep data collected during testing. See response to 3.f.10. Debris Materials and Preparation The following materials were used as conventional debris for head loss testing: Nukon, lnteram E-54A, green silicon carbide powder, and silica sand. The method of preparation prior to introduction to the test tank for each material is discussed below. Nukon fines were used as surrogate for latent fiber, as recommended in NEI 04-07 and associated NRC SE (References 2 and 3, respectively). Nukon was also used to represent fines and small pieces of LDFG insulation debris. To prepare Nukon fines, Nukon fiberglass sheets were first shredded and inspected to ensure that the shredded Nukon met the size distribution requirements defined in NUREG/CR-6808 (Reference 39). Afterward, the required quantity was weighed out and boiled for 10 minutes to remove the binder. The boiled fiber was then placed in a bucket of water that was within +/-10 degrees F of the testing water temperature. The fiber was mixed thoroughly with a paint mixer attached to an electric drill until a homogeneous slurry was formed. Prepared fiber fines consisted of Class 1-3 fibers as defined in NUREG/CR-6224 (Reference 29). For small pieces of Nukon, the preparation was similar. However, the prepared small pieces consisted of interwoven strands of fiber, equivalent to or smaller than the Class 4 small fiber clusters, as defined in NUREG/CR-6224. lnteram fire blanket was processed (double-shredded) through a leaf shredder, similar to the manner in which the fiber debris was shredded. After shredding, the lnteram was added to buckets with sufficient water to suspend the debris. The buckets were then stirred to wet and suspend the lnteram debris in the bucket. Silica sand was used as a surrogate for latent particulate debris and was prepared by Performance Contracting, Inc. (PCI). The size distribution of the silica sand was ES-67

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) prepared to be consistent with that of latent particulate debris provided in the NRC SE for NEI 04-07 (References 3). Green silicon carbide powder was used as a surrogate for both qualified and unqualified coatings. Since the density of the green silicon carbide powder and that of the actual coating are different, the mass of the surrogate material added was adjusted such that the volumes of the surrogate material and actual coatings debris were matched. Per NEI 04-07 and the associated NRC SE (References 2 and 3, respectively), the coatings particulate debris was assumed to be 10 µm diameter spheres. The majority of the silicon carbide surrogate had a size distribution range from 4 µm to 20 µm, a median size of 10.25 µm, and a mean size of 10.46 µm. The required amount of silicon carbide and silica sand was weighed out and placed in a bucket of water with a temperature within +/-10 degrees F of the testing temperature. The particulate was mixed with water using an electric paint stirrer until no agglomeration or clumping was observed. Before introducing the particulate to the test tank, all particulate batches were mixed once again with an electric paint stirrer to create a thin slurry. Two types of chemical debris surrogates were used for the head loss testing: sodium aluminum silicate (SAS) and calcium phosphate. The chemical debris was prepared in accordance with WCAP-16530-NP-A (Reference 73). The 1-hour settling volume for each batch of chemical precipitates was determined at the time the batch was produced. The chemical precipitate settling time was also measured within 24 hours from the time the surrogate was to be used. Chemical precipitates that did not meet the settling requirements were discarded and not used for testing. Debris Introduction Debris was added at the side of the tank adjacent to the return flow exit into the tank and away from the simulated sump floor and walls. This allowed for an even and representative debris accumulation on the test strainer. A sparger system was installed on the return flow exit in the tank away from the strainer to aid in suspension of debris. Two mechanical mixers were also installed at the tank corners opposite from the test strainer. During the thin-bed test, all of the debris was added over the mixers while the recirculation pump was running. Additional manual agitation was applied to help the lnteram arrive near the test strainer. During the later full-load tests, the lnteram was introduced to the floor area immediately adjacent to the test strainer to aid the transport. The agitation from the sparger, mechanical mixers, and manual agitation helped keep the debris suspended in the water, and near-field settling was not credited. The debris bed formed on the strainer was not affected by debris addition or agitation in the tank. ES-68

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) After conventional debris introduction was completed for each test, chemical precipitate debris was added to the test tank. Calcium phosphate was first introduced in batches, and the head loss was allowed to stabilize between batches. SAS batches were added last. Head Loss Test Cases and Results Three head loss tests were performed for VEGP: one thin bed test and two full debris load tests. The two full-load tests targeted the same flow conditions and debris loads. For the thin bed test (VOG-1-TB), all of the particulate debris (including lnteram, coatings surrogate, and latent particulate surrogate) was introduced to the test tank at the beginning of the test. Once all of the particulate debris was added and allowed to circulate through the test loop, fine fiber batches were incrementally added in batch sizes equivalent to a 1/8-inch theoretical uniform debris bed thickness. Only fiber fines were used for the thin bed test. Thin-bed formation was observed visually, via head loss and turbidity measurement. The head loss was allowed to stabilize (less than or equal to 1 percent change over a 1-hour period) after the final batch of fiber fines was added. Once the head loss stabilized, chemical precipitates were incrementally added. The debris batch composition and size for the thin bed test are summarized in Table 3.f.4-1. Table 3.f.4-1: Debris Batches Added for the Thin Bed Head Loss Test Test Nukon Test !Test Silicon Test Test Calcium Quantity lnteram Carbide Dirt/Dust Phosphate Test SAS Batch (lbm) Quantity Quantity Quantity Quantity Quantity (lbm) (lbm) (lbm) (L) (L) Fines Smalls VOG-1.2-TB-P 0 0 29.15 358.42. 5.3 0 0 VOG-1.3-TB-F1 1.49 0 0 0 0 0 0 VOG-1.4-TB-F2 1.49 0 0 0 0 0 0 VOG-1.5-TB-F3 1.49 0 0 0 0 0 0 VOG-1.6-TB-F4 1.49 0 0 0 0 0 0 VOG-1.7-TB-F5 1.49 0 0 0 0 0 0 VOG-1.10-TB-CP1 0 0 0 0 0 160.24 0 VOG-1.11-TB-CP2 0 0 0 0 0 160.24 0 VOG-1.12-TB-CP3 0 0 0 0 0 160.24 0 VOG-1. 13-TB-NAS 1 0 0 0 0 0 0 122.86 VOG-1.14-TB-NAS2 0 0 0 0 0 0 122.86 VOG-1.15-TB-NAS3 0 0 0 0 0 0 122.86 Total 7.45 0 29.15 358.42 5.3 480.72 368.58 E5-69

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Table 3.f.4-2 summarizes the stabilized head losses after adding each batch of debris during the thin bed test. Table 3.f.4-2: Thin Bed Head Loss Test Results Stabilized Head Loss Batch (ft-H20) VOG-1.2-TB-P 0.171 VOG-1.3-TB-F1 0.226 VOG-1.4-TB-F2 0.262 VOG-1.5-TB-F3 0.308 VOG-1.6-TB-F4 0.368 VOG-1.7-TB-F5 0.625 VOG-1.10-TB-CP1 1.02 VOG-1.11-TB-CP2 1.54 VOG-1.12-TB-CP3 1.65 VOG-1.13-TB-NAS 1 2.12 VOG-1.14-TB-NAS2 2.27 VOG-1.15-TB-NAS3 2.56 For the full-load tests (VOG-2-FL-B and VOG-2-FL-82), the particulate and fiber debris was introduced simultaneously in equal batches maintaining the same fiber to particulate ratio until the full conventional debris load was reached. This debris addition sequence resulted in a homogenous debris bed accumulation. Each debris batch consisted of Nukon fiber fines and small pieces, and particulate debris. After the final batch of conventional debris was introduced, the head loss was allowed to stabilize with a less than 1 percent change over a 1-hour period. Chemical precipitates were then incrementally added. The debris batch composition and size for the full-load tests are summarized in Table 3.f.4-3. I , E5-70

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Table 3.f.4-3: Debris Batches Added for the Full-load Head Loss Tests Test Nukon Test Test Silicon Test Test Calcium Test SAS Quantity lnteram Carbide Dirt/ Dust Phosphate Quantity Batch (lbm) Quantity Quantity Quantity Quantity (L) Fines Smalls (lbm) (lbm) (lbm) (L) VOG-2.2-FL-F1 4.58 2.04 7.29 89.61 1.32 0 0 VOG-2.3-FL-F2 4.58 2.04 7.29 89.61 1.32 0 0 VOG-2.4-FL-F3 4.58 2.04 7.29 89.61 1.32 0 0 VOG-2.5-FL-F4 4.58 2.04 7.29 89.61 1.32 0 0 VOG-2.6-FL-CP1 0 0 0 0 0 160.24 0 VOG-2. 7-FL-CP2 0 0 0 0 0 160.24 0 VOG-2.8-FL-CP3 0 0 0 0 0 160.24 0 VOG-2.9-FL-NAS1 0 0 0 0 0 0 122.86 VOG-2.1 O-FL-NAS2 0 0 0 0 0 0 122.86 VOG-2.11-FL-NAS3 0 0 0 0 0 0 122.86 Total 18.32 8.16 29.16 356.64 5.28 480.72 368.58 As discussed above, two full-load head loss tests were performed. However, the first full-load test VOG-2-FL-B reported higher head losses. Table 3.f.4-4 summarizes the stabilized head losses after adding each debris batch during test VOG-2-FL-B. The measured head losses from the full-load test are much higher than the thin-bed test. Table 3.f.4-4: Bounding Full-load Head Loss Test Results Stabilized Head Loss Batch (ft-H20) VOG-2.2-FL-F1 0.276 VOG-2.3-FL-F2 1.06 VOG-2.4-FL-F3 2.42 VOG-2.5-FL-F4 5.46 VOG-2.6-FL-CP1 5.29 VOG-2.7-FL-CP2 6.22 VOG-2.8-FL-CP3 6.57 VOG-2.9-FL-NAS 1 7.16 VOG-2.1 O-FL-NAS2 7.24 VOG-2.11-FL-NAS3 11.81 E5-71

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary)

5. Address the ability of the design to accommodate the maximum volume of debris that is predicted to arrive at the screen.

Response to 3.f.5: The 2009 head loss test program evaluated debris loads based on the Nukon debris quantities calculated using a 7D ZOI from WCAP-16710-P, which was later rejected by the NRG. The current Nukon debris quantities were calculated with a 17D ZOI in BADGER, and this results in fiber debris loads greater than that tested. Therefore, the debris quantities used in the 2009 test program do not bound what is predicted for some of the breaks using BADGER. Debris limits were implemented in the NARWHAL analysis for each strainer in operation for a given scenario. The debris limits were applied to individual strainers, not to the total amount of transported debris. If the debris on the strainer exceeded any of the debris limits, a failure was recorded for that postulated break. The debris limits are based on the maximum quantity of conventional and chemical debris that was tested in 2009. Table 3.f.5-1 shows the debris limits for each of the debris types at the test scale and plant scale for one RHR strainer. Note that the debris limits at the plant scale are determined by multiplying the debris limits at the test scale by the ratio of the RHR strainer area (677.6 ft2) to the test strainer area (65.57 ft 2 ), as discussed in the response to 3.f.4. Note that the calcium phosphate debris and SAS debris are converted from volume to mass using concentrations of 5 g/L and 11 g/L, respectively. Table 3.f.5-1: Debris Limit Failure Criteria Debris Limit at Debris Limit Debris Type Plant Scale for One at Test Scale RHR Strainer Fiber 11.03 ft3 113.98 ft 3 Particulate 363.72 lbm 3758.68 lbm 4 Fire Barrier (particulate) 20.41 lbm 210.92 lbm Fire Barrier (fiber) 3.65 ft3 37.72 ft 3 Calcium Phosphate 5.30 lbm 54.77 lbm Sodium Aluminum Silicate 8.94 lbm 92.39 lbm 4 Fire barrier debris was treated as 70% particulate and 30% fiber. E5-72

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary)

6. Address the ability of the screen to resist the formation of a "thin bed" or to accommodate partial thin bed formation.

Response to 3.f.6: The "thin-bed effect" is defined as the relatively high head losses associated with a low-porosity (or high particulate to fiber ratio) debris bed formed by a thin layer of fibrous debris that can effectively filter particulate debris. The 2009 VEGP head loss testing included a test for thin-bed effects. During this test, the full particulate load was added into the test tank first, followed by fiber fines in batches equivalent to a 1/8-inch theoretical uniform bed thickness. This batching schedule allowed the formation of a debris bed with a high particulate to fiber ratio. As a result, any thin-bed effects, should they occur, would be captured by the measured head losses. As discussed in Section 3.f.10, head loss testing results from the thin bed test are used by the NARWHAL CFP calculation for postulated debris loads less than or equal to 3.1 ft3 of fiber at test scale. See Section 3.f.10 for additional discussion of how the total head loss is determined.

7. Provide the basis for strainer design maximum head loss.

Response to 3.f.7: There are several failure criteria evaluated by NARWHAL based on the head loss across the strainer: strainer structural margin, strainer debris limits, strainer partial submergence limits, void fraction limits, flashing, and pump NPSH. A postulated break that exceeds one or more of these criteria for the RHR strainers/pumps is considered to be a failure of the ECCS system. Each of the failure criteria was evaluated at each time step within the NARWHAL model to determine if an ECCS system failure would occur. Strainer Structural Margin Limits The strainer structural margin for each strainer is 24.0 ft. The head loss across each of the RHR and CS strainers due to conventional and chemical debris loading is compared to this value to ensure that the structural margin was not exceeded. See Section 3.k.1 for additional information on how the structural margin was determined. Debris Limits See Section 3.f.5 for discussion of conventional and chemical debris limits used in NARWHAL. ES-73

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Unsubmerged Strainer Limits If the strainers are partially submerged, the NARWHAL CFP calculation assumed that the strainer would fail if the head loss across the debris bed and strainer is equal to or greater than half of the submerged strainer height per RG 1.82 (Reference 25). Note that the pump NPSH and strainer structural limits are also applicable for a partially submerged strainer. The NARWHAL CFP calculation tracks time-dependent accumulation of debris on the strainer. When the strainer is partially submerged, the evaluation only credits the active (i.e., submerged) portion of the strainers for flow and debris accumulation. Void Fraction Limits A pump failure due to degasification was recorded if the steady state gas void fraction at the pump is greater than 2 percent by volume. Note that bubble compression was not credited. Flashing Failure Limits A flashing failure was recorded for a postulated break if, at any time during sump recirculation, the pressure downstream of the strainer was lower than the vapor pressure at the sump temperature. The pressure downstream of the strainer was calculated by NARWHAL based on the strainer submergence, containment pressure and head loss across the strainer. Note that a small increase in containment pressure was credited in the flashing analysis, see Section 3.f.14 for additional information. Pump NPSH Limits A pump failure was recorded if the head loss across the strainer exceeded the clean strainer NPSH margin. (i.e., margin that is available for debris laden head loss). It should be noted that because the safety injection (SI) pumps and centrifugal charging pumps (CCPs) take suction from the RHR pumps during recirculation, only the NPSH margins of the RHR and CS pumps were calculated in NARWHAL. See Section 3.g.16 for details of the NPSH margin used in NARWHAL.

8. Describe significant margins and conservatisms used in head loss and vortexing calculations.

Response to 3.f.8: Vortexing Testing Testing was conducted to determine if vortexing is expected to occur. As discussed in the Response to 3.f.3, the vortex tests were performed at both clean strainer and debris-laden conditions. ES-74

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) All vortex tests used strainer approach velocities higher than those expected for the plant strainer (0.0122 ft/sand 0.0098 ft/s for the RHR and CS strainers, respectively, see Table 3.f.3-1 ). The clean strainer vortex tests used strainer approach velocities up to 0.0258 - 0.0355 ft/s. For the debris laden vortex tests, a strainer approach velocity of 0.0136 ft/s was used. As shown in the response to 3.f.3, plant strainer minimum submergence at the start of the recirculation is compared with the submergence limit established by the debris-laden vortex tests. It should be noted that these tests were performed after all conventional and chemical debris has been added to the test tank. This is conservatively bounding because, at the start of recirculation, the strainer is expected to be clear of debris. Strainer Head Loss The quantity of latent debris used to determine the strainer head loss is 200 lbm, but the actual amount of latent debris documented for the plant is only 60 lbm. Similarly, the amount of miscellaneous debris used in the analysis is 50 ft2 , but, as stated in Response to 3.b.5, the amount of miscellaneous debris was conservatively assumed in the debris generation calculation to be 4 ft2, which bounds the 2 ft2 identified in containment during the walkdown. When correcting the debris head loss from the test conditions (e.g., water temperature and strainer approach velocity) to plant conditions, head loss coefficients from both the full debris load test and thin bed test were applied in the analysis. The coefficients that result in the higher head loss are used to calculate the debris head loss. Additionally, specific flow sweep data points were excluded from the analysis if the use of the points would result in lower corrected head loss values. See section 3.f.10 for additional discussion. Finally, the rule-based approach described in Section 3.f.10 conservatively applies the maximum head loss result from the thin bed test for all postulated breaks with minimal fiber debris generation. Similarly, all postulated breaks that result in debris at the strainer greater than the thin bed test and less than or equal to the maximum amount of debris tested in the full load test have the maximum head loss from the full load test.

9. Provide a summary of methodology, assumptions, bases for the assumptions, and results for the clean strainer head loss calculation.

Response to 3.f.9: The clean strainer head loss was calculated to be 4.40 in-H20 at the RHR pump runout flow rate of 4,500 gpm. This flow rate conservatively bounds the flow rate of the RHR and CS strainers. ES-75

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) The clean strainer head loss was calculated by modeling flow through one 18-disk strainer stack per the following steps:

• The cross-sectional areas of flow for various parts of the strainer stack were calculated from the physical dimensions of the strainer components.
• Loss coefficients were calculated for the flow paths through the strainer components based on flow path geometry. Loss coefficients were determined for the perforated plate, wire cloth, and converging cross flow from the flow through the disk with flow through the core tube.
• A system of mass balance and energy balance equations were iteratively solved to calculate the flow and resulting pressure drop for each disk in the stack.
• The difference between the initial pressure and the pressure of the fluid before entering the plenum was calculated and reported as the head loss through the strainer stack.
• The head loss inside the sump pit below the strainer stacks was also included.

For each pit, the flow through the four strainer stacks combines inside the space below the strainer assembly and upstream of the ECCS or CS suction pipe openings. Head losses due to flow exiting the strainer stacks and turning inside the pit were accpunted for based on conservative loss coefficients and velocities. Several assumptions were used when applying the above methodology to determine the clean strainer head loss. The temperature of water was assumed to be 120 degrees F, the strainer was assumed to be fully submerged (i.e., flow is through all disks), and head loss along the outside face of the disks, elevation head, and coupling effects were ignored. Additionally, friction loss between and within the strainer disks was ignored as it is negligible compared to the screen and perforated plate losses. The surface friction loss in the strainer core was also ignored because the radial influx of water from the strainers disks and spacers prevents significant flow and friction loss along the surface of the strainer core. The flow turn inside the pit after exiting a strainer stack was conservatively assumed to be confined in a 90° steel mitre bend. It is acceptable to use the clean screen head loss calculated for an 18-disk strainer as the clean screen head loss for a 16-disk strainer for the following reasons:

• The flow distribution in the clean screen head loss calculation shows that the vast majority of flow is through the first six disks closest to the pump suction.

In fact, only 0.1 percent of the total flow comes from the top two disks.

• A flow rate of 3,700 gpm through the RHR strainer is used in the NARWHAL analysis rather than the pump runout flow rate of 4,500 gpm used in the clean screen head loss calculation. The clean screen head loss at 3,700 gpm is less than the clean screen head loss at 4,500 gpm.

ES-76

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary)

10. Provide a summary of methodology, assumptions, bases for the assumptions, and results for the debris head loss analysis.

Response to 3.f.10: The total head loss across the strainer is the sum of the clean strainer head loss, the conventional debris head loss, and the chemical head loss. The conventional and chemical head loss values were based on VEGP-specific head loss test results that were corrected from the test conditions (i.e., strainer approach velocity and water temperature) to plant conditions. Additionally, the test results were extrapolated to the end of the 30-day strainer mission time. Debris Head Loss Correction A head loss correction factor (based on the strainer approach velocity and pool temperature) was implemented into NARWHAL to scale the measured head losses from test conditions to plant conditions. For each time step for which conventional and chemical head losses are evaluated, the head loss value is corrected based on the plant flow rate through the strainer and the pool temperature. The correction was performed based on the debris bed characteristics obtained through flow sweeps conducted during head loss tests. For the 2009 test program, flow sweeps were performed at the end of the thin bed and full-load tests. To account for the uncertainty in the flow sweeps for each test, the resulting correction parameters from both tests were applied at each time step, and the maximum resulting head loss was returned. Table 3.f.10-1 and Table 3.f.10-2 show the flow sweep data for the thin bed and full load tests, respectively. Note that the thin bed test flow sweep was conducted at a water temperature of 86 degrees F, and that the full load test was conducted at a water temperature of approximately 93 degrees F. The flow sweep data was used to determine the correction parameters, which w then used to scale measured head losses. The volumetric flow rates documented in Tables 3.f.10-1 and 3.f.10-2 were converted to approach velocities using the test strainer area of 65.57 ft2. Table 3.f.10-1: Thin Bed Test Flow Sweep Data Flow Rate Head Loss Approach Velocity (gpm) (ft-H20) (Ws) 403 2.6 0.0137 371 2.37 0.0126 200 1.31 0.0068 436 2.87 0.0148 395 2.56 0.0134 E5-77

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Table 3.f.10-2: Full Load Test Flow Sweep Data Flow Rate Head Loss Approach Velocity (gpm) (ft-H20) (ft/s) 393 11 .81 0.0134 369 10.61 0.0125 200 3.47 0.0068 446 11 .57 0.0152 395 8.81 0.0134 Figure 3.f.10-1 and Figure 3.f.10-2 show the debris head loss as a function of approach velocity for the thin bed and full load test flow sweeps , respectively. In addition , the data was fit with a second -order polynomial in the following form : Here, K1 and K2 are the fitting coefficients (as shown in the figures) , and v is the strainer approach velocity in fUs . Note that the polynomial was forced through the origin , because the head loss would be zero at an approach velocity of zero. Also , note that for the full load test, the test data points represented by the two orange points were excluded from the curve fit. This produces a conservative curve fit because it results in a higher predicted head loss for lower approach velocities. 3.5 3 y = 240.89x2 +187 .72x R' =0.9995 2.5

        -~      2                                                                                                    .. **

1:1! **** ***** _g ..*.... **

         ~    1.5 I                                                       ........ .**********

1 .... *** 0.5 1 . .* ********************** 0 .* ************* - 0 0 .002 0.004 0.006 0.008 0.01 0.012 0.014 0.016 Average Approach Velocity {ft/s) Figure 3.f.10-1: Head Loss Fitting Coefficients for Thin Bed Test E5-78

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) 14 y = 57252x 2 +123.56x 12 2 R = 0.9999 ..*** *

         ~

3 VI VI 10 8

         ~C1I 6                                                                                       ... ***

I .** 4 2 ... **** 0 . ........... *** 0 0.002 0.004 0.006 0.008 0.01 0.012 0.014 0.016 Average Approach Velocity (ft/s) Figure 3.f.10-2: Head Loss Fitting Coefficients for Full Load Test With these curves defined , the head loss fitting parameter can be calculated . NARWHAL accepts a, b , and flow sweep head loss as inputs into the following correction equation . X _ a X µ X Vstrainer + b xp X v~tra i ner HL - liPHL Nomenclature : XHL = Head loss correction factor Vstrainer = Approach velocity of the strainer at plant condition , ft/s a = Coefficient determined from flow sweep curve fitting parameter K 2 and water viscosity at test temperature , K2 /µ b = Coefficient determined from flow sweep curve fitting parameter K1 and water density at test temperature , K1 /p

µ                 =       Viscosity of water at plant condition , lbm/(ft-s) p                 =       Density of water at plant condition , lbm/ft 3 LlPH L            =       Head loss at the test approach velocity and temperature , and the flow sweep debris load , ft-H20 The thin bed head loss test was conducted at a temperature of 86 degrees F, which corresponds to a water density of 62 .16 lbm/ft3 and a water viscosity of 0.000536 lbm/(ft-s). The head loss correction coefficients were calculated as follows for the thin bed test flow sweep data :

ES-79

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) K2 187.725 52 *ft a= - = = 350,223 .9 - -

     µ     0.000536 lbm                        lbm ft

• 5 52 Ki 240.89 ft 52 . ft 2 b=-= =3.875--

p 62.161~1;'1 lbm LlPHL = 2.6 ft The full load head loss test was conducted at a temperature of 93 degrees F, which corresponds to a water density of 62.08 lbm/ft3 and a water viscosity of 0.000494 lbm/(ft-s) . The head loss correction coefficients are calculated as follows for the full load test flow sweep data : K2 123.56 5 52 *ft a= - = lb = 250,121.5 u;---

     µ    0.000494 ~                             m ft

• 5 52 Ki 57,252 ft 52 . ft2 b=- = lbm = 922.229 lbm p 62.08 ft3 LlPHL = 11.81 ft By substituting the values of a, b, and LlPHL into the formula above , a head loss correction factor XHL can be calculated for each set of flow sweep data. As stated earlier, the two correction factors calculated from the thin-bed and full debris load flow sweeps were multiplied by the total debris head loss at each time step and the higher resulting head loss was returned. Note that the total debris head loss is the sum of the conventional debris head loss, the chemical head loss , and the extrapolation constant (where applicable) . The clean screen head loss is not corrected to different strainer approach velocities or pool temperatures.

Debris Head Loss Extrapolation To address extrapolation of the head loss tests to the end of 30-day mission time , a head loss extrapolation constant was applied . The extrapolation constant was determined using the raw test data from the end of the head loss test. The raw test data was smoothed using a locally weighted least-squares method. The first order derivative of the smoothed data was reviewed to ensure that the slope of the data was trending towards zero, suggesting that the head loss profile was stabilizing . ES-80

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) A natural logarithmic function was fitted to the smoothed data and the function was shifted upwards , which bounded any peaks observed after the last chemical addition . The curve fit also had a similar slope (i .e. , rate of increase in head loss over time) as the recorded head losses at the end of the test. This head loss extrapolation was performed for the thin -bed test and both full debris load tests. Figure 3.f.10-3 shows the recorded head losses and the logarithmic curve fit before and after the adjustment for the first full debris load test. Note that the test data used for the extrapolation analysis was recorded at least 12 hours after adding the last batch of chemical debris. The extrapolation constant was calculated for all three tests . Since the constant of the first full debris load test (3.89 ft-H 20) is larger than the other tests , this value was conservatively used for all NARWHAL analysis . Note that this extrapolation constant is at the testing condition and is corrected to plant conditions using the same approach as the debris head losses (see discussions earlier in this response). The extrapolation constant was applied at 7.5 hours after the accident. 14 13.5 13

                                ..                     Adjusted head loss correlation 12.5      ***                  Head Loss" 12.9 + 0.42591.ln(Time) 12 Natural log curve flt to smoothed data
                                                                        .~                       -
                                                                  ~
                                                            ~

y = m1 + rn2 ' ln( mO) 11 Value Erra mi 10.644 0.028743 m2 0.42591 0.019719 10.5 Chl&q 3.9565 NA thed data R 0.90907 NA 10

• 0 2 3 4 5 6 7 8 Time [hrj Figure 3.f.10-3: Logarithmic Curve Fit of Bounded Test Data Used for 30-Day Head Loss Extrapolation Clean Strainer Head Loss The clean strainer head loss varies as a function of strainer approach velocity.

However, the bounding clean strainer head loss of 4.40 ft at 4 ,500 gpm was used for all cases in NARWHAL. No flow or temperature correction was applied to the clean strainer head loss. ES-81

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Conventional Debris Head Loss NARWHAL uses a rule-based approach to calculate head loss based on the results of head loss testing. As shown in Table 3.f.10-3, if the fiber debris load at the strainer is less than the tested quantity from the thin bed test (3.1 ft3 at test scale) , the maximum thin bed conventional debris head loss was returned. If the quantity was greater than what was tested in the thin bed test, the conventional head loss of the full-load test was returned . For a given time step, NARWHAL scaled the plant debris load to the test scale based on the active plant strainer surface area before determining the conventional debris head loss . Table 3.f.10-3: Conventional Head Loss Values Fiber Debris Load at Test Scale Head Loss (ft3) (ft-H20) s;3 .1* 0.625

                           >3 .1*                                      5.46 3                                              3
     *The 3.1 ft of fiber at test scale corresponds to 32.04 ft at the plant scale for one RHR strainer.

Chemical Head Loss A conservative chemical head loss model was implemented in the NARWHAL CFP calculation . The head loss effects of calcium phosphate and SAS were each analyzed separately from the 2009 head loss test results. Table 3.f.10-4 shows the head loss applied to each strainer once precipitate starts to accumulate on the strainer. Table 3.f.10-4: Chemical Head Loss Values Quantity Head Loss Chemical Precipitate (lbm) (ft-H20) Calcium Phosphate >O 1.11 Sodium Aluminum Silicate >O 5.24 Note that chemical head loss is not applied until a 0.45-inch thick theoretical uniform fiber debris bed has formed on the strainer. This approach is reasonable because , for fiber quantities smaller than this , large areas of open screen are present on the strainer. This is supported by the 2009 thin bed head loss test data . During the thin bed test, particulate debris was added to the test tank before fiber debris was batched in. Figure 3.f.10-4 shows a negligible increase in debris head loss for fiber loads up to and including 5.96 lbm . This fiber load corresponds to a uniform fiber bed thickness of 0.45 inches. ES-82

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) 0 .3 0£ 0.25

                 '+-

0

                !E.. 0 .2
                  ~

_,0 "C 0.15 ro (I) I

                 .!:   0 .1
                  ~

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0.05 c

0 0.11 0.23 0 .34 0.45 0.57 Theoretical Fiber Bed Thickness (inches) Figure 3.f.10-4: Increase in Head Loss as a Function of Fiber Bed Thickness 11 . State whether the sump is partially submerged or vented (i.e ., lacks a complete water seal over its entire surface) for any accident scenarios , and describe what failure criteria in addition to loss of NPSH margin were applied to address potential inability to pass the required flow through the strainer. Response to 3.f.11: As shown in Table 3.g.1-3 , for some of the postulated breaks (specifically, reactor cavity breaks with CS actuated) , the strainers could be partially submerged at the start of recirculation for a short period but become fully submerged before the switchover to recirculation is completed . When strainers are not fully submerged, the unsubmerged strainer head loss failure criterion discussed in Section 3.f.7 was used. The NARWHAL CFP calculation showed that the head loss during the time when the RHR strainer is partially submerged does not challenge the failure criterion which states that head loss cannot be greater than half of the submerged strainer height for any of the break scenarios . The calculation evaluated the most limiting break in terms of fiber at the RHR strainer during the time that the RHR strainer is partially submerged . The DEGB at Weld 11201-001-1-RB (located in the reactor cavity on the hot leg) with a single train failure configuration resulted in the most amount of fiber on the RHRA strainer for all breaks that have partially submerged strainers during recirculation . For this break, the strainer is partially submerged for a total of 11 minutes. The amount of debris accumulated during this period of time did not challenge the failure criterion for partially submerged strainers. ES-83

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary)

12. State whether near-field settling was credited for the head-loss testing , and if so ,

provide a description of the scaling analysis used to justify near-field credit. Response to 3.f.12: No near-field settling was credited in head loss testing . Sufficient turbulence was provided in the tank to ensure that all debris had an opportunity to collect on the surfaces of the test strainer, while not disturbing the debris bed formation . Additionally, manual stirs were applied as necessary to prevent debris from settling after introduction . Two mechanical stirrers were required to suspend the debris due to the strainer configuration and flow rate , one in the pit below the strainer, and another within the area underneath the strainer, bounded by the simulated containment walls and floo r. A sparger system was installed on the return line to aid in suspension of debris. Additionally, a sump pump and attendant tub ing were used to provide flow from beneath the simulated containment floor to ensure that particulate debris did not accumulate there . Hand-stirring and manual adjustment of the mechanical stirrers was performed as necessary during the add itions of the fibrous and particulate debris , with much care and consideration given to avoid disturbing the bed or otherwise artificially influencing the bed formation .

13. State whether temperature/viscosity was used to scale the results of the head loss test to actual plant conditions. If scaling was used , provide the basis for concluding that boreholes or other differential-pressure induced effects did not affect the morphology of the test debris bed .

Response to 3.f.13 : Head loss values were scaled from test conditions to plant conditions using both temperature (i.e ., viscosity and density as a function of temperature) as well as strainer approach velocity. As shown in the response to 3.f.10, these equations were derived from VEGP-specific flow sweep data . During the 2009 head loss testing , flow sweeps were conducted at the end of each test to characterize the flow through a prototypical debris bed . Therefore , any boreholes and other differential-pressure induced effects on bed morphology were captured and properly accounted for when scaling the head loss . In addition , as stated in the NARWHAL CFP calculation , two sets of flow sweep data were collected following the thin -bed and full debris load tests . To account for the uncertainties in the flow sweeps , the resulting correction parameters from both tests were applied at each time step and the higher resulting head loss was used . E5-84

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary)

14. State whether containment accident pressure was credited in evaluating whether flashing would occur across the strainer surface, and if so , summarize the methodology used to determine the available containment pressure.

Response to 3.f.14: The NARWHAL software was used to evaluate the potential for flashing due to the pressure drop across the strainer and debris bed . For a given break, a flashing failure was recorded if, at any time during sump recirculation , the pressure downstream of the strainer was lower than the vapor pressure at the sump temperature . The pressure downstream of the strainer was calculated by NARWHAL based on the strainer submergence , containment pressure-, and head loss across the strainer. Note that, for flashing analysis , the strainer submergence is evaluated from the top of the strainer. As discussed in Enclosure 3, Section 6.7, up to 3.5 psi of accident pressure was credited in order to preclude flashing. This approach is reasonable , since, as shown below, even the smallest margin in the containment pressure for preventing flashing is higher than the 3.5 psi credited in the analysis. The margin in containment pressure for preventing flashing immediately downstream of the strainer is evaluated for time-dependent post-accident containment and sump conditions . For each given set of conditions , the sump pool temperature is obtained from the design basis profile evaluated for a double-ended reactor coolant pump (RCP) suction break with minimum safeguards. The strainer head loss at each given pool temperature is taken from the NARWHAL CFP calculation . The post-accident containment pressure is from the design basis profile evaluated for a double-ended RCP suction break with maximum safeguards. The minimum containment pressure that is required to prevent flashing is calculated by adding the strainer head loss to the water vapor pressure . Afterwards , this minimum required containment pressure is compared with the expected post-accident containment pressure to determine the margin . The evaluation contains the following conservatisms : When calculating the minimum containment pressure required to prevent flashing , the submergence of the strainer is conservatively neglected . Including the submergence would reduce the minimum pressure required and increase the margin .

• For sump temperatures above 212 degrees F, the strainer head loss is conservatively assumed to be the same as that at 212 degrees F. In reality, the head loss adjusted to the actual temperature would be lower due to the lower water viscosities at higher temperatures.

When determining the post-accident containment pressures from the Vogtle FSAR chart, the values are rounded down, which results in conservatively smaller margins. ES-85

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) As shown in Table 3.f.14-1 , the minimum margin in the containment pressure to prevent flashing is over 6 psi at 3,000 seconds after the accident. Therefore , crediting 3.5 psi of accident pressure in the NARWHAL CFP calculation for the flashing evaluation is reasonable and conservative . Figure 3.f.14-1 compares the design basis post-accident containment pressure with the minimum containment pressure required to prevent flashing . The vertical difference between the two curves represent the margin in containment pressure for preventing flashing . Table 3.f.14-1: Margin in Containment Pressure for Preventing Flashing based on OBA Curves Min Cont. Margin in Sump Pool Vapor Strainer Accident Time Pressure Req'd Containment Temperature Pressure Head Pressure (s) (oF) to Prevent Pressure (psia) Loss (ft) (psia) Flashing (psia) (psi) 1,800 251 30.66 5.515 32.90 39.4 6.5 3,000 249 29.47 5.515 31 .72 37.8 6.1 3,400 248 28.70 5.515 30.95 38.4 7.4 7,020 212 14.81 5.515 17.10 33.4 16.3 7,980 205 12.88 5.544 15.19 33.4 18.2 10,020 195 10.49 5.589 12.83 31.4 18.6 19,980 165 5.42 5.729 7.85 27.4 19.5 30 ,000 153 4.08 9.006 7.92 25.4 17.5 60 ,000 140 2.96 9.128 6.86 22.9 16.0 90 ,060 133 2.47 13.615 8.30 21 .9 13.6 500,460 120 1.74 13.826 7.68 19.4 11 .7 50

                                                                  - Min Containment Pressure to Prevent Flashing 40 +---.c--- - - - - + - - - - 1
                                                                  ---.- Deisgn Basis Accident ro                                                       Containment Pressure
               - ~ 30 + - - -....;:.;;111-----'"""";::--- - - - - - + - - - - - - - I 10 -+-------~'                 ___,,__ _ _ _ _-+------~

0 -+---'--'-~'-'--'-'-'-1---L---'-..__,_-'--'--'-'+-_...._..___._,_......_,__,~ 1.E+3 1.E+4 1.E+S 1.E+6 Time (s) Figure 3.f.14-1: Margin in Containment Pressure for Preventing Flashing based on OBA Curves E5-86

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) To demonstrate that margin also exists for breaks that have a lower pressure than the OBA profile, the same approach was also applied to the best-estimate post-accident containment pressure and sump pool temperature profiles to derive the margins in containment pressure for preventing flashing. Note that the best-estimate curves have lower values than the design basis curves because the thermal hydraulic modeling for the best-estimate cases used less conservative inputs . Figure 3.f.14-2 compares the best-estimate containment pressure curve for a double-ended guillotine cold leg break with the minimum pressure required for preventing flashing evaluated using the corresponding sump pool temperature profile. The vertical difference between the two curves is the margin for preventing flashing . The results show that the minimum margin is 9.3 psi at approximately 30,000 seconds after the accident. Note that, for the cold leg break, sump recirculation starts at 3,398 seconds after the accident. 40 ~---- --~--- 1 ----- Min Pressure to Prevent Flash ing

                                                                                                         ==;-1 i I 30 i--"
                      """""'o::--- -                                       - - - Best Estimate CLB          Y i
                                     ---+-------1
                    ~                                                    L         Containment Press ~

1

            ~ 20   r-------1-__::=--...:;;;;;:::::::::::::j:::::::=-----1 Ill Ill
            <ll 0...

10 L ............... !.... - .... ___ .. ,~- ------...... _.,-.. t I 0 ....._._"'"" 1.E+2 1.E+3 1.E+4 1.E+5 lime (s) Figure 3.f.14-2: Margin in Containment Pressure for Preventing Flashing based on Best-Estimate Cold Leg Break Curves Similar evaluations were also performed using the best-estimate hot leg break curves. The results are shown in Figure 3.f.14-3. The minimum margin is 8.7 psi at approximately 2,500 seconds after the accident. Note that, for the hot leg break, sump recirculation starts at 2,263 seconds after the accident. ES-87

Enclosure 5 Supplemen'tal Response to NRC Generic Letter 2004-02 (Nori-Proprietary)

• Min Pressure to Prevent Flashing

                                                 -       Best Estimate HLB Contai nment Pressure 0  +-~~~~~~-'-+--~~~~~~--'--'-<>----~~~~~~~

1.E+2 1.E+3 1.E+4 1.E+5 Time (s) Figure 3.f.14-3: Margin in Containment Pressure for Preventing Flashing based on Best-Estimate Hot Leg Break Curves In summary, for the best-estimate containment pressure and sump temperature curves , the minimum margin in containment pressure for preventing flashing is at least 8.7 psi. Therefore , crediting 3.5 psi of accident pressure in the NARWHAL CFP calculation for the flashing evaluation is reasonable and conservative. Note that containment pressure and sump temperature are intrinsically related. While the best-estimate containment pressures are lower than the design basis case , the corresponding pool temperatures are also lower, which results in lower pressures required for preventing flashing. The resulting margins in containment pressure for the best-estimate curves are either comparable or actually greater than those derived based on the design basis curves . ES-88

Enclosure 5 Supplemental Response to NRC Generic Letter 2004;02 (Non-Proprietary)

g. Net Positive Suction Head The objective of the NPSH section is to calculate the NPSH margin for the ECCS and CSS pumps that would exist during a LOCA considering a spectrum of break sizes.

1. Provide applicable pump flow rates, the total recirculation sump flow rates, sump temperature(s), and minimum containment water level.

Response to 3.g.1: Pump/ Sump Flow Rates ECCS and CS pump design flow rates used in the NARWHAL model are presented in Table 3.g.1-1. The total recirculation sump flow rates are provided in Table 3.g.1-

2. Note that the SI pumps and CCPs piggyback off of the RHR system during recirculation. Additionally, each of the RHR and CS pumps has its own dedicated sump and strainer.

Table 3.g.1-1: Applicable Pump Flow Rates Pump Design Flow Rate (gpm) RHR 3,700 SI 425 cc 150 cs 2,600 Table 3.g.1-2: Total Recirculation Sump Flow Rates Sump Design Flow Rate (gpm) RHR 3,700 cs 2,600 Minimum Water Level Minimum sump pool levels were calculated in the NARWHAL CFP calculation and in a bounding hand calculation. The NARWHAL calculation performs comprehensive evaluation of GSl-191 phenomena in a self-consistent and time-dependent manner. For each accident evaluated, the entire duration of RWST injection and sump recirculation was divided into smaller time steps. The minimum sump pool volume was calculated for each time step by subtracting the transitory and geometric hold up volumes from the total quantity of water in containment. The NARWHAL CFP calculation evaluated the NPSH margin for each break scenario. Impact on the results due to variabilities in the inputs was evaluated by sensitivity analyses. ES-89

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) The minimum water level hand calculation evaluated bounding minimum sump pool volumes and levels which were used as inputs in the vortexing evaluation (see the Response to 3.f.3) and chemical precipitate debris hand calculation (see the Response to 3.o.1 ). Table 3.g.1-3 summarizes the results of the minimum water level hand calculation. The short-term water level values (prior to 60 hours post-LOCA) and long-term water level values (at 60 hours post-LOCA) differ due to the transient inputs, such as RWST injection, reactor cavity hold-up, and containment temperature (which influences vapor hold-up). The submergence values in Table 3.g.1-3 were calculated by subtracting the RHR strainer height (4.438 ft, discussed in Response to 3.f.2) from the water level above the containment floor. A negative value for strainer submergence indicates the strainer is partially submerged. The submergence values in Table 3.g.1-3 bound the minimum submergence of the CS strainers because the CS strainers are shorter than the RHR strainer and switchover of CS pumps to recirculation occurs after the RHR pumps. The VEGP sump recirculation switchover evaluation showed that, for breaks that do not actuate CS, the ECCS pumps continue drawing all of their flow from the RWST until the Empty level setpoint is reached. In other words, for these breaks, recirculation will not start until the time labeled as "completion of switchover" in Table 3.g.1-3. Therefore, for the breaks that do not actuate CS, no strainer submergence values are shown at the time when the sump suction valves open. For those breaks that do actuate CS, the VEGP sump recirculation switchover evaluation showed that the ECCS pumps start to draw flow from the sump as soon as the sump suction valves open. Therefore, for these breaks, sump recirculation begins when the sump suction valves open at the RWST Low-Low level. ES-90

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Table 3.g.1-3: Minimum Sump Pool Water Levels from Hand Calculation Time of Pool Strainer Break Case Time Occurrence Height Submergence Description (sec) (ft) (ft) 1 Sump Suction Valves Open 1,929 4.536 0.098 LBLOCAwith Completion of Switchover 3,288 6.241 1.803 Containment Spray 5.5 Hours 19,800 6.058 1.620 60 Hours 216,000 5.311 0.873 Sump Suction Valves Open 18, 109 4.108 NIA SBLOCA without Completion of Switchover2 24, 137 5.739 1.301 Containment Spray 5.5 Hours 19,800 6.003 1.565 60 Hours 216,000 5.186 0.748 Sump Suction Valves Open 6,038 4.648 N/A MBLOCA without Completion of Switchover2 8,048 6.353 1.915 Containment Spray 5.5 Hours 19,800 6.318 1.880 60 Hours 216,000 5.501 1.063 Sump Suction Valves Open 3,655 4.692 N/A LBLOCA without Completion of Switchover2 5,104 6.407 1.969 Containment Spray 5.5 Hours 19,800 6.318 1.880 60 Hours 216,000 5.501 1.063 Sump Suction Valves Open 1 1,929 3.054 -1.384 Reactor Nozzle Break LBLOCA Completion of Switchover 3,288 4.788 0.350 with Containment 5.5 Hours 19,800 4.971 0.533 Spray 60 Hours 216,000 5.039 0.601 Sump Suction Valves Open 3,655 3.235 N/A Reactor Nozzle Break LBLOCA Completion of Switchover2 5,104 4.977 0.539 without Containment 5.5 Hours 19,800 5.161 0.723 Spray 60 Hours 216,000 5.229 0.791 Notes: 1 Beginning of recirculation for the breaks that actuate CS is when the RWST level reaches Low-Low setpoint and the sump suction valves for the RHR pumps open. 2 Beginning of recirculation for the breaks that do not actuate CS is when the switchover of the RHR pump suctions from the RWST to sump is completed. E5-91

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 {Non-Proprietary) Sump Temperature The VEGP NARWHAL CFP calculation used the design-basis sump temperature profile calculated for a double-ended pump suction LOCA with minimum safeguards. Note that the minimum safeguards temperature profile shown in Figure 3.g.1-1 is conservatively higher than the temperature profile for the maximum safeguards case. 300

  -u. 250 L

100 . *-*-* *~**- "" - . **-

                                                ~      ',,,_._  ......,.... *-*-***- --

1

                                                                      '~                     ~

50 '  ! I 1.E-01 1.E+01 1.E+03 1.E+05 1.E+07 Time (s) Figure 3.g.1-1: Sump Temperature for Double-Ended Pump Suction Break with Minimum Safeguards As discussed above, the recirculation duration was divided into smaller time steps. When applying the sump temperature profile, the value that is closest to the current time step is used. Consider an example where the current time-step is 220 seconds and the profile has values corresponding to 219 and 229 seconds. NARWHAL would return the value at 219 seconds because it is closer to the current time step.

2. Describe the assumptions used in the calculations for the above parameters and the sources/bases of the assumptions.

Response to 3.g.2: Pump/Sump Flow Rate As discussed in the VEGP NARWHAL CFP calculation, the RHR flow rate was assumed to be 3, 700 gpm. The design flow rate for the RHR pumps is 3,000 gpm. Using a higher flow rate is generally conservative in terms of recirculation timing, ES-92

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) flashing calculations, and head loss correction as performed in the NARWHAL CFP calculation. This 3,700-gpm flow rate is also consistent with the value used in the design-basis NPSH calculation and the single train value used in the ECCS system head curve. In the NARWHAL CFP calculation, the flow rates for the SI pumps and CCPs are their design flow rates based on the SI system description. Similarly, the CS pump flow rate is also the design flow rate from the CS system description. Using the design flow rates for these pumps is reasonable because they are more closely aligned to what would be expected in post-LOCA mitigation than run-out flow rates when taking into consideration system pipe losses. In the NARWHAL CFP calculation, the same pump flow rates were used consistently for all breaks regardless of break size. Using higher flow rates for the smaller breaks is conservative in terms of NPSH and flashing failures because recirculation starts sooner when the pool temperature is higher. Minimum Water Level As stated in the response to 3.g.1, minimum sump pool water levels were calculated in both the NARWHAL CFP calculation and a hand calculation. The major assumptions used in these evaluations are listed as follows.

1. The density of the inventory of the RWST, the reactor coolant system (RCS), and the SI accumulators is assumed to be the same as pure water. This is a reasonable assumption because the concentration of boric acid in the water is extremely small, with a maximum of 1,900 ppm for the RCS; 2,600 ppm for the RWST; and 2,600 ppm for the accumulators.

2. It is assumed that SBLOCAs will not result in rapid, full depressurization of the RCS; therefore, the SI accumulators will not inject when evaluating the minimum water levels for SBLOCAs. This is a conservative assumption because this will minimize the pool volume.

3. It is assumed that MBLOCAs and LBLOCAs will result in full depressurization of the RCS; therefore, during recirculation, the RCS will retain water up to the elevation of the break. This is a reasonable assumption because these breaks result in rapid cooling from the SI accumulators, which are triggered through RCS depressu rization.

4. The hand calculation reported minimum sump water levels for three break size categories, defined as follows. This definition matches that used in the Vogtle PRA model.

a. An SBLOCA is defined as a break smaller than 2 inches.

b. An MBLOCA is defined as a break greater than or equal to 2 inches, less than 6 inches.

c. An LBLOCA is defined as a break greater than or equal to 6 inches with the largest break being a double-ended guillotine break of the crossover leg.

ES-93

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary)

5. It is assumed that a reactor nozzle break will cause the entire reactor cavity (up to the seal ring) to fill with water before any water reaches the containment floor.

This conservatively maximizes the transient reactor cavity hold-up, thereby minimizing the pool level.

6. By maximizing the vapor hold-up in the atmosphere of containment, water is withheld from the pool, thereby conservatively minimizing the pool level. To

   'maximize the vapor hold-up in the atmosphere of containment, three complementary assumptions were made.

a. The relative humidity within containment post-LOCA was maximized. A post-LOCA relative humidity of 100 percent was used to saturate the atmosphere in containment completely, thereby maximizing the change in water vapor in the air from pre-LOCA to post-LOCA conditions.

b. The relative humidity within containment pre-LOCA was minimized. A pre-LOCA relative humidity of 0 percent was used to obtain a pre-LOCA vapor hold-up of 0 gal, thereby maximizing the change in water vapor in the air from pre-LOCA to post-LOCA conditions.

c. The transient values used for containment temperature are maximum values, which result in maximum vapor pressures. These maximum vapor pressures maximize the vapor hold-up in the air.

7. The containment sprays were assumed to only be activated for hot leg breaks greater than 15 inches, which includes all partial breaks and DEGBs greater than 15 inches on the hot legs. However, no failures on the cold or intermediate legs were assumed to actuate containment sprays. This assumption is consistent with the results of best-estimate thermal-hydraulic modeling for a range of potential break sizes on the hot and cold leg piping. This modeling showed that a hot leg DEGB resulted in containment pressures exceeding the CS actuation setpoint of 21.5 psig, while all other evaluated breaks (including a cold leg DEGB and partial 15 inch breaks on both the hot and cold legs) did not. Assuming that hot leg breaks greater than 15 inches activate CS is reasonable because it represents what was learned from the best-estimate thermal hydraulic modeling.

It is recognized that there is some uncertainty in which breaks initiate CS. Sensitivity runs were therefore performed on actuation limits and spray duration using NARWHAL, as summarized in Enclosure 3.

8. The accumulators were assumed to not inject for any secondary side break. This is a reasonable assumption because secondary side breaks do not result in rapid depressurization of the RCS, which would trigger accumulator injection.

Sump Temperature The sump temperature profile used for the GSl-191 analysis was from the design-basis containment analysis for a DEGB on the crossover leg with minimum safeguards and 11.12 percent fan cooler degradation. This analysis was performed for evaluating post-LOCA containment integrity to support the VEGP Units 1 and 2 measurement uncertainty recapture power uprate program. Note that, in addition to the crossover leg break, the containment analysis also modeled the containment response following a main steam line break (MSLB). The results showed that the ES-94

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) crossover leg break with minimum safeguards resulted in higher sump temperatures than the MSLB and the crossover leg break with maximum safeguards.

3. Provide the basis for the required NPSH values, e.g., 3 percent head drop or other criterion.

Response to 3.g.3: The NPSH required (NPSHR) values were taken from the bounding pump vendor curves. These curves were obtained by the pump manufacturer through testing in accordance with the Hydraulic Institute guidelines in effect at the time. Typically, the 3 percent head drop criterion was used in pump NPSH testing.

4. Describe how friction and other flow losses are accounted for.

Response to 3.g.4: The verification of adequate NPSH margin to the RHR and CS pumps from the containment sump was performed using the NARWHAL model. For each time step, NARWHAL calculates the pump NPSH available (NPSHA), NPSHR, and strainer head loss using the inputs of that time step (e.g., sump water level, sump temperature, and pump flow rates). Note that the calculated NPSHA accounted for the piping head loss from the sump to pump suction but not the strainer head loss. If the NPSH margin, determined by subtracting NPSHR from NPSHA, is less than the total strainer head loss, a failure is recorded. The total strainer head loss was calculated by combining the clean strainer and debris bed head losses, and extrapolation constant as necessary. The head loss of the suction piping between the strainer exit and the pump suction was accounted for when calculating NPSHA. The piping frictional loss was calculated using the standard Darcy formula with the friction factor determined from an empirical equation. The head losses of the components (e.g., valves, elbows, reducers, and tee junctions) on the pump suction piping were calculated using the loss coefficients from standard industry handbooks.

5. Describe the system response scenarios for LBLOCAs and SBLOCAs.

Response to 3.g.5: In response to a LOCA, the RHR pumps, SI pumps, and CCPs automatically start upon receipt of an SI signal. These pumps take suction from the RWST and inject to the RCS cold legs. This system line-up is referred to as the ECCS injection phase. The CS pumps start automatically when the containment pressure reaches the high-pressure setpoint for CS actuation. The CS pumps also take suction from the RWST during the injection phase. When the RCS depressurizes to approximately 600 psia, all four accumulators begin to inject borated water into the RCS loops. E5-95

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Before the RWST inventory is depleted, the suction source of the pumps must be switched to the recirculation sumps. The surrip suction valves for the ECCS pumps open automatically when the RWST level reaches the Low-Low setpoint. The switchover for the CS pumps starts manually when the RWST level reaches the Empty setpoint. The switchover is complete when the suction valves from the RWST for all pumps are manually closed, which occurs between the RWST Empty and Dead Volume levels. For the breaks that do not actuate CS, the switchover to sump recirculation for the ECCS pumps follows the same logic. Approximately 7.5 hours following an accident, the ECCS line-up is modified for simultaneous cold and hot leg recirculation. For this operating mode, the SI pumps and CCPs continue taking suction from th.e RHR pump discharge. The RHR and SI pumps are aligned to supply flow to the RCS hot legs, but the CCPs continue supplying flow to the cold legs. The response sequence described above is typical for the ECCS and CSS following a LOCA. The differences between the responses to an LBLOCA and an SBLOCA are:

• Depending on the size of the break, the RCS pressure may stabilize at a value that does not allow injection from the SI accumulators and/or the RHR pumps.
• For an SBLOCA, the containment pressure will likely remain below the actuation setpoint for the CSS.

For an SBLOCA, the outflow from the RWST may be sufficiently low that the plant may be taken to a safe shutdown condition before the RWST level reaches the Low-Low setpoint. As a result, sump recirculation may not be required.

6. Describe the operational status for each ECCS and CSS pump before and after the initiation of recirculation.

Response to 3.g.6: Residual Heat Removal Pumps In the event of a LOCA, both RHR pumps are started automatically on receipt of an SI signal. During the injection phase, the RHR pumps take suction from the RWST and supply flow to the RCS cold legs. When the RWST level reaches the Low-Low setpoint, the suction valves to the sump automatically open. The RHR pumps could take suction simultaneously from the RWST and the containment sumps. After the RWST level reaches the Empty setpoint, the suction valves to the RWST are manually closed. Afterwards, the RHR pumps take suction from the sumps only. The RHR pumps continue to supply flow to the RCS cold legs and to the SI pumps and CCPs. ES-96

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Centrifugal Charging Pumps In the event of a LOCA, both CCPs start automatically on receipt of an SI signal and take suction directly from the RWST during the injection phase. The CCPs supply flow to the RCS cold legs. After switching to the sump recirculation phase, flow to the CCPs is provided from the RHR pump discharge. The CCPs continue supplying flow to the RCS cold legs during simultaneous cold leg and hot leg recirculation. Safety Injection Pumps In the event of a LOCA, both SI pumps start automatically on receipt of an SI signal. During the injection phase, these pumps take suction from the RWST and deliver water to the RCS cold leg. Similar to the CCPs, flow to the SI pumps is supplied from the containment emergency sump via the RHR pumps during the recirculation phase. Containment Spray System Pumps The CS pumps can be actuated manually from the control room or automatically on receipt of two out of four containment pressure (high-3) signals. These signals start the CS pumps and open the discharge valves to the spray headers. During the injection phase, the CS pumps take suction from the RWST. As discussed in the Response to 3.g.5, the pump suction is manually switched to the containment recirculation sump when the RWST level reaches the Empty setpoint.

7. Describe the single failure assumptions relevant to pump operation and sump performance.

Response to 3.g. 7: As described in Enclosure 3, Sections 4.0, 6.3, and 14.1, the VEGP risk-informed evaluation considered many different equipment configurations and wasn't limited to the worst single failure. The high likelihood configuration calculation used the VEGP PRA model of record, which accounts for human reliability analysis (HRA), and identified the following twelve equipment failure combinations.

1. No Equipment failure

2. RHR Pump 8 failure

3. RHR Pump A failure

4. Charging Pump A failure

5. Charging Pump 8 failure

6. SI Pump 8 failure

7. SI Pump A failure

8. Train 8 failure

9. Train A failure

10. CS Pump 8 failure ES-97

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary)

11. CS Pump A failure

12. Both CS Pumps failure Note that the VEGP inputs and NARWHAL methodology allow for this list to be reduced based on train symmetry. A pump failure for one train is analytically identical to a pump failure for the other train. Therefore, the following seven equipment configurations were analyzed in the NARWHAL CFP calculation.

1. No Equipment failure

2. RHR Pump B failure

3. Charging Pump B failure

4. SI Pump B failure

5. Train B failure

6. CS Pump B failure

7. Both CS Pumps failure It was assumed that all random equipment failures evaluated occur at the beginning of recirculation. This is a conservative assumption because it results in a quicker switchover to recirculation when compared to failure at the beginning of the event.

Additionally, for CS pump and/or RHR pump failure cases, it results in more debris accumulation on the remaining active strainers. The CCP B and SI pump B failure cases are identical to the no equipment failure case. This is because the failure is applied at the start of recirculation. The flow rate through the RHR strainers is not affected by the charging pump failure because the RHR pump provides the same flow rate regardless of which piggybacked pump fails. The CS pump B failure case is similar to the no equipment failure case. It only affects hot leg breaks greater than 15 inches. Thus, this case only has a slight effect on the results even though there is one less active strainer during recirculation.

8. Describe how the containment sump water level is determined.

Response to 3.g.8: As discussed in the response to 3.g.1, the post-LOCA minimum sump pool level was determined in both the NARWHAL CFP calculation and a hand calculation. The two calculations used the methodology described below:

1. A correlation was first developed for the relationship between the containment water level and the water volume using a 3-D CAD model.

2. The quantity of water added to containment from the RWST, RCS, and SI accumulators was calculated.

3. The quantity of water that is diverted from the containment sump by the following effects was evaluated:

• Hold-up within the reactor cavity and in-core tunnel.
• RCS hold-up volume required to fill the RCS steam space.

ES-98

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary)

• Water volume required to fill the CS pump discharge piping that is empty pre-LOCA.
• Water in transit from the containment spray nozzles and the break to the containment sump.
• Steam hold-up in the containment atmosphere.
• Miscellaneous hold-up volumes throughout containment, such as containment sumps, the elevator pit, and containment floor drains.

4. Given the net mass of water added to the containment floor based on Items 2 and 3 listed above, the post-LOCA containment water level is calculated using the correlation developed in Item 1.

As discussed earlier, the NARWHAL CFP calculation used self-consistent inputs and evaluated time-dependent pool volumes and water levels for each postulated break. The hand calculation determined bounding minimum containment water levels for LBLOCA, MBLOCA, and SBLOCA and provided inputs for evaluating chemical precipitate debris quantities and vortexing. While the NARWHAL CFP calculation determines the water level at each time step within the simulation, the hand calculation only reported water levels at a few different times after the accident, as shown in Table 3.g.1-3.

9. Provide assumptions that are included in the analysis to ensure a minimum (conservative) water level in determining NPSH margin.

Response to 3.g.9: The assumptions provided in the Response to 3.g.2 ensure that minimum (conservative) containment water levels are calculated in the VEGP containment water volume calculation.

10. Describe whether and how the following volumes have been accounted for in pool level calculations: empty spray pipe, water droplets, condensation, and holdup on horizontal and vertical surfaces. If any are not accounted for, explain why.

Response to 3.g.10: As described in the Response to 3.g.8, the following volumes are treated within the VEGP containment water volume calculation as hold-up volumes that remove water from the containment pool: CS discharge piping (initially empty spray piping), water in transit from both the containment spray nozzles and the break itself, and the water droplets on containment walls. ES-99

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary)

11. Provide assumptions (and their bases) as to what equipment will displace water resulting in higher pool level.

Response to 3.g.11: The volumes occupied by structures, equipment, and equipment supports, etc. will displace water and result in a higher pool level. Examples of such equipment and or structures include concrete walls, accumulator tanks, piping, and cable trays. These volumes were accounted for in the VEGP containment water volume calculation. The 3D CAD model of containment was used to determine the correlation between the containment pool volume and water level. Smaller equipment, cables, and instruments are excluded from the CAD model and therefore provide some conservatism in the resulting water levels.

12. Provide assumptions (and their bases) as to what water sources provide pool volume and how much volume is from each source.

Response to 3.g.12: The following design inputs provided the basis for water sources and their volumes to determine the minimum containment water level for VEGP:

• The VEGP TS minimum initial RWST level was used for the initial RWST water level. As discussed in 3.g.1, when evaluating the minimum containment water level, the RWST level at the beginning of sump recirculation is either at the Low-Low level (minimum volume of water injected from the RWST at this level is 435,522 gal) or at the Empty level (the minimum volume of water injected from the RWST at this level is 580,497 gal).
• Four SI accumulators have a minimum volume of 6,555 gal/accumulator. The total minimum volume of the SI accumulators is therefore 26,220 gal. This volume is not credited in the SBLOCA cases because the RCS pressure is assumed to remain above their injection pressure as stated in Response to 3.g.2.
• The inventory of the RCS is assumed to remain relatively constant during normal operations. This is a reasonable assumption because during full power operation, the RCS remains at a fixed volume and remains at constant temperature and pressure. Due to the small volume of the RCS as compared to the RWST and its negligible variation in water volume, a best estimate value is representative. The best estimate RCS liquid volume is that associated with the total RCS liquid volume at hot full power conditions: 86,729 gal. The RCS represents both a source of water and a hold-up volume. The mass of water held up in the RCS may be more or less than the initial RCS mass depending on the elevation of the break (e.g., for a break at the top of the pressurizer, the vapor space of the pressurizer would be filled).

ES-100

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary)

13. If credit is taken for containment accident pressure in determining available NPSH, provide description of the calculation of containment accident pressure used in determining the available NPSH.

Response to 3.g.13: Containment accident pressure was not credited in the VEGP analysis for pump NPSH. Using the VEGP NARWHAL model, pump NPSH margin was calculated at each time step using inputs from that time step. The containment pressure is assumed to be equal to the saturation pressure at the sump temperature for sump temperatures greater than 210.96 degrees F. Note that the temperature of 210.96 degrees F corresponds to the saturation temperature at the VEGP TS minimum containment pressure of -0.3 psig. For sump temperatures below 210.96 degrees F, the minimum containment pressure of-0.3 psig (or 14.396 psia) was used as the containment pressure to calculate the pump NPSHA.

14. Provide assumptions made which minimize the containment accident pressure and maximize the sump water temperature.

Response to 3.g.14: Containment Pressure As discussed in the Response to 3.g.13, the VEGP TS minimum containment pressure is -0.3 psig, which corresponds to a saturation temperature of 210.96 degrees F. When calculating pump NPSH margin, the containment pressure was minimized by using the minimum containment pressure of -0.3 psig for sump temperatures below 210.96 degrees F. When the sump temperature is higher than 210.96 degrees F, the containment pressure was assumed to be equal to the saturation pressure at the sump temperature, which is necessary to maintain the sump in a liquid phase. No accident pressure was credited for NPSH calculations. As stated in the NARWHAL CFP calculation, an accident pressure of 3.5 psi was used when evaluating flashing and degasification. However, a sensitivity case was run to show that it is unnecessary to credit accident pressure for degasification analysis. The sensitivity case without crediting any accident pressure resulted in

• degasification failures for 12 large hot leg breaks (greater than 28.5 inches). These 12 breaks already failed in the base case due to exceeding the debris limits failure criterion. As a result, the additional degasification failures without crediting any accident pressure will not increase the overall risk.

Sump Temperature As discussed in the Response to 3.g.1, the VEGP NARWHAL model used the design-basis sump temperature profile calculated for a double-ended pump suction LOCA with minimum safeguards. Note that the minimum safeguards temperature n ES-101

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) profile is conservatively higher than the temperature profile for the maximum safeguards case during recirculation through the sump strainers. The time-dependent sump temperature profile applied to all breaks is included as Figure 3.g.1-1.

15. Specify whether the containment accident pressure is set at the vapor pressure corresponding to the sump liquid temperature.

Response to 3.g.15: See the Responses to 3.g.13 and 3.g.14.

16. Provide the NPSH margin results for pumps taking suction from the sump in recirculation mode.

Response to 3.g.16: The RHR and CS pump NPSH margins were evaluated using the VEGP NARWHAL model. Table 3.g.16-1 provides a summary of the minimum NPSH margins for the RHR pumps in recirculation mode at various sump temperatures between 120 degrees F and 212 degrees F. The RHR pump NPSH margins shown in Table 3.g.16-1 are based on one of the four breaks within Table 3.b.4-2, all of which have the same NPSH margins due to the rule-based approach used in calculating head loss. ES-102

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Table 3.g.16-1 Limiting NPSH Margin vs. Sump Temperature NPSH Margin Net NPSH Margin Pool Strainer Before Subtracting After Subtracting Temperature Head Loss (oF) Strainer Head Strainer Head (ft-H20) Loss (ft-H20) Loss (ft-H20) 212 22.681 5.515 8 17.2 205 27.215 5.544 8 21.7 195 33.572 5.589 8 28.0 165 45.300 5.7298 39.6 153 46.599 9.006b 37.6 140 48.265 9.128b 39.1 133 46.810 13.615c 33.2 120 48.397 13.826c 34.6 a This includes clean strainer, conventional and chemical debris (calcium phosphate) head losses. b This includes clean strainer, conventional and chemical debris (calcium phosphate) head losses, and extrapolation constant. c This includes clean strainer, conventional and chemical debris (calcium phosphate and SAS) head losses and extrapolation constant. Although the minimum net NPSH margins shown in Table 3.g.16-1 are for RHR Pump A, they are bounding for all of the RHR and CS pumps. The RHR pumps are expected to have less NPSH margins than the CS pumps because of the higher head losses of the RHR strainers associated with the higher RHR pump flow rate and greater strainer debris loads. An NPSH evaluation was not performed for the SI pumps and CCPs because these pumps take suction from the RHR pumps during recirculation. Table 3.g.16-1 shows the NPSH margins before and after subtracting the total strainer head losses. The total strainer head losses include the clean strainer head loss, conventional debris (particulate and fiber) head loss, and chemical debris (calcium phosphate and SAS) head loss, as appropriate. Bounding head loss values, as shown in the Response to 3.f.10, were used in the evaluation. The head losses were also extrapolated to the end of the 30-day mission time as described in the Response to 3.f.10. As shown in the table, the minimum net NPSH margin for any given sump temperatures is over 17 ft. Therefore, adequate NPSH margin is available for Unit 1 and Unit 2 RHR and CS pumps to ensure their design functions. E5-103

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary)

h. Coatings Evaluation The objective of the coatings evaluation section is to determine the plant-specific ZOI and debris characteristics for coatings for use in determining the eventual contribution of coatings to overall head loss at the sump screen.

1., Provide a summary of type(s) of coating systems used in containment, e.g., Carboline CZ 11 Inorganic Zinc primer, Ameren 90 epoxy finish coat. Response to 3.h.1: The types of coating and systems used in containment are presented in Table 3.h.1-1. Qualified Coatings Table 3.h.1 Coatings Systems Used in Analyses OFT Density Substrate Layer Type (mil) (lbm/ft3) 1st Coat Carbozinc 11 5 208 2nd Coat Ameren 90 6 99.6 Steel Surfaces 3rd Coat Ameren 90 6 99.6 Total 17 1st Coat K&L 4129 1.5 69.0 2nd Coat K&L 4000 25 107.2 Concrete Surfaces 3rd Coat K&L D-Series 9 98.0 Total 35.5 Unqualified Coatings Unqualified coatings could include coatings within containment that do not have a specified preparation, application, or inspection compliant with plant specifications; previously qualified coatings that have noticeably deteriorated; coatings inaccessible for inspection; and coatings applied by vendors on vendor-supplied items that cannot be qualified. There are several types of unqualified coatings applied over numerous substrates within containment, including various types of epoxy, inorganic zinc, and alkyds. ES-104

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary)

2. Describe and provide bases for assumptions made in post-LOCA paint debris transport analysis.

Response to 3.h.2: The following assumptions related to coatings were made in the NARWHAL model:

• It was assumed that 100 percent of unqualified coatings were in the containment pool at the start of recirculation. This is a conservative assumption since no credit is taken for retention of unqualified coatings in upper containment regardless of the failure time or if containment sprays are initiated.
• It was assumed that the unqualified and degraded qualified coatings in VEGP have a recirculation transport fraction of 100%. This is consistent with the debris transport calculation, and is conservative since settling of this debris is not credited.

3. Discuss suction strainer head loss testing performed as it relates to both qualified and unqualified coatings. Identify surrogate material and what surrogate material was used to simulate coatings debris.

Response to 3.h.3: Silicon carbide was used to simulate both qualified and unqualified coatings debris. See the Response to 3.f.4 for detailed information on coating surrogates and the amount added to the test.

4. Provide bases for the choice of surrogates.

Response to 3.h.4: See the Response to 3.f.4. ES-105

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary)

5. Describe and provide bases for coatings debris generation assumptions. For example, describe how the quantity of paint debris was determined based on ZOI size for qualified and unqualified coatings.

Response to 3.h.5: The following assumptions related to coatings were made in the debris generation calculation:

• Qualified coatings within the ZOls were assumed to fail as 1Oµm diameter spheres; qualified coatings outside the ZOls were assumed to remain intact.

This is based on the guidance of NEI 04-07.

• It was assumed that the FN-8 qualified coatings system was applied to steel structures, including columns, equipment supports and grating. The FN-14/19 system was applied to all concrete surfaces within containment. Using these two systems is conservative because they have the largest number of coats and the largest final dry film thickness of all field coating systems present within containment for their respective substrates. Both field coating systems, including the type, dry-film thickness, and density are presented in Table 3.h.1-1.

The masses of unqualified coatings in containment are quantified based on detailed logs maintained over the life of the plant. The entire quantity of unqualified coatings, as shown in Tables 3.h.5-1 and 3.h.5-2, are assumed to fail for all breaks. The amount of coating debris generated at VEGP is shown Tables 3.h.5-1 and 3.h.5-2. Table 3.h.5-1: Coatings Debris for the Four Overall Worst-Case Breaks Break Location 11201-004-6-RB 11201-001-5-RB 11201-001-3-RB 11201-004-4-RB Break Size 29" 29" 29" 29" Break Type DEGB DEGB DEGB DEGB Qualified Epoxy 50.4 50.2 50.2 47.5 (lbm) Qualified IOZ 43.8 43.7 43.6 41.3 (lbm) Unqualified 2728.7 2728.7 2728.7 2728.7 Epoxy (lbm) Unqualified 58.9 58.9 58.9 58.9 Alkyd (lbm) Unqualified IOZ 55.7 55.7 55.7 55.7 (lbm) ES-106

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Table 3.h.5-2: Coatings Debris for the Four Worst-Case Breaks that Do Not Fail the Strainer Acceptance Criteria Break Location 11201-004-4-RB 11201-001-3-RB 11201-003-5-RB 11201-002-5-RB Break Size 20" 23" 25" 17" Break Type Partial Partial Partial Partial Qualified Epoxy 6.7 8.8 15.6 6.3 (lbm) Qualified IOZ 5.8 7.7 13.6 5.5 (lbm) Unqualified 2728.7 2728.7 ' 2728.7 2728.7 Epoxy (lbm) Unqualified 58.9 58.9 58.9 58.9 Alkyd (lbm) Unqualified IOZ 55.7 55.7 55.7 55.7 (lbm)

6. Describe what debris characteristics were assumed, i.e., chips, particulate, size, distribution, and provide bases for the assumptions.

Response to 3.h.6: In accordance with the guidance provided in NEI 04-07 (Reference 2) and the associated NRC SE (Reference 3), all coating debris was treated as particulate and therefore transported entirely to the sump strainer. See the Response to 3.h.1, 3.h.2, and 3.h.5 for additional description of debris characteristics.

7. Describe any ongoing containment coating conditions assessment program.

Response to 3.h.7: SNC conducts condition assessments of coatings inside containment every outage under the site work control system. As localized areas of degraded coatings are identified, those areas are evaluated and scheduled for repair or replacement as necessary. The periodic condition assessments and resulting repair and replacement activities assure that the amount of coatings that may be susceptible to detachment from the substrate during a LOCA event is minimized. ES-107

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary)

i. Debris Source Term The objective of the debris source term section is to identify any significant design and operational measures taken to control or reduce the plant debris source term to prevent potential adverse effects on the ECCS and CSS recirculation functions.

Provide the information requested in GL 2004-02 Requested Information Item 2(f) regarding programmatic controls taken to limit debris sources in containment. GL 2004-02 Requested Information Item 2(f) A description of the existing or planned programmatic controls that will ensure that potential sources of debris introduced into containment (e.g., insulations, signs, coatings, and foreign materials) will be assessed for potential adverse effects on the ECCS and CSS recirculation functions. Addressees may reference their responses to GL 98-04, "Potential for Degradation of the Emergency Core Cooling System and the Containment Spray System after a Loss-of-Coolant Accident Because of Construction and Protective Coating Deficiencies and Foreign Material in Containment," to the extent that their responses address these specific foreign material control issues. In responding to GL2004-02 Requested Information Item 2(f), provide the following:

1. A summary of the containment housekeeping programmatic controls in place to control or reduce the latent debris burden. Specifically for RMI/low-fiber plants, provide a description of programmatic controls to maintain the latent debris fiber source term into the future to ensure assumptions and conclusions regarding inability to form a thin bed of fibrous debris remain valid.

Response to 3.i.1: SNC procedure, "Containment Exit Inspection," provides detailed guidance for containment inspection to ensure no loose debris (e.g., rags, trash, clothing, etc.) is present in the containment that could be transported to the containment sump and cause restriction of pump suctions during LOCA conditions. This procedure contains an extensive checklist detailing all areas of containment that must be inspected for cleanliness prior to plant startup after each outage. SNC procedure, "Containment Entry," establishes guidance to inventory and control items carried into containment during non-outage entries. This procedure ensures that no loose debris (e.g., rags, trash, clothing, etc.) is present in the containment, which could be transported to the containment sump and cause restriction of pump suctions during LOCA conditions. E5-108

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary)

2. A summary of the foreign material exclusion programmatic controls in place to control the introduction of foreign material into the containment.

Response to 3.i.2: SNC procedure, "Foreign Material Exclusion Program," establishes the administrative controls and personnel responsibilities for the foreign material exclusion (FME) program. The procedure describes methods for controlling and accounting for material, tools, parts, and other foreign material to preclude their uncontrolled introduction into an open or breached system during work activities. This procedure also provides guidance for establishing and maintaini,ng system cleanliness, recovering from an intrusion of foreign material, and re-establishing system cleanliness requirements.

3. A description of how permanent plant changes inside containment are programmatically controlled so as to not change the analytical assumptions and numerical inputs of the licensee analyses supporting the conclusion that the reactor plant remains in compliance with 10 CFR 50.46 and related regulatory requirements.

Response to 3.i.3: An enhancement to the screening guidelines and considerations for the design input process, which is part of the design change procedure, has introduced a requirement to review the impact of a proposed change on the documentation that forms the design basis for the response to GL 2004-02. The specific areas that are addressed are:

• Insulation inside containment
• Fire barrier material inside containment
• Coatings inside containment
• Inactive volumes in containment
• Labels inside containment
• Buffer changes (iodine and pH control)
• Structural changes (i.e., choke points) in containment
• Downstream effects (piping components downstream of the ECCS sump strainers)

Inclusion in the design input process ensures all design changes consider these attributes during the design process. ES-109

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary)

4. A description of how maintenance activities including associated temporary changes are assessed and managed in accordance with the Maintenance Rule, 10 CFR 50.65.

Response to 3.i.4: Maintenance activities, including temporary changes, are subject to the provisions of 10 CFR 50.65(a)(4), as well as VEGP TSs. SNC fleet procedures also provide guidance. For instance, the 50.59 review process procedure provides details on maintenance activities and temporary modifications, while the on-line work management process procedure establishes administrative controls for performing on-line maintenance of structures, systems, components (SSCs) to enhance overall plant safety and reliability. Further guidance is also available in the temporary configuration change procedure.

5. If any of the following suggested design and operational refinements given in the guidance report (guidance report, Section 5) and SE (SE, Section 5.1) were used, summarize the application of the refinements.

a. Recent or planned insulation change-outs in the containment which will reduce the debris burden at the sump strainers.

Response to 3.i.5.a: All of the Min-K insulation located inside the steam generator compartments (original ZOI analyzed for GL 2004-02) was removed from VEGP Unit 1 and Unit 2 containments during refueling outage 1R13 (Fall 2006) and refueling outage 2R12 (Spring 2007). There are no known quantities of Min-K in VEGP Unit 1 and Unit 2 containments outside of the secondary shield wall (outer wall of the steam generator compartments). However, Min-K was only used as insulation in penetrations, which are difficult to inspect. This leaves the containment wall as the only ,. possible location of remaining Min-K. As shown in Figure 3.a.1-1, there is only one line outside of the steam generator compartments with analyzed breaks (i.e. welds inside the first isolation valve). If a worst case ZOI of 28.60 is assumed for this 2" line, the ZOI radius would be about 4.8 feet. Using the strainer as a reference dimension (square with each side approximately 5 feet), it is apparent that the ZOI could not reach the containment wall. Therefore, Min-K is not considered when analyzing the sump strainers. ES-110

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary)

b. Any actions taken to modify existing insulation (e.g., jacketing .or banding) to reduce the debris burden at the sump strainer.

Response to 3.i.5.b: This suggested design and operational refinement was not used in the VEGP evaluation.

c. Modifications to equipment or systems conducted to reduce the debris burden at the sump strainers.

Response to 3.i.5.c: This suggested design and operational refinement was not used in the VEGP evaluation.

d. Actions taken to modify or improve the containment coatings program.

Response to 3.i.5.d: No specific actions were taken to modify or improve the containment coatings program; however, enhancements were made to the screening guidelines and considerations for the design input process to ensure that all design changes consider GL 2004-02 attributes during the design process. The specific areas that are addressed are listed in the Response to 3.i.3.

j. Screen Modification Package The objective of the screen modification package section is to provide a basic description of the sump screen modification.

1. Provide a description of the major features of the sump screen design modification.

Response to 3.j.1: The currently installed strainers for RHR and CS consist of four parallel, vertically stacked, modular disk strainer assemblies that are connected to a plenum installed over each sump. Each RHR strainer assembly consists of 18 stacked disks that are 30 inches long by 30 inches wide, and the height of the disk portion of the strainer is

53. 75 inches. The RHR strainer assemblies are 59.6 inches tall, measured from the containment floor. The four RHR strainer assemblies provide approximately 765 ft2 of perforated plate surface area and 179 ft2 of circumscribed surface area per sump.

Each CS strainer assembly consists of 14 stacked disks that are 30 inches long by 30 inches wide, and the height of the disk portion of the strainer is 41. 75 inches. The CS strainer assemblies are 47.6 inches tall measured from the containment floor. Figure 3.j.1-1 below shows a picture of one CS strainer. Each of the four CS ES-111

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) strainer assemblies provides approximately 590 ft2 of perforated plate surface area and 139 ft2 of circumscribed surface area. Subsequent risk-informed analysis has led to the proposed modification of the Unit 1 and Unit 2 RHR sump strainer assemblies. The RHR strainers will be modified to reduce the overall height approximately 6 inches by removing the two top disks per disk stack. The modified RHR strainer assembly will consist of 16 stacked disks, and the disk portion of the strainer is approximately 47.75 inches high (53.75 in. - 6 in. =47.75 in.). As shown in the response to 3.f.2, the overall height of the modified RHR strainer is 53.25 inches, measured from the containment floor to the highest strainer disks. The four RHR modified strainer assemblies provide approximately 677 .6 ft2 of perforated plate surface area and 159 ft2 of circumscribed surface area per sump as calculated below. 4modules (4 sides ) (30 in)(47.75 in)(l ftz) (circumsc:ibed area)) module side = 1 59 ft2 ( (144 in 2 )

• All of the analyses shown in this submittal were performed for the modified strainer configuration. Operating procedures are being revised, in addition to the planned physical modification, to ensure that the RHR strainers are completely submerged for an increased number of postulated LOCA scenarios.

ES-112

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Figure 3.j.1-1 Containment Spray Strainer

2. Provide a list of any modifications , such as reroute of piping and other components ,

relocation of supports , addition of whip restra ints and missile shields, etc., necessitated by the sump strainer modifications. Response to 3.j.2: The following modifications were necessitated by those of the sump strainer:

• Installation of new and replacement of existing ECCS flow orifices to allow new ECCS throttle valve settings.
• Cage assembly vortex suppressors installed in the sumps removed .
• Temperature elements for the Units 1 and 2 RHR sumps replaced and relocated.
• Two conduit interferences at the Unit 2 RHR sump Train A screen rerouted through an area outside of the sump screen envelope.
• Three electrical interferences for the new Unit 2 CS sump Train A screen relocated/rerouted through an area outside of the sump screen envelope .
• The RHR strainers will be reduced in height by the removal of two disks from each stack to ensure full submergence for an increased number of postulated break scenarios as described in the Response to 3.j .1.

ES-113

Enclosure 5 Suppleme"ntal Response to NRC Generic Letter 2004-02 (Non-Proprietary)

k. Sump Structural Analysis The objective of the sump structural analysis section is to verify the structural adequacy of the sump strainer including seismic loads and loads due to differential pressure, missiles, and jet forces.

Provide the information requested in GL2004-02 Requested Information Item 2(d)(vii). GL 2004-02 Requested Information Item 2(d)(vii) Verification that the strength of the trash racks is adequate to protect the debris screens from missiles and other large debris. The submittal should also provide verification that the trash racks and sump screens are capable of withstanding the loads imposed by expanding jets, missiles, the accumulation of debris, and pressure differentials caused by post-LOCA blockage under flow conditions.

1. Summarize the design inputs, design codes, loads, and load combinations utilized for the sump strainer structural analysis.

Response to 3.k.1: Design Codes (1) American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section Ill, Subsection NC and ND, 1989 Edition. (2) American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section Ill, Appendix I, 1989 Edition, Table 1-6.0 for Modulus of Elasticity, Table 1-5.0 for thermal expansion, and Table 1-7.2 for allowable stress (S). ES-114

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Material Properties The Material properties come from the ASME Code and are tabulated in Table 3.k.1-1 below. Table 3.k.1-1 Material Properties Material I Property @Room @ Maximum Water Temperature Temperature (70°F) (250°F) SA-240 Type SS304 (Strainer): E, Elastic modulus, psi 28.3x1 QB 27.45 x1QB Coefficient of thermal expansion, in/in/°F 8.6x1 Q-B 8.995 x1Q-B Poisson's ratio Q.3 Q.3 SA-479 Type SS410 (Tie Rod): E, Elastic modulus, psi 28.3x1 QB 27.45x1QB Coefficient of thermal expansion, in/in/°F 5.9 x1 Q-B 6.1 x1 Q-B Load Combinations Table 3.k.1-2 shows the load combinations specified for the VEGP passive suction strainer design. Table 3.k.1-2 Load Combinations for VEGP Strainer Design Load Combination Strainer Assembly Design W+ Po+ OBE1 LevelB WD + Pd + OBE2 + TEmax + Per LevelD WD + Pd +SSE2 + Per Support Structure Design W +Po+ OBE1 Level B WD + Pd + OBE2 + TEmax LevelD WD +Pd +SSE2 Nomenclature: W = Weight (Dry strainer Assembly Weight) WD = Weight+ Debris Weight + Hydrodynamic Mass (LOCA Event with Strainer in Water) Per = Crush Pressure (During Suction Strainer Operation in Water Post LOCA) Pd = Design Pressure (LOCA Event) +Water Head (Strainer Open System) Po = Design Pressure (Strainer Open System) ES-115

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) OBE1 = Operating Basis Earthquake (Inertia Load in Air) OBE2 = Operating Basis Earthquake (Inertia Load with Strainer in Water - Include Debris Weight+ Hydrodynamic Mass) TE max = Thermal Expansion (Accident Condition) SSE1 = Safe Shutdown Earthquake (Inertia Load with Strainer in Air) SSE2 = Safe Shutdown Earthquake (Inertia Load with Strainer in Water - Include Debris Weight + Hydrodynamic Mass) The seismic loads are based on the lateral and vertical accelerations of the response spectrum according to the first mode of frequency of the strainer assembly in water. The natural frequency checks in the original analysis show that the system is in the rigid range. The design pressure, Po or Pd, has no impact on the system because the strainer is an open system. The hydrodynamic mass and debris weight are distributed evenly and are added to the strainer finite element model by adjusting the density of the material. A combined load table for the strainer component evaluation is summarized in Table 3.k.1-3. For the, design load case, the dry strainer weight (or 1G) is combined with the QBE vertical acceleration for a combined loading of 1.375G vertically. In addition, QBE horizontal acceleration of 0.27G is applied in both X and Y lateral directions. For the Level B load case, the strainer weight in water including debris and hydrodynamic mass (or 1G) is combined with the QBE vertical acceleration for a combined loading of 1.375G vertically. In addition, QBE horizontal acceleration of 0.27G is applied in both X and Y lateral directions as well as crush pressure and thermal loading. For the Level D load case, the strainer weight in water including debris and hydrodynamic mass (or 1G) is combined with SSE vertical acceleration for a combined loading of 1.6G vertically. In addition, SSE horizontal acceleration of 0.4125G is applied in both X and Y lateral directions as well as crush pressure. Table 3.k.1-3 Load Table for the VEGP Strainer Design Strainer Load Combination Inertia Z* Inertia X Inertia Y Per Temp*** Assembly (G) (G) (G) (psi) (oF) Design W+ Po+ OBE1 1.375 0.27 0.27 LevelB WO + Pd + OBE2 + TEmax + Per 1.375 0.27 0.27 4.46** 180 LevelD WO+ Pd +SSE2 +Per 1.6 0.4125 0.4125 4.46**

• Axis orientation: Z Vertical, X and Y Lateral

   ** Equivalent to 10.1 ft of head loss
   *** Stress free temperature is assumed to be 70 °F, t::..T =(250- 70) °F =180 °F E5-116 I

~

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Modal Analysis Modal analyses were performed using the suction strainer finite element models. Modal results were obtained for the dry strainer and for the wet strainer with added debris weight and hydrodynamic mass during LOCA and post-LOCA events. The strainer structural mass and natural frequencies are calculated for the first four modes and are summarized in Table 3.k.1 ~4. Table 3.k.1-4 Replacement Strainer Weight and Frequency Typical RHR Strainer in Air W= 7,150 lbm Mode 1 37.311 Hz Mode 2 37.722 Hz Mode 3 39.240 Hz Mode4 85.327 Hz Typical RHR Strainer in Water WO = 10,655 lbm Mode 1 30.566 Hz Mode2 30.902 Hz Mode 3 32.146 Hz Mode4 69.901 Hz RHR Train B Strainer in Water WO= 11,256 lbm Mode 1 31.421 Hz Mode 2 32.037 Hz Mode 3 33.426 Hz Mode4 68.966 Hz Load Application Loads used in the stress analysis of the strainer models include the weight of the strainer assembly, hydrodynamic mass and debris mass, the crush pressure due to suction strainer operation, and the lateral and vertical inertial accelerations of. Response Spectrum (OBE & SSE) corresponding to the first mode frequency of strainer assembly in water. The crush pressure is applied on the top and bottom surfaces of the disk sets accounting for debris blockage. The weight of the strainer assembly model in water (WD) is the sum of the weight of the strainer assembly in air (W), the debris weight, and the hydrodynamic mass. The debris and hydrodynamic mass are uniformly distributed over the strainer assembly and support for mode shape and stress analysis. The crush pressure is applied on the plenum for Level D load case. The ASME code combination stress limits are summarized in Tables 3.k.1-5 and 3.k.1-6. ES-117

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Table 3.k.1-5 Stress Limits for Strainer Components (250 degrees F) Service Level Stress Category Stress Limit (ksi) Design Pm s 17.15 Pm+ Pb 1.5 s 25.725 Service Level B Pm 1.1 s 18.865 Pm+ Pb 1.65 s 28.3 Pm* s 16.35 Pm+ Pb+ Q* 3 Sm 69.9 Service Level D Pm 2.0 s 34.3 Pm+ Pb 2.4 s 41.16 S: 17,150 psi for SS304 Sm: 23,300 psi for SS410 Table 3.k.1-6 Weld Stress Limits (250 degrees F) Type Service Level Stress Stress Limit (ksi) Category Fillet ND-3929 & ND-5260* Shear 0.85x0.7xS 10,200 Plug ND-3929 & ND-5260* Shear 0.65x0.8xS I 8,918 S: 17, 150 psi for SS304

• No specific weld inspection requirements. VT-visual test inspection will be performed.

ES-118

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary)

2. Summarize the structural qualification results and design margins for the various components of the sump strainer structural assembly.

Response to 3.k.2: Table 3.k.2-1 Stress Ratio Summary for Strainer Components Based on ASME Subsection NC Component Service Level Stress Ratio* Perforated Plates Design - RHR model 18.55 Fingers Design - RHR model 21.10 Finger Frames Design - RHR model 40.70 Perforated Spacers Design - RHR model 17.23 Center Post Design - RHR model 39.27 Connecting Plates Design - RHR model 41.69 Support Base Design - RHR model 16.40 Base Frame Design - RHR model 16.84 I-Beams Design - RHR model 16.40 Perforated Plates Level B - RHR model 2.30 Fingers Level B- RHR model 3.65 Finger Frames Level B- RHR model 13.78 Perforated Spacers Level B - RH R model 10.49 Center Post Level B- RHR model 26.13 Connecting Plates Level B - RH R model 14.63 Support Base Level B- RHR model 9.58 Base Frame Level B - RHR model 9.58 I-Beams Level B - RHR model 18.04 Tie Rods Level B - RHR model 2.38 Perforated Plates Level D - RHR model 3.33 Fingers Level D - RH R model 5.30 Finger Frames Level D - RHR model 19.83 Perforated Spacers Level D - RHR model 12.44 Center Post Level D - RHR model 30.11 Connecting Plates Level D - RHR model 21.09 Support Base Level D - RHR model 14.96 Base Frame Level D - RHR model 14.96 I-Beams Level D - RHR model 14.36 Perforated Plates Level D - RHR Train B model 3.35 Fingers & Frames Level D - RHR Train B model 5.21 Perforated Spacers Level D- RHR Train B model 10.67 Center Post Level D- RHR Train B model 26.78 Connecting Plates Level D- RHR Train B model 20.47 Support Base Level D- RHR Train B model 7.16 Base Frame Level D - RHR Train B model 9.97 I-Beams Level D- RHR Train B model 9.40

• Stress Ratio= ASME Code Stress Limit I Calculated Max Stress ES-119

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Table 3.k.2-2 Stress Summary for Welds based on Service Level D Load Weld Location Weld Stress Allowable Stress (psi) (psi) Stress** Ratio* Perforated Plate to Finger 3,708 8,918 2.4 Perforated Plate to Finger 9,016 10,200 1.13

• Stress Ratio= ASME Code Stress Limit I Calculated Max Stress

          ** Conservative Level A Stress Limits, ASME Code Section Ill, Subsection ND-3923 at 250 °F The ASME Code combination stress limits are summarized in Tables 3.k.1-5 and 3.k.1-6.

Table 3.k.2-3 Typical RHR Strainer Stress Ratios for Service Level D Component Stress Max Stress Intensity Stress Limit Stress Ratio* Category (psi) (psi) Perforated Plates Pm less than 12,347 34,300 2.78 minimum Pm+ Pb 12,347 41,160 3.33 Fingers Pm less than 7,764 34,300 4.42 minimum Pm+ Pb 7,764 41, 160 5.30 Finger Frames Pm less than 2,076 34,300 16.52 minimum Pm+ Pb 2,076 41, 160 19.83 Perforated Spacers Pm less than 3,308 34,300 10.37 minimum Pm+ Pb 3,308 41, 160 12.44 Center Post Pm less than 1,367 34,300 25.09 minimum Pm+ Pb 1,367 41, 160 30.11 Connecting Plates Pm less than 1,952 34,300 17.57 minimum Pm+ Pb 1,952 41, 160 21.09 Support Base Pm less than 6,855 34,300 5.00 minimum Pm+ Pb 6,855 41,160 6.00 Base Frame Pm less than 6,855 34,300 5.00 minimum Pm+ Pb 6,855 41, 160 6.00 I-Beams Pm less than 5,684 34,300 6.03 minimum Pm+ Pb 5,684 41, 160 7.24

• Stress Ratio= ASME Code Stress Limit I Calculated Max Stress E5-120

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Table 3.k.2-4 RHR Train B Strainer Stress Ratios for Service Level D Component Stress Max Stress Intensity Stress Limit Stress Ratio* Category (psi) (psi) Perforated Plates Pm less than 12,289 34,300 2.79 minimum Pm+ Pb 12,289 41, 160 3.35 Fingers & Frames Pm less than 7,894 34,300 4.35 minimum Pm+ Pb 7,894 41, 160 5.21 Perforated Spacers Pm less than 3,856 34,300 8.90 minimum Pm+ Pb 3,856 41, 160 10.67 Center Post Pm less than 1,537 34,300 22.32 minimum Pm+ Pb 1,537 41,160 26.78 Connecting Plates Pm less than 2,011 34,300 17.06 minimum Pm+ Pb 2,011 41,160 20.47 Support Base Pm less than 5,750 34,300 5.97 minimum Pm+ Pb 5,750 41, 160 7.16 Base Frame Pm less than 4,130 34,300 8.31 minimum Pm+ Pb 4,130 41, 160 9.97 I-Beams Pm less than 4,380 34,300 7.83 minimum Pm+ Pb 4,380 41, 160 9.40

• Stress Ratio= ASME Code Stress Limit I Calculated Max Stress Weld Analysis Since the finite element model with the typical RHR strainer configuration has slightly higher overall stress results, the ANSYS analysis results in this load case were used to calculate the load transfer through the welds. For a given weld location, the elements and corresponding nodes at the weld were selected on one side of the node, and the AN SYS post-processor was used to calculate the forces transferred across the weld section. These forces were then used to calculate the stresses based on the weld section properties. If the welds consisted of more than one weld, then the group section properties were used.

Weld stresses were calculated for simultaneous application of loads for Service Level D. These calculated stress values were compared with the ASME Code shear stress limits. The minimum weld stress ratios for all the weld locations are summarized in Tables 3.k.2-1 and 3.k.2-2. For the welds between the fingers and perforated plate, robotic welding will be utilized to ensure a weld diameter of 3/16 inches. At the worst stress intensity finger location, a net shear Fx of 115.8 pounds force (lbf), a net shear Fy of 184 lbf, and a net tensile Fz of 184 lbf are obtained between two sides of the finger as seen in Figure 3.k.2-1. ES-121

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) AFz= 184 lbf . z AFy= 184 lbf M'x= 115 .8 lbf Figure 3.k.2-1 Worst Stress Intensity Finger Location Free Body Diagram Considering a line of welds, net forces are reacted by the circular areas of the plug weld, Aw. The unbalanced Fz causes a moment of 30 inch-pounds force (in-lbf) and is reacted by the section modulus of the weld, Sw, when the weld is treated as a line. The unbalanced Fx and Fy cause torsion and are reacted by the twisting property of the weld, Jw. Jw is large because the line of weld is approximately 8 inches long. The stresses caused by torsion are therefore negligible. Weld load treated as a line: 2 2 2 _ ( M Fz ) ( Fx) + (AFwy) f- Sw +Aw + Aw f Sa= Nt where Sa= 8,918 psi and t = 0.078 inches Five welds along each finger will satisfy the stress allowable of 8,918 psi. The welds are to be distributed evenly along the finger. Similarly, at the weld location between the finger frame and the perforated plate, a net shear Fx of 180 lbf., a net shear Fy of 325 lbf, and a net tensile Fz of 1,437 lbf are obtained between two sides of the frames. The moment from the unbalanced Fz is 566 in-lbf and is reacted by Sw of 1,200 in 2 for the square frame shape weld line. The fillet weld area, Aw, is 0.707 x 2t, where t equals 0.078 inches. In addition, an intermittent weld has a knock down factor of 0.66 for weld length of 3 inches and pitch distance of 5 inches. The weld calculation shows that 18 inches of weld length is recommended along each edge of the disk. The fillet welds should cover corners and at finger protrusion areas. Based on a width of 0.070 inches for the weld and stress allowable of 10,200 psi, the recommended intermittent welds should be 3 inches with 5-inch pitch. ES-122

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Interface Load The original Vogtle strainers were re-evaluated when the original 16 bolt anchor configuration was revised to a 32 bolt anchor configuration. The updated anchor bolt loads from the simplified GEH finite element model are documented in GEH letter no. JXDR7-2006-02, Rev. 1. The worst case anchor loads were for Level D load cases (Wd + SSE2). The largest tension load (z-direction) that an individual anchor bolt sees is 195 lbf. The largest lateral X-direction force is 237 lbf, and the largest Y-direction force is 279 lbf. The worst overall anchor bolt interaction ratio for each load case is provided below in Table 3.k.2-5. Table 3.k.2-5 Worst Case Interaction Ratios (l.R.) for Anchor Bolts Fx (lbf) Fy (lbf) Fz (lbf) l.R.* WD + SSE2 152.4 167.59 -195.4 0.423 (downward) WD + SSE2 137.91 278.26 -38.734 0.350 (upward)

• The Interaction Ratio (l.R.) is the inverse of a stress ratios in the tables above and equals the resultant force per anchor divided by the allowable force per anchor.

16-Disk ECCS Suction Strainer Summary (( ES-123

Enclosure 5 Supplemental Response to 'NRC Generic Letter 2004-02 (Non-Proprietary) Table 3.k.2-6: Service Level D Stress Summary for 16-Disk Strainer II

3. Summarize the evaluations performed for dynamic effects such as pipe whip, jet impingement, and missile impacts associated with high-energy line breaks (as applicable).

Response to 3.k.3: As shown in Figure 3.a.1-1, there is only one line outside of the steam generator compartments with analyzed breaks (i.e. welds inside the first isolation valve). However, this 2" line is at an elevation of 208 ft, which is 32 ft above the strainer at approximately 176 ft. Thus, pipe whip from this line impacting the strainer is not considered a credible scenario. The strainers are seismically qualified, robust structures designed with a crush pressure of approximately 10.4 psi, which is approximately the impingement pressure at 11.5 pipe diameters from a break (Reference 3). Considering the distance from the analyzed break, jet impingement loads are not a credible concern. Finally, the line in question is on the side of the pressurizer cubicle wall, where no unsecured items would be located. Therefore, missiles generated by this analyzed break are not a credible concern. There are no other high-energy lines in the area of the emergency sumps except for the RHR and HHSI lines that are used for accident mitigation and are not assumed to be the accident initiator. ES-124

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) The 12-inch RHR hot leg recirculation line is located more than 6 ft above the Train B CS strainer outside of the secondary shield wall. The RHR suction header has two isolation valves in series to isolate it from the RCS hot leg recirculation line. The valves are normally closed except when the RHR is operating. The inboard isolation valve is on the other side of the secondary shield wall inside the steam generator compartment preventing high-energy RCS discharge outside of the secondary shield wall. Therefore, there is no possibility of pipe whip impacts or jet loads associated with this pipe. In addition, this line is only pressurized during shutdown, refueling, and accident mitigation. During normal operation, this line is not pressurized. Thus, this line is not considered for evaluation as a postulated location for a high-energy line break accident. The 6-inch HHSI line is located approximately 69 inches above the Train B RHR strainer outside of the secondary shield wall. There is a check valve on the other side of the secondary shield wall inside the steam generator compartment preventing high-energy RCS discharge outside of the secondary shield wall. Therefore, there is no possibility of pipe whip impacts or jet loads associated with this pipe. In addition, this line is only pressurized during accident mitigation. During normal operation, this line is not pressurized. Thus, this line is not considered for evaluation as a postulated location for a high-energy line break accident. The strainers are located outside of the steam generator compartments and inside the outer containment wall. Therefore, the strainers are adequately protected from the hazardous effects of missiles.

4. If a backflushing strategy is credited, provide a summary statement regarding the sump strainer structural analysis considering reverse flow.

Response to 3.k.4: Backflushing of the sump strainers, or any other active approach, is not credited in the VEGP analysis. The RHR strainer suction lines have check valves to prevent reverse flow through the strainers. If containment pressure is high enough to actuate containment spray, the pressure is high enough to prevent reverse flow through the CS strainers. Therefore, no structural analysis considering reverse flow is required. ES-125

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non~Proprietary) I. Upstream Effects The objective of the upstream effects assessment is to evaluate the flowpaths upstream of the containment sump for holdup of inventory, which could reduce flow to and possibly starve the sump. Provide a summary of the upstream effects evaluation including the information requested in GL 2004-02 Requested Information Item 2(d)(iv). GL 2004-02 Requested Information Item 2(d)Ov) The basis for concluding that the water inventory required to ensure adequate EGGS or GSS recirculation would not be held up or diverted by debris blockage at choke points in containment recirculation sump return flowpaths.

1. Summarize the evaluation of the flowpaths from the postulated break locations and containment spray washdown to identify potential choke points in the flow field upstream of the sump.

Response to 3.1.1: The following areas I items were considered as part of the evaluation to determine potential choke points for flow upstream of the sump: Refueling Cavity Evaluations of containment, along with review of the CFO model, indicated no significant areas would become blocked with debris and hold up water during the sump recirculation phase. The area of the refueling cavity, which is the area around the reactor head that is flooded prior to fuel movement, is the only significant area in containment that can retain water during an event that requires containment spray. However, this area is drained by a large clear flow path that cannot be easily blocked with debris. See the Response to 3.1.4 for additional information. Inside Secondary Shield Wall A postulated LOCA inside the secondary shield wall in the lower elevations of the containment was considered limiting with respect to flow restrictions upstream of the sump. The flow path from this break area to the sump strainers is primarily through two labyrinth-like walkways through the shield wall. There are also smaller openings through the shield wall for piping, but these are much smaller than the walkways. The walkways provide a large clear flow path from inside the shield wall to the screen area. In addition, any restriction of the smaller through-wall piping openings would have minimal effect on the overall flow path to the strainers, since water would simply flow through the open walkways. ES-126

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Containment Spray Washdown Containment spray washdown has a clear path to the containment sump area. Large sections of the floor on each level in containment are covered with grating that allows the water to pass. A complete evaluation of containment, along with a review of the CFO model, indicated no significant areas would become blocked with debris and hold up water during the sump recirculation phase.

2. Summarize measures taken to mitigate potential choke points.

Response to 3.1.2: Per the Response to 3.1.1, no measures were necessary to mitigate potential choke points.

3. Summarize the evaluation of water holdup at installed curbs and/or debris interceptors.

Response to 3.1.3: There are no curbs or debris interceptors that provide water volume holdup in the VEGP containments.

4. Describe how potential blockage of reactor cavity and refueling cavity drains has been evaluated, including likelihood of blockage and amount of expected holdup.

Response to 3.1.4: The refueling cavity is drained by two 12-inch pipes. During refueling, these drains are secured by installing flanges. These flanges are removed prior to entry into Mode 4 and above. The VEGP limiting break with respect to upstream flow blockage occurs under the operating deck and inside the secondary shield wall. This break would result in a torturous path for large debris to travel above the operating deck and land in the refueling cavity. Large debris would have to travel through the RCP access ports or the Steam Generator and Pressurizer cubicles. The RCS access ports are covered with grating, and there are significant structural elements that would prevent any large pieces of debris from entering upper containment. The same is true for the path through the Steam Generator and Pressurizer cubicles. Each cubicle contains several levels of grating and significant structural elements that would make large pieces of debris entering upper containment via these paths highly unlikely. Therefore, the clogging of the refueling cavity drains is minimized. ES-127

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 {Non-Proprietary) The drains into the area under the reactor (e.g., reactor cavity) could become blocked. For breaks outside of the reactor cavity, there is no detrimental impact of this blockage, as it would inhibit loss of water from the active EGGS sump to an inactive area beneath the vessel. The flooding analysis assumes this area floods during the event. For breaks inside the reactor cavity, there is also no detrimental impact of this blockage, since the majority of the flow from the break would travel to the EGGS sump through the hot leg and cold leg penetrations.

m. Downstream Effects - Components and Systems The objective of the downstream effects, components and systems section is to evaluate the effect of debris carried downstream of the containment sump screen on the function of the EGGS and GSS in terms of potential wear of components and blockage of flow streams.

Provide the information requested in GL 2004-02 Requested Information Item 2(d)(v) and 2(d)(vi) regarding blockage, plugging, and wear at restrictions and close tolerance locations in the EGGS and GSS downstream of the sump. GL 2004-02 Requested Information Item 2(d)(v) The basis for concluding that inadequate core or containment cooling would'not result due to debris blockage at flow restrictions in the ECCS and CSS flowpaths downstream of the sump screen (e.g., a HPSI throttle valve, pump bearings and seals, fuel assembly inlet debris screen, or containment spray nozzles). The discussion should consider the adequacy of the sump screen's mesh spacing and state the basis for concluding that adverse gaps or breaches are not present on the screen surface. GL 2004-02 Requested Information Item 2(d)(vi) , Verification that the close-tolerance subcomponents in pumps, valves and other ECCS and CSS components are not susceptible to plugging or excessive wear due to extended post-accident operation with debris-laden fluids.

1. If NRG-approved methods were used (e.g., WGAP-16406-P-A with accompanying NRG SE), briefly summarize the application of the methods. Indicate where the approved methods were not used or where exceptions were taken, and summarize the evaluation of those areas.

Response to 3.m.1: The following methodology was employed in the ex-vessel downstream effects* evaluations. The evaluations did not use any unapproved methods or take any exceptions to NRG-approved methods. ES-128

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Maximum Debris Ingestion Determination Debris blockage of flow restrictions in the ECCS and CSS flowpaths downstream of the sump screen was addressed within the downstream effects evaluations for ECCS valves and equipment. Each unit has two sets of screens - RHR and CS emergency sump screens. The adequacy of the sump screens' mesh spacing or strainer hole size [nominal hole diameter of 0.09375 inches (3/32 inches)] is conservatively addressed by assuming that the maximum amount of particulate (coatings and latent debris) transported to the strainers passes through the strainers. Additionally, the evaluation used a quantity of fiber debris that passes through the strainers (100 g/FA), which is greater than the maximum total reactor vessel fiber load amount shown for a hot-leg break in Table 3.n.1-6. The ex-vessel downstream effects evaluations were based on this maximum amount of ingested debris (see Initial Debris Concentrations below). The Unit 1 strainers were inspected after installation and found to conform to design specifications. No adverse gaps or breaches were found on the screen surface. The Unit 2 strainers were inspected upon installation, and deficiencies in the fabrication of the Unit 2 CS sump screens were discovered; specifically, there were 124 holes greater than the nominal specified sump screen hole-diameter of 0.09375 inches (3/32 inches). No holes greater than a 0.25-inch diameter were found in the Unit 2 strainers; therefore, ingestion of debris will not cause plugging of downstream CS components because the smallest component diameter is 0.375 in. Initial Debris Concentrations Initial debris concentrations were developed using the assumptions and methodology described in Chapter 5 of WCAP-16406-P-A. Additionally, for conservatism, the maximum amount of particulate (coatings and latent debris) transported to the strainer were assumed to pass through the strainer. The total maximum initial debris concentration was determined to be 919.17 ppm, with fiber debris contributing 11.61 ppm, and particulate and coating debris contributing 907.56 ppm. Flowpaths and Alignment Review Both trains of the RHR system, SI system, component cooling system (CCS), and CSS were reviewed to ensure that all of the flowpaths and components impacted by the debris passing through the sump screens were considered. Documents used for this effort included piping and instrumentation diagrams (P&IDs), vendor manuals, equipment specifications, and other documents as applicable. ES-129

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Component Blockage and Wear Evaluations Methodology All component evaluations were performed based on WCAP-16406-P-A. Components addressed in the evaluations include pumps, heat exchangers, orifices, spray nozzles, instrumentation tubing, system piping, and valves required for the post-LOCA recirculation mode of operation of the ECCS and CSS. The evaluations included the following steps:

                                                          ]a,c

2. Provide a summary and conclusions of downstream evaluations.

Response to 3.m.2: Summary and Conclusions of Downstream Evaluations The following is the summary of results and conclusions of the downstream effects evaluations: ECCS/CSS Pumps For pumps, the effects of debris ingestion through the sump* screen on three aspects of operability (hydraulic performance, mechanical-shaft seal assembly performance, ES-130

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary} and mechanical performance) were evaluated. The hydraulic and mechanical performances of the ECCS and CSS pumps were determined to be unaffected by the recirculating sump debris. The mechanical shaft seal assembly performance evaluation resulted in one action item with the suggested replacement of the RHR pumps' carbon/graphite backup seal bushings with a more wear-resistant material, such as bronze. However, because VEGP has an engineered safety feature (ESF) atmospheric filtration system in its auxiliary building, this action is not required per WCAP-16406-P-A. ECCS/CSS Heat Exchangers, Orifices, Spray Nozzles, and System Piping Heat exchangers, orifices, spray nozzles, and system piping were evaluated for the effects of erosive wear for an initial concentration of 919.17 ppm over the mission time of 30 days. The erosive wear on these components was determined to be insufficient to affect system performance. The smallest clearance found for VEGP heat exchangers, orifices, spray nozzles, and system piping in the ECCS recirculation flow path is 0.188 inches, for the SI pressure breakdown orifices. Therefore, no blockage of the ECCS flow path is expected with the current ECCS sump screen hole size of 0.09375 inches. The smallest clearance found in the CSS recirculation flow path is 0.375 inches, for the CS nozzles. No blockage of the CSS flow path is expected because the maximum debris size able to bypass the CSS sump screens is 0.25 in. System piping was evaluated for plugging based on system flow and material settling velocities .. For all piping, the minimum flow velocity was found to be greater than 0.42 ft/s, the minimum velocity required to prevent debris sedimentation. All system piping passed the acceptable criteria for plugging due to sedimentation. ECCS/CSS Valves Valves were evaluated for plugging impact in the downstream effects evaluations. Valves that were determined to be "Not Critical" did not warrant further evaluation, but those valves identified as "Evaluation Required" received a more detailed evaluation. It was determined that all valves passed the acceptance criteria for the plugging evaluation. Valves were evaluated for debris sedimentation. Valves identified as "No Evaluation" did not require additional analysis, but valves identified as "Evaluation Required" were analyzed further. The line velocity for all valves analyzed was found to be greater than [ ]a,c thus, debris sedimentation was not an issue. Valves were evaluated for wear impact. Valves determined to be "Not Critical" did not warrant further evaluation, but valves identified as "Evaluate" were analyzed further. It was found that four throttle valves did not pass the acceptance criteria of [

                                                                  ]a,c Because of this, ES-131

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) they were evaluated for additional opening. This additional opening is necessary for the valves to demonstrate acceptable wear. The turns open value (from a fully closed position) was calculated and rounded to the nearest half turn. With the new

  . calculated turns open value, the valves passed the acceptance criteria of [
                                                              ]a,c ECCS/CSS Instrumentation Lines Instrumentation tubing was evaluated for debris settling. It was found that the largest settling velocity of debris source material found inside a PWR containment is

[ ]a,c as provided by WCAP-16406-P-A. Therefore, as long as the recirculation flow velocity though the ECCS and CSS is greater than [

                                  ]a,c failure of instrumentation due to debris settlement will not occur. The evaluation showed that all flow velocities are greater than [
    ]a,c

3. Provide a summary of design or operational changes made because of downstream evaluations.

Response to 3.m.3: As noted in the Response to 3.m.2, an adjustment to the minimum opening of four throttle valves was required in order to resolve erosion concerns. The minimum turns open value required to demonstrate acceptable valve wear was implemented into plant procedures. The results of the VEGP downstream effects evaluations demonstrate that the evaluated components are acceptable for the expected mission time.

n. Downstream Effects - Fuel and Vessel The objective of the downstream effects, fuel and vessel section is to evaluate the effects that debris carried downstream of the containment sump screens and into the reactor vessel has on core cooling.

1. Show that the in-vessel effects evaluation is consistent with, or bounded by, the industry generic guidance (WCAP-16793-NP), as modified by NRC staff comments on that document. Briefly summarize the application of the methods. Indicate where the WCAP methods were not used or where exceptions were taken, and summarize the evaluation of those areas.

Response to 3.n.1: In-vessel downstream effects for VEGP were evaluated per the methodology in WCAP-16793-NP-A (Reference 22) and the associated NRC SE (Reference 94) using assumed values for in-vessel debris accumulation limits. The evaluation included the following: E5-132

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary)

1. Peak cladding temperature (PCT) due to deposition of debris on fuel rods 0/VCAP-16793-NP-A).

2. Deposition thickness (OT) due to collection of debris on fuel rods 0/VCAP-16793-N P-A).

3. Amount of fiber accumulation at the reactor core inlet and inside the reactor vessel (limits assumed for the purpose of exercising the risk-informed methodology).

These analyses concluded that post-accident long~term core cooling (L TCC) will not be challenged by deposition of debris on the fuel rods, accumulation of debris at the core inlet, or accumulation of debris in the heated region of the core for all postulated LOCAs inside containment. A brief summary of the relevant testing and analyses is provided below as it was used to inform the WCAP evaluations. VEGP Fiber Penetration Testing VEGP conducted fiber penetration testing in 2015. The purpose of the testing was to collect time-dependent fiber penetration data of a prototypical VEGP strainer under various conditions (e.g., approach velocity, water chemistry) and strainer configurations (e.g., number of strainer disks). The test results were used to derive a model that can be used to quantify fiber penetration for the RHR and CS strainers at plant conditions. Twelve penetration tests were conducted, nine of which are useful to inform the resolution of GL 2004-02. Within those nine tests, the approach velocity, the water chemistry, and the number of strainer disks were varied to investigate their effects on fiber penetration. Test Loop Design The test loop consisted of a metal test tank, which housed a test strainer at its downstream end. Water was circulated by a pump through the test strainer, a fiber filtering system, and various piping components. The test tank had a flume geometry, as shown in Figure 3.n.1-1. Debris was introduced in the high-agitation region. This region was equipped with two mechanical mixers to create adequate mixing and prevent the debris from settling. Mixing inside the low-agitation region was created by directing a portion of the returning flow through the perforated bottom plate of the region. This mixing motion kept fiber in suspension without disturbing the fiber bed on the strainer. The strainer region was designed such that the spacing between the test strainer and tank walls imitated the gaps between adjacent strainer stacks, or between the nearest object inside the containment and the strainer. The spacing between the strainer and tank walls was also designed to minimize settling of debris. ES-133

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) FLOW 0 LOW AGITATION HIGH AGITATION REGION REGION ( I

                    /         --------~
                                                 ----~-----1------~

STRAINER __/ l Mechanical Mixers _J Figure 3.n.1-1 General Arrangement of Test Tank The effectiveness of the agitation regions is displayed in Table 3.n.1-1, which documents the quantity of fiber that did not transport to the strainer and was collected from the high or low agitation regions after the conclusion of each test. Table 3.n.1-1: Summary of Useful Fiber Transport Tests Gross Fiber Non-Transported Net Fiber  % of Fiber Test # 1 Added (g) Fiber (g) Added (g) Transport 1 17350.2 236.7 17113.5 98.6% 2 17350.2 397.9 16952.3 97.7% 3 11403.1 203.9 11199.2 98.2% 4 11401.4 205.2 11196.3 98.2% 5 14375.5 190.6 14184.9 98.7% 6 14466.6 171.9 14294.7 98.8% 7 17281.3 578.1 16703.2 96.7% 8 17281.3 278.3 17003.1 98.4% 10 14375.6 0.0 14375.6 100.0% 1 The test numbers shown correspond to the number assigned to each test in the VEGP fiber penetration test report. Note that Tests 9, 11, and 12 are not shown because they were the three tests that were not applicable to this resolution. Test Strainer The test strainer was a prototypical strainer stack; the only difference was the number of disks installed on the test strainer. While the VEGP RHR and CS strainer stacks consist of 16 and 14 disks, respectively, the number of disks on the test strainer was varied among 12, 15, and 18 disks for different tests. The testing flow rate and debris load for each test were determined based on the area ratio of the test strainer to the prototypical strainer assembly, which was varied according to the ES-134

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) number of disks on the test strainer. As a result, the debris load per unit strainer area and water approach velocity were preserved among tests with varying number of strainer disks. This approach allowed the effect of varying the number of strainer disks on penetration quantities to be investigated separately. Though the number of disks for the tested strainer did not exactly match the plant strainers, the tested configurations bounded the plant configurations. As discussed later in this response, the results from those tests with different number of disks were interpolated to derive the penetration model for the plant strainer configurations. Debris Types and Preparation Nukon was the only debris type used in testing. This is appropriate because the only other type of fibrous debris in containment, fire barrier material, transports in negligible amounts of fine fiber. All Nukon used in testing was introduced as fines. Nukon fines were prepared according to the NEI protocol (Reference 46). Nukon sheets, with an overall thickness of 2 inches, were baked single-sided into approximately half the thickness. The heat-treated sheets were then cut into cubes and weighed out according to batch size. Batches were then pressure-washed with test water following the NEI protocol (Reference 46). Debris Introduction Debris was introduced in eight separate batches of increasing batch size for each test. The first two batches corresponded to a theoretical uniform bed thickness of 1/16 inch. The third through seventh batches corresponded to a theoretical uniform bed thickness of 1/8 inch. The final batch corresponded to a theoretical uniform bed thickness of 1/4 inch. The total debris load was equivalent to a theoretical uniform bed thickness of 1 inch. This debris load was chosen because it was sufficient to circumscribe the test strainer. Subsequent debris addition after the development of a circumscribed debris bed would not result in an appreciable amount of penetration. Note that the debris introduction rate was controlled to maintain a prototypical debris concentration in the test tank. Debris Capture Fiber can penetrate through the strainer by two different mechanisms: prompt penetration and shedding. Prompt penetration occurs when fiber reaching the strainer travels through the strainer immediately. Shedding occurs when fiber that already accumulated on the strainer migrates through the bed and ultimately travels through the strainer. Both mechanisms were considered during testing. Fibers that passed through the strainer were collected by the fiber filtering system downstream of the test strainer. The filtering system had 5-micron filter bags installed in filter housings. All of the flow downstream of the strainer travelled through the filter bags before returning to the test tank. The capture efficiency of the filter bags was verified to be above 95 percent. The filtering system allowed the ES-135

     -------------------------------------------~

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) installation of two filter bags in parallel lines such that a filter bag could be left online at all times, even during periods in which filter bags were swapped. Before and after each test, all of the filter bags required for the test were uniquely marked and dried, and their weights were recorded. The weight gain of the filter bags was used to quantify fiber penetration. After testing, the debris-laden filter bags were rinsed with DI water to remove residual chemicals before weighing. For each bag, the drying and weighing process was repeated until two consecutive bag weights (taken at least 1 hour apart) were within 0.05 g of each other. A clean filter bag was placed online before a debris batch was introduced to the test tank and was left on line for a minimum of five pool turnovers (PTOs) to capture the prompt fiber penetration. For Batches 1 and 3, two additional filter bags were used to capture the fiber penetration due to shedding. Before further debris addition, a visual confirmation was required to verify that all introduced debris had transported to the strainer. This approach allowed the testing to capture time-dependent fiber penetration data, which was used to develop a model for the rate of fiber penetration as a function of fiber quantity on the strainer. Test Parameters The test water used for fiber penetration testing had a chemical composition prototypical to VEGP. The plant conditions selected for testin'g were those of minimum and maximum boron concentrations and the corresponding buffer (trisodium phosphate, TSP) concentrations. The low boron concentration was taken from an SBLOCA event, and the high boron concentration was taken from an LBLOCA with the CSS active. For the low boron concentration, a corresponding high plant TSP concentration was used, and the high boron concentration was coupled with a corresponding low plant TSP concentration. The chemical concentrations used in testing are shown in Table 3.n.1-2. Test water was prepared by adding chemicals to DI water until the prescribed concentrations were achieved. Table 3.n.1-2: Summary of Penetration Test Water Chemistry Chemical High Level Low Level Boron (ppm) 2,522 2,169 TSP (ppm) 2,181 3,147 Several different strainer approach velocities, ranging between 0.0043 ft/s and 0.0130 ft/s, were determined from plant operating conditions and were used for the VEGP fiber penetration testing. As shown in Table 3.f.3-1, the 16-disk RHR strainers have a surface area of 677.6 ft2, and the 14-disk CS strainers have a surface area of 590 ft2

• The design flow rates of the RHR and CS pumps are 3,700 gpm and 2,600 gpm, respectively. The average approach velocities of the ES-136

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) RHR and CS strainers are therefore 0.0122 ft/s and 0.0098 ft/s, respectively. Both of these velocities were enveloped by the range tested. In addition to water chemistry and approach velocity, the number of disks was also varied. A test matrix was designed to quantify fiber penetration for different combinations of test conditions. The test matrix is displayed in Table 3.n.1-3. Only the nine tests that are applicable to plant design conditions are shown. Table 3.n.1-3: Large Scale Penetration Test Matrix Approach No. of Boron I TSP Test# Velocity (ft/s) Disks Concentration (ppm) 1 0.0130 18 2,522/2,181 2 0.0043 18 2,522/2,181 3 0.0130 12 2,522/2,181 4 0.0043 12 2,522/2,181 5 0.00314 15 2,522/2,181 6 0.0087 15 2,169 / 3,147 7 0.0043 18 2,169 / 3,147 8 0.0087 18 2,522 / 2,181 10 0.0087 15 2,522 / 2,181 Strainer Penetration Model Development Data gathered from VEGP fiber penetration tests were used to develop a model for quantifying the RHR and CS strainer fiber penetration under prototypical plant conditions. Given the different characteristics and flow rates of the RHR and CS strainers, separate formulas were derived for the two strainers. The models were developed per the following steps:

• General governing equations were developed to describe both the prompt fiber penetration and shedding through the strainer as a function of time and fiber quantity on the strainer. The summation of the developed equations can be used to describe total fiber penetration. The equations contain coefficients whose values were determined separately for each test based on the test results.
• The results for each test were fit to the governing equations using various optimization techniques to refine the coefficient values. This produced a unique set of equations and thus a unique penetration model for each test. Figure 3.n.1-2 compares the curve fit with the test data for Test 8. As the figure shows, it is the summation of the prompt and shedding penetration curves that was fit to the test data. Since three parameters were varied during testing (approach velocity, water chemistry, and number of disks), the fitting coefficients are functions of these three parameters.

ES-137

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) 1800 1600 ... _ 1400 ~~ 1200 { _/ bO

                              ~

VI VI ~ 1000 r -c Q) 4J -

...ro
                                                                                                  ~

4J 800 - Prompt Bypassed Mass f-Q) c Q)

                                                                       -   Shed Mass a_
                          ~                                                                        I 600
                                                                       -   Total Penetrated Mass r

___.! Actual Penetration [g) 400 . A ... { 200 0 0 10000 20000 30000 40000 50000 60000 Time [s] Figure 3.n.1-2: Test 8 Penetration Model Fit

• The test models from the previous step were then used to develop separate models for the RHR and CS strainers , respectively, by interpolating from the model coefficients of each test. During this process , different weighting factors were applied to each test model based on the similarity of the test conditions to the conditions of the actual RHR and CS strainer configurations (as shown in Table 3.n.1-4). Since the RHR and CS strainer conditions are different, each test was weighted separately for the RHR and CS model according to the similarity of its conditions to those of the RHR and CS strainers . Note that the RHR strainer configuration modeled below is based on the strainer modification described in the Response to 3.j .1. The effects of water chemistry on the penetration model were accounted for differently. Instead of interpolating between tests , the low-boron/high-TSP tests were used to represent the water chemistry of interest and were weighted more heavily when calculating the model parameters. This is conservative , firstly , because the large-scale fiber penetration test report showed that tests using low-boron and high-TSP concentrations resulted in higher fiber penetration under otherwise identical conditions , as shown in Figure 3.n .1-3.

Secondly, the low-boron water chemistry conditions used in testing were E5-138

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) determined with the intent of finding the minimum sump boron concentration by using an unlikely combination of conservative inputs. Table 3.n.1-4: RHR and CS Strainer Model Parameters Strainer Stack Disk# Velocity (ft/s) Boron (ppm) RHR 16 0.0122 2,169 cs 14 0.0098 2,169 The resulting RHR fiber penetration model was applied to prototypical plant conditions to calculate the total fiber penetration as a function of time , which is shown as the dark solid line in Figure 3.n.1-3. The curve fit models for the actual test cases were also applied to the same plant conditions , and the results are shown in Figure 3.n.1-3 as lines of different colors. It should be noted that the bracketed values shown in the legend of the figure are the parameters used for each test, not the model conditions, which are common for each curve . As shown in the figure , the model developed for the RHR strainer provides a higher total penetrated fiber quantity than all of the test conditions . This is expected because the RHR strainer model was developed with high approach velocity and low boron concentration (referred to as "low chem" in the figure legend) , which are the most conservative values for those parameters, and none of the tests were run with that combination . As shown in the figure , high approach velocity and low boron concentration , or "low chem" , increase fiber penetration under otherwise identical conditions . 16000 14000

               ~
               ~
                                                                    -   RHR [16 Disk, 0.0122 ft/s, Low Chem]
                                                                    - - Test 8 (18 Disk, 0.00876 ft/s. High Chem]

12000 {;// Ei Ji ii: "O 10000 fl/ / Test 1 [18 Disk, 0.01302 ft/s, High Chem] Test 7 [18 Disk, 0.00435 ft/s, Low Chem] Test 6 [15 Disk, 0.00879 ft/s, Low Chem] 2

~

Q) cQ) Q_ 8000 6000 r; - Test 10 [15 Disk, 0.00885 ft/s, High Chem] Test 3 [12 Disk, 0.01323 ft/s, High Chem]

            ' ~-

'iii 0 - - Test 2 [18 Disk, 0.00433 ft/s, High Chem] I-r 4000 - - Test 5 [15 Disk, 0.00317 ft/s, High Chem]

                                                                    -   Test 4 (12 Disk, 0.00443 ft/s, High Chem]

2000 0 0 20000 40000 60000 80000 100000 120000 Time[s] Figure 3.n.1-3: Comparison of RHR Model with Test Cases at Plant Scale It should be noted that the 95% confidence interval uncertainty in the RHR model output is 137.7 g. This quantity is not included in the model output displayed in E5-139

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Figure 3.n .1-3, but it is added to the calculated in-vessel fiber accumulation quantity found in the in-vessel hand calculation discussed later in this section . Figure 3.n.1-4 shows the prompt fiber penetration fraction as a function of fiber quantity on the strainer derived using the RHR strainer model. As expected , the prompt penetration fraction decreases as a fiber debris bed forms on the strainer. Because shedding penetration is a function of both fiber quantity on the strainer and time , it cannot be similarly shown . 45% 40% 35% 30% +-

          '*:ti"'a. 25%

co> a.E 20%

            ~

0.. 15% 10% 5% 0% 0 2000 4000 6000 8000 10000 12000 14000 16000 18000 Fiber on Strainer [g] Figure 3.n.1-4: Prompt Fiber Penetration Fraction for RHR Strainer Model In-Vessel Effects Evaluations Peak Cladding Temperature and Deposition Thickness The LOCA deposition model (LOCADM) , which is contained as part of WCAP-16793-NP-A (Revision 22) , was used to determine the scale thickness due to deposition of debris that passes through the strainer on the fuel rod surfaces and the resulting peak cladding temperature. The calculated scale thickness was then combined with the thickness of existing fuel cladding oxidation and crud build-up to determine the total deposition thickness . The calculated total deposition thickness and peak cladding temperature were compared with the acceptance criteria provided in WCAP-16793-NP. Note that the VEGP evaluation also considered the applicable E5-140

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) ,. requirements and recommendations from the following PWROG letters: OG-07-419, OG-07-534, OG-08-64, and OG-10-253. Two different cases were considered in this evaluation per the WCAP: minimum containment sump pool volume (Case 1) and maximum containment sump pool volume (Case 2). Table 3.n.1-5 below summarizes the maximum PCT and DT for these two cases. Table 3.n.1-5: Summary of PCT and OT for Cases 1 and 2 PCT (°F) DT (mils) Case Acceptance Acceptance Results Results Criteria Criteria Case 1: Minimum Initial 412 22.7 Sump Pool Volume

                                                < 800                    < 50 Case 2: Maximum Initial 412                     22.3 Sump Pool Volume For either case, the PCT is much lower than the acceptance criterion of 800 degrees F, and the DT is well within the acceptance criterion of 50 mils.

Therefore, deposition of post-LOCA chemical precipitate on the fuel rods will not block the LTCC flow through the core, nor will it create unacceptable local hot spots on the fuel cladding surfaces. The list below summarizes the key inputs and conservatisms used in the LOCADM analysis:

1. When calculating the "bump-up" factor to account for the fiber that passes through the strainer, a bounding value of 100 g/FA was used. As shown in Table 3.n.1-6, this quantity bounds the actual fiber loads for VEGP.

2. The surface area of aluminum coatings was conservatively calculated with operating margin.

3. The maximum sump pH, rather than the actual sump pH profile, was used for the entire 30-day mission time. This is conservative because higher sump pH values result in greater DT.

4. A combination of inputs was used to conservatively determine the PCT and/or DT in LOCADM.

a. Spray was assumed to start immediately and continue for 30 days after the beginning of the LOCA.

b. A conservatively high value for sump temperature was used to set the density of the sump water in order to minimize its mass for the given volume, thereby resulting in higher chemical concentrations within the sump.

c. A conservatively high value for reactor vessel coolant temperature was used to set the density of the reactor coolant in order to minimize its mass for the ES-141

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-f>roprietary) given volume, thereby resulting in higher chemical concentrations within the reactor vessel. The fiber limit at the reactor core inlet given in WCAP-16793-NP-A (15 g/FA) was not used. Instead, accumulation of fiber at the reactor core inlet and inside the reactor vessel was evaluated using the methodology described as discussed in the following section titled "Accumulation of Fiber inside Reactor Vessel." The NRC Safety Evaluation of WCAP-16793-NP provided analysis and recommendations on the use of Westinghouse's WCAP-16793-NP, Revision 2 methodology and identified 14 limitations and conditions that must be addressed. VEGP's responses to these limitations and conditions are summarized below.

1. Assure the plant fuel type, inlet filter configuration, and ECCS flow rate are bounded by those used in the FA testing outlined in Appendix G of the WCAP. If the 15 g/FA acceptance criterion is used, determine the available driving head for a hot leg (HL) break and compare it to the debris head loss measured during the j, FA testing. Compare the fiber bypass amounts with the acceptance criterion given in the WCAP.

VEGP Response: This limitation/condition is associated with the 15 g/FA limit established in WCAP-16793-NP-A, which does not apply to VEGP. In-vessel fiber accumulation is not calculated using WCAP-16793-NP-A.

2. Each licensee's GL 2004-02 submittal to the NRC should state the available driving head for an HL break, ECCS flow rates, LOCADM results, type of fuel and inlet filter, and amount of fiber bypass.

VEGP Response: This limitation/condition is associated with the 15 g/FA limit established in WCAP-16793-NP-A, which does not apply to VEGP. In-vessel fiber accumulation is not calculated using WCAP-16793-NP-A.

3. If a licensee credits alternate flow paths in the reactor vessel in their LTCC evaluations, justification is required through testing or analysis.

VEGP Response: This limitation/condition is associated with the 15 g/FA limit established in WCAP-16793-NP-A, which does not apply to VEGP. In-vessel fiber accumulation is not calculated using WCAP-16793-NP-A. ES-142

Enclosure 5 Suppremental Response to NRC Generic Letter 2004-02 (Non-Proprietary)

4. The numerical analyses discussed in Sections 3.2 and 3.3 of the WCAP should not be relied upon to demonstrate adequate LTCC.

VEGP Response: VEGP does not use any of the conclusions drawn based on the fuel blockage modeling discussed in Sections 3.2 and 3.3 of the WCAP report. In-vessel fiber accumulation is not calculated using WCAP-16793-NP.

5. The SE requires that a plant must maintain its debris load within the limits defined by the testing (e.g., 15 g/FA), and any debris amounts greater than those justified by generic testing in the WCAP must be justified on a plant-specific basis.

VEGP Response: This limitation/condition is associated with the 15 g/FA limit established in WCAP-16793-NP-A, which does not apply to VEGP. In-vessel fiber accumulation is not calculated using WCAP-16793-NP-A.

6. The debris acceptance criterion can only be applied to fuel types and inlet filter configurations evaluated in the WCAP FA testing.

VEGP Response: This limitation/condition is associated with the 15 g/FA limit established in WCAP-16793-NP-A, which does not apply to VEGP. In-vessel fiber accumulation is not calculated using WCAP-16793-NP-A.

7. Each licensee's GL 2004-02 submittal to the NRC should compare the PCT from LOCADM with the acceptance criterion of 800 degrees F.

VEGP Response: As shown in Table 3.n.1-5 above, the calculated VEGP PCT is well within the acceptance criterion of 800 degrees F.

8. When utilizing LOCADM to determine PCT and OT, the aluminum release rate must be doubled to predict aluminum concentrations in the sump pool in the initial days following a LOCA more accurately.

VEGP Response: The methodology outlined in PWROG Letter OG-08-64 was followed to double the aluminum release rate in the LOCADM analysis.

9. If refinements specific to the plant are made to the LOCADM to reduce conservatisms, the licensee should demonstrate that the results still adequately bound chemical product generation.

VEGP Response: The VEGP LOCADM runs do not employ any conservatism-reducing refinements specific to the plant. Therefore, no additional justification is required. ES-143

Enclosure 5 Supplemental Response to NRC Generic Letter"2004-02 (Non-Proprietary)

10. The recommended value for scale thermal conductivity of 0.11 BTU/(h-ft-°F) should be used for LTCC evaluations.

VEGP Response: As stated in Appendix E of WCAP-16793-NP, the recommended thermal conductivity of 0.11 BTU/(h-ft-°F) can be converted to 0.2 W/m-K, which is used in the LTCC calculation.

11. The licensee's submittals should include the means used to determine the amount of debris that bypasses the EGGS sump strainer and the fiber loading at the fuel inlet expected for the HL and cold leg (CL) break scenarios. Licensees should provide the debris loads, calculated on a fuel assembly basis, for both the HL and CL break cases in their GL 2004-02 responses.

VEGP Response: This limitation/condition is associated with the 15 g/FA limit established in WCAP-16793-NP-A, which does not apply to VEGP. In-vessel fiber accumulation is not calculated using WCAP-16793-NP-A.

12. Plants that can qualify a higher fiber load based on the absence of chemical deposits should ensure that tests for their conditions determine limiting head losses using particulate and fiber loads that maximize the head loss with no chemical precipitates included in the tests. In this case, licensees must also evaluate the other considerations discussed in the first limitation and condition.

VEGP Response: This limitation/condition is associated with the 15 g/FA limit established in WCAP-16793-NP-A, which does not apply to VEGP. In-vessel fiber accumulation is not calculated using WCAP-16793-NP-A.

13. The size distribution of the debris used in the FA testing must represent the size distribution of fibrous debris expected to pass through the EGGS sump strainer at the plant.

VEGP Response: This limitation/condition is associated with the 15 g/FA limit established in WCAP-16793-NP-A, which does not apply to VEGP. In-vessel fiber accumulation is not calculated using WCAP-16793-NP-A

14. Each licensee's GL 2004-02 submittal to the NRG should not utilize the "Margin Calculator" as it has not been reviewed by the NRG.

VEGP Response: The VEGP evaluation does not use the "Margin Calculator. In summary, the evaluation showed that the peak cladding temperature and deposition thickness due to accumulation of debris on the fuel rods met the acceptance criteria and did not cause any failures. ES-144

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Accumulation of Fiber inside Reactor Vessel During the post-LOCA sump recirculation phase, debris that passes through the strainer could accumulate at the reactor core inlet or inside the reactor vessel, thereby potentially challenging LTCC. This effect is evaluated for both hot leg break (HLB) and cold leg break (CLB) scenarios using the assumed acceptance criteria. The evaluation used time-dependent fiber penetration fractions obtained from VEGP testing based on plant-specific inputs, as described earlier in this response. The penetration fraction varies with the amount of fiber on the strainer and the amount of time passed since the onset of recirculation. The evaluation was performed in the NARWHAL CFP calculation as well as a bounding hand calculation. The NARWHAL model used assumed values and acceptance criteria to evaluate every break in a self-consistent and time-dependent manner. The model showed no failures for any of the postulated breaks due to accumulation of fiber at the core inlet or inside the reactor vessels. Impact on the results due to variabilities in the inputs was evaluated by sensitivity analyses. The hand calculation served as a bounding evaluation in which the worst case combination of input parameters (e.g., pool volume, transport fiber load, number of RHR and CS trains in operation, RHR and CS pump flow rates, sump recirculation and hot leg switchover times, and CS duration) were used. The uncertainty of the fiber penetration model was added to the calculated fiber quantities for conservatism. The results of the hand calculation are summarized in the table below and are compared with the acceptance criteria. The resulting fiber quantities for both the HLB and CLB are bounded by the assumed acceptance criteria. This conclusion is consistent with the results of the NARWHAL CFP calculation that showed no failures due to accumulation of fibers at the core inlet or inside the reactor vessel. Table 3.n.1-6: Bounding Fiber Loads for HLB and CLB Scenarios a,c In summary, no failures were recorded for any of the postulated breaks due to accumulation of debris at the core inlet or inside the reactor vessel. ES-145

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary)

o. Chemical Effects The objective of the chemical effects section is to evaluate the effect that chemical precipitates have on head loss and core cooling.

1. Provide a summary of evaluation results that show that chemical precipitates formed in the post-LOCA containment environment, either by themselves or combined with debris, do not deposit at the sump screen to the extent that an unacceptable head loss results, or deposit downstream of the sump screen to the extent that long-term core cooling is unacceptably impeded. Content guidance for chemical effects is provided in Enclosure 3 to a letter from the NRC to NEI dated March 2008 (Reference 106).

Response to 3.o.1: The chemical effects strategy for VEGP includes:

• Quantification of chemical precipitates using the WCAP-16530-NP-A methodology with refinements for phosphate passivation of aluminum surfaces.
• Introduction of those pre-prepared precipitates in prototypical array testing.
• Application of an aluminum solubility correlation and integrated autoclave chemical test results to determine formation temperature/timing.
• Time-based determination of acceptable head losses.
• Extrapolation of the resulting head losses to 30 days.

As discussed in the Response to 3.a.1, VEGP has evaluated breaks at all Class 1 weld locations on the primary RCS piping, upstream of the first isolation valve. The amount of chemical precipitates was quantified individually for each of these breaks using the amount of LOCA generated debris for that respective break location. Other plant-specific inputs such as pH, temperature, aluminum amount, and spray times were selected to maximize the generated amount of precipitates. These amounts were scaled by the ratio of the test strainer area to the plant strainer surface area and are compared with the chemical debris quantities used in the prototypical array tests to determine the resulting head loss across the strainers. Before the tests were conducted, the SAS and calcium phosphate were prepared according to the WCAP-16530-NP-A "recipes" and were verified to meet the settling criteria within 24 hours of the test. During the test, a fiber and particulate debris bed was established on the strainer surfaces, the stabilization criteria was satisfied, and the pre-prepared precipitates were added to the test tank in batches. See the Response to 3.f.4 for further details on the head loss measured after introduction of chemical precipitates. See the Response to 3.f.10 for further details on how the chemical precipitate head loss was utilized in the NARWHAL CFP calculation. See the in-vessel effects evaluations in the Response to 3.n.1 for the evaluation of chemical precipitate deposition on the fuel rod surfaces. ES-146 I

                                                                                  ---- - - - - _ _ _ _ _I

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary)

2. Content guidance for chemical effects is provided in Enclosure 3 to a letter from the NRC to NEI dated March 2008 (Reference 106).

Response to 3.o.2: The NRC identified evaluation steps in "NRC Staff Review Guidance Regarding Generic Letter 2004-02 Closure in the Area of Plant-Specific Chemical Effect Evaluations" in March of 2008 (Reference 106). VEG P's responses to the GL Supplement Content evaluation step for each debris characteristic are i I I summarized below.

1. Sufficient 'Clean' Strainer Area: Those licensees performing a simplified chemical effects analysis should justify the use of this simplified approach by providing the amount of debris determined to reach the strainer, the amount of bare strainer area and how it was determined, and any additional information that is needed to show why a more detailed chemical effects analysis is not needed.

VEGP Response: As discussed in the Response to 3.a.1, VEGP has evaluated breaks at all ISi weld locatio_ns on the primary RCS piping, upstream of the first isolation valve. Many of the breaks analyzed resulted in fiber loads sufficient to fully cover the sump strainer screens. Therefore, VEGP is not crediting clean strainer area to perform a simplified chemical effects analysis. See the Figure 1 flow chart in "NRC Staff Review Guidance Regarding Generic Letter 2004-02 Closure in the Area of Plant-Specific Chemical Effect Evaluations" from March 2008 (Reference 106).

2. Debris Bed Formation: Licensees should discuss why the debris from the break location selected for plant-specific head loss testing with chemical precipitate yields the maximum head loss. For example, plant X has break location 1 that would produce maximum head loss without donsideration of chemical effects.

However, break location 2, with chemical eff~cts considered, produces greater head loss than break location 1. Therefore, the debris for head loss testing with chemical effects should be based on break location 2. VEGP Response: Three head loss tests were completed for VEGP: thin bed, full load, and confirmatory full load. The full load produced the highest head loss at each stage of the test, as shown in the table below. Therefore, the full load test was utilized to develop the contributions from conventional debris, calcium phosphate, aluminum precipitates, and the 30-day extrapolation. Table 3.o.2-1 lists the chemical head loss contributions. ES-147

Enclosure 5 I Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Table 3.o.2-1: Chemical Head Loss Values Confirmatory Thin Bed Full Load Test Point Full Load (ft-H20) (ft-H20) (ft-H20)

onventional Debris 0.625 5.46 . 3.50 Full Calcium Phosphate Load 1.65 6.57 5.75 Full Aluminum Precipitate Load 2.60 11.81 8.99 130-day Extrapolation 3.15 15.70 11.12 See the Response to 3.f.10 for additional chemical head loss information.

3. Plant-Specific Materials and Buffers: Licensees should provide their assumptions (and basis for the assumptions) used to determine chemical effects loading: pH I. range, temperature profile, duration of containment spray, and materials expected to contribute to chemical effects.

VEGP Response: The VEGP chemical model requires a number of plant-specific inputs. Each input is chosen to maximize the calculated quantity and minimize the solubility (aluminum only) of the chemical precipitates. VEGP uses TSP to buffer the post-LOCA containment sump pool to a final pH between 7.12 and 7.78. In order to maximize chemical release, TSP is conservatively assumed to dissolve immediately, and the maximum pH of 7.8 was used for the containment sump pool for the entire 30-day event and for the containment spray while recirculating from the containment sump pool. A maximum pH of 5.72 was used for the containment spray during the post-LOCA RWST injection mode. To minimize aluminum solubility, the minimum containment sump pool pH of 7.0 was used, and precipitation was forced at 24 hours whether the solubility limit was reached or not. Different pH values for release and solubility were combined in a non-physical way, bounding the effects

'*'     of all potential pH profile variations. The pH values are summarized in Table 3.o.3-1:

E5-148

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Table 3.o.3-1: VEGP pH Values Sump and Recirculation Spray pH Used 7.8 To Determine Chemical Release Rates Maximum VEGP Long-Term Sump pH 7.78 Minimum VEGP Long-Term Sump pH 7.12 Sump pH Used To Determine Aluminum 7.0 Solubility Injection Spray pH Used To Determine 5.72 Chemical Release Rates The maximum temperature profiles of containment and the sump pool used for the analysis are from the double-ended pump suction LOCA with minimum safeguards case. The total amount of unsubmerged aluminum exposed to containment sprays was assumed to be 926.6 ft2. The total amount of submerged aluminum exposed to the containment sump fluid at VEGP Units 1 and 2 was assumed to be 348.4 ft2. The total amount of concrete assumed to be exposed and submerged in th.e containment sump pool is 10,000 ft2. The quantity of chemical precipitates is negligibly impacted by this large assumed surface area of exposed concrete. Therefore, exposed concrete is not a significant impact to chemical product generation in the VEGP post-LOCA containment sump pool and is not tracked for this purpose. The NARWHAL software (see Enclosure 3, Section 13.1 for general description of the software) analysis accounts for the change in water volume with respect to time and uses assumptions that minimize the water volume in containment. The water volume is dependent on break size, break location, and whether the containment sprays actuate, among other factors. The break size affects the volume in that the accumulators do not inject for breaks less than 2 inches. The RCS holdup volume is dependent on the break location/elevation. Finally, the containment spray activation affects the volume of water in transit. It is acknowledged that water volume has competing effects with respect to chemical release versus solubility; therefore, water volume was included in the sensitivities evaluated in Enclosure 3. A spray duration of 24 hours is used in the analysis. Although there is some operational response flexibility in the spray duration, this is a reasonably conservative assumption because sprays are required to operate for at least 2 hours if they are initiated (assuming there are no containment radiation monitor alarms). Phosphate inhibition minimizes the effect of chemical release from extended spray durations. Enclosure 3, Section 14.3 ES-149

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) includes sensitivity studies with both 2 hour and 30-day containment spray durations. Variable Debris Amounts Of the debris types described in WCAP-16530-NP-A, both E-Glass (Nukon) and lnteram fire barrier are present in the VEGP containments. The quantities of these materials are specific to each break analyzed. Figure 3.a.3-1 through Figure 3.a.3-8 show the ranges of Nukon and lnteram debris versus break size for DEGBs and partial breaks. The mass of latent fiber included as E-Glass for all breaks is 30 lbm. lnteram is a potential source of aluminum from its metal foil surface and is a source of leachable silicon (Reference 73). The lnteram fire barrier at VEGP uses stainless steel foil and is, therefore, not a source of aluminum. Additionally, as discussed in the Response to 3.o.2.7.i, silicon release is not tracked because aluminum is assumed to precipitate only as SAS. Therefore, lnteram does not impact the VEGP chemical model.

4. Approach to Determine Chemical Source Term (Decision Point): Licensees should identify the vendor who performed plant-specific chemical effects testing.

VEGP Response: VEGP is using the separate chemical effects approach to determine the chemical source term. Alien Science and Technology Corporation performed the testing in their test lab in Warrenville, IL.

5. Separate Effects Decision (Decision Point): Within this part of the process flow chart, two different methods of assessing the plant-specific chemical effects have been proposed. The WCAP-16530-NP-A study (Box 7 WCAP Base Model) uses predominantly single-variable test measurements. This provides baseline information for one material acting independently with one pH-adjusting chemical at an elevated temperature. Thus, one type of insulation is tested at each individual pH, or one metal alloy is tested at one pH. These separate effects are used to formulate a calculational model, which linearly sums all of the individual effects. A second method for determining plant-specific chemical effects that may rely on single-effects bench testing is currently being developed by one of the strainer vendors (Box 6, AECL).

VEGP Response: VEGP is primarily using the WCAP-16530-NP-A chemical effects base model to determine the chemical source term. Refinements to this model for aluminum solubility and phosphate inhibition of aluminum release from metallic aluminum are discussed in the Response to 3.o.2.8 and Response to 3.o.2.9.i. E5-150

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary)

6. AECL Model:

i. Since the NRG is not currently aware of the complete details of the testing approach, the NRG staff expects licensees using it to provide a detailed discussion of the chemical effects evaluation process along with head loss test results.

VEGP Response: This question is not applicable because VEGP is not using the AECL model. See the Figure 1 flow chart in "NRG Staff Review Guidance Regarding Generic Letter 2004-02 Closure in the Area of Plant-Specific Chemical Effect Evaluations" from March 2008 (Reference 106). ii. Licensees should provide the chemical identities and amounts of predicted plant-specific precipitates. VEGP Response: This question is not applicable because VEGP is not using the AECL model. See the Figure 1 flow chart in "NRG Staff Review Guidance Regarding Generic Letter 2004-02 Closure in the Area of Plant-Specific Chemical Effect Evaluations" from March 2008 (Reference 106).

7. WCAP Base Model:

i. Licensees proceeding from block 7 to diamond 10 in the Figure 1 flow chart

[in Enclosure 3 to a letter from the NRG to NEI dated March 2008 (Reference 106)] should justify any deviations from the WCAP base model spreadsheet (i.e., any plant specific refinements) and describe how any exceptions to the base model spreadsheet affected the amount of chemical precipitate predicted. VEGP Response: The VEGP chemical model includes quantification of chemical precipitates using the WCAP-16530-NP-A (Reference 73) methodology with refinements for phosphate passivation of aluminum surfaces and the application of an aluminum solubility correlation to determine formation temperature/timing. Silicon inhibition of aluminum release is not credited. Refinements to this model for aluminum solubility and phosphate inhibition of aluminum release from metallic aluminum are discussed in the Response to 3.o.2.9.i. The chemical precipitates assumed by the VEGP chemical model are calcium phosphate (Ca3(PQ4)2) and SAS (NaAISbOa). Although the WCAP-16530-NP-A model typically includes aluminum oxyhydroxide ES-151

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) (AIOOH), the VEGP model assumes that all precipitated aluminum forms SAS, independent of the quantity of silicon available. Per the WCAP-16530-NP-A SE, both aluminum precipitates are acceptable surrogates for aluminum precipitate in head loss testing. As described in Enclosure 3, Section 8.0, the chemical quantity methodology is integrated into the NARWHAL software. The quantity of chemical precipitate is determined for each break as a function of thermal hydraulic inputs and debris generation quantities. The temperature profiles, pH profiles, aluminum metal surface area, and concrete surface area are constant for each break and were selected to maximize aluminum and calcium release and minimize aluminum solubility. The sump fluid volume, spray duration, and debris quantities are break-dependent variables in the NARWHAL calculations. There are two parts to the determination of the chemical precipitate quantity: the elemental chemical release from substrates in containment and chemical product formation. Elemental Chemical Release The two classifications of substrates for which chemical release is analyzed are debris and exposed surfaces. Fiber debris and lnteram debris contribute to chemical release, which is quantified using the WCAP-16530-NP-A release equations. Note that the quantity of each of these debris types is break-specific; therefore, the quantity of elemental chemical release will vary for each break analyzed. Also, note that the lnteram only releases silicon, which does not contribute to the chemical precipitates being tracked for VEGP (see chemical product formation discussion below). This is because SAS is the only aluminum precipitate that is being tracked in the VEGP NARWHAL model, and NARWHAL conservatively assumes an infinite source of silicon when SAS is the only aluminum precipitate tracked. The amount of elemental chemical release from a given debris source is limited by the quantity. Table 3.o.2.7-1 shows the chemical limits of fiberglass used in the chemical release model. Table 3.o.2.7-1: Chemical Mass Limits E-Glass Aluminum Mass Available per Material Mass 1.95% Calcium Mass Available per Material Mass 2.16% The exposed surfaces include aluminum metal and concrete surfaces that either are submerged in the containment pool or are exposed to containment E5-152

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) sprays. The same surface areas were analyzed for each break. Table 3.o.2.7-2 shows these areas. Note that the chemical release from exposed concrete was evaluated using the WCAP-16530-NP-A release equations, and the chemical release from aluminum was evaluated using the University of New Mexico (UNM) release equations. Table 3.o.2.7-2: Other Source Release Values Surface Area Substrate (ft2) Aluminum Metal: Submerged 348.4 Aluminum Metal: Unsubmerged 926.6 Exposed Concrete: Submerged 10,000 Exposed Concrete: Unsubmerged 0 Chemical Product Formation The chemical precipitates analyzed for the VEGP NARWHAL CFP calculation are SAS and calcium phosphate. The calcium phosphate is conservatively assumed to precipitate immediately. The SAS precipitates when the concentration of aluminum in the pool exceeds the aluminum solubility limit as calculated with the Argonne National Laboratory (ANL) solubility equation. Note that if precipitation of SAS is not predicted before 24 hours, then precipitation is forced at that time. Also, note that aluminum does not remain dissolved in the pool after precipitation occurs. Forcing precipitation at 24 hours and not taking credit for aluminum remaining dissolved in the pool are conservative factors in the chemical product formation model. ii. Licensees should list the type (e.g., AIOOH) and amount of predicted plant-specific precipitates. VEGP Response: Chemical precipitate quantities were calculated in the NARWHAL CFP calculation and in a bounding hand calculation. The NARWHAL calculation performs comprehensive evaluation of GSl-191 phenomena in a self-consistent and time-dependent manner. It should be noted that the chemical debris quantities used for quantifying head loss were directly calculated in NARWHAL, not from the hand calculation. The NARWHAL calculation uses the plant-scale precipitate loads from the 2009 head loss testing as the maximum debris limit acceptance criteria (see Response to 3.f.5). The results from the hand calculation are provided below as bounding numbers. The bounding precipitate surrogate masses that would be generated at VEGP are 40.2 kg (88.6 lbm) SAS and 63.2 kg (139.3 lbm) calcium phosphate. ES-153

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) These precipitate masses represent the bounding quantity of aluminum and calcium that could precipitate in the VEGP Unit 1 and Unit 2 containment sump pool. Both AIOOH and SAS chemical surrogates are considered equivalent generators of head loss across a debris bed. Therefore, AIOOH and SAS surrogates may be substituted for each other stoichiometrically relative to aluminum. The maximum temperature where aluminum precipitation could occur in the containment sump pool was calculated to be 136.8 degrees F. Furthermore, the use of the aluminum solubility model for a TSP buffered solution is supported by the integrated autoclave experiments, as shown in Table 3.o.2.9-2. However, as the duration of these experiments was only 24 hours, aluminum precipitation is assumed to occur at 24 hours for calculating strainer head loss unless the solubility model predicts precipitation at an earlier time. Calcium phosphate is assumed to precipitate at all temperatures. These results are bounding for both Unit 1 and Unit 2.

8. WCAP Refinements: State whether refinements to WCAP-16530-NP-A were utilized in the chemical effects analysis.

VEGP Response: The chemical effects strategy for VEGP includes quantification of chemical precipitates using the WCAP-16530-NP-A (Reference 73) methodology with refinements for phosphate passivation of aluminum surfaces and the application of an aluminum solubility correlation to determine formation temperature/timing. Silicon inhibition of aluminum release was not credited. Refinements to the model for aluminum solubility and phosphate inhibition of aluminum release from metallic aluminum are discussed in the Response to 3.o.2.9.i.

9. Solubility of Phosphates. Silicates and Al Alloys:

i. Licensees should clearly identify any refinements (plant-specific inputs) to the base WCAP-16530-NP-A model and justify why the plant-specific refinement is valid.

VEGP Response: The VEGP chemical model includes quantification of chemical precipitates using the WCAP-16530-NP-A (Reference 73) methodology with refinements for phosphate passivation of aluminum surfaces and the application of an aluminum solubility correlation to determine formation temperature/timing. Silicon inhibition of aluminum release was not credited. ES-154

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 {Non-Proprietary) Phosphate Inhibition of Aluminum Surfaces The release of aluminum from metallic aluminum material into the TSP sump solution was modeled using the equations developed by Howe et al. Equations 3.o.2.9-1 through 3.o.2.9-3 (UNM aluminum release equations) calculate the release rate for aluminum from sprayed and submerged aluminum metal in containment. (Equation 3.o.2.9:-1) 6.2181-4.6454 1000 +1.7716(pH)-1.95SO(pH)TK Rmax = 10 TK 1000 (Equation 3.o.2.9-2) (Equation 3.o.2.9-3) Nomenclature: RAI,m = release rate of aluminum from aluminum metal, mg/(m 2 min) Rmax = non-passivated aluminum release rate, mg/(m 2 min) P = phosphate passivation term, unitless TK =temperature, K TKP =temperature utilized in the phosphate passivation term, K pH = pH at 25 degrees C tTsP = time elapsed with phosphate present in solution, min The above equations were developed in testing that was performed at temperatures from 55 degrees C to 85 degrees C (131 degrees F to 185 degrees F, 328.15 K to 358.15 K) and at pH values from 6.84 to 7.84. The following two constraints were used to extend the applicability of these equations:

1. The passivation term, P, is an exponential decay function that approaches zero as trsP increases. This term models the decrease in aluminum surface area available for release as the passivation layer forms. Since testing was not performed below a pH of 6.84, it is not known if this term is applicable at very low TSP concentrations. Therefore, phosphate is not considered "present in solution" unless the pH is above 6.84, and trsP is held at 0 minutes. When the pH rises above 6.84, as TSP dissolves into solution, the trsP "clock" starts.

In practice, the VEGP analysis assumes that TSP is present in the containment sump pool at the start of the LOCA by assuming that the initial sump pH is at its maximum value of 7.8 (see Response to 3.o.2.3). This assumption conservatively increases the aluminum release rate by non-physically combining the highest sump pH with the high initial sump E5-155

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) temperatures. Similarly, the containment sprays are assumed to be at a pH of 7.8 (TSP is present) immediately upon the switch to recirculation.

2. The phosphate passivation function is not assumed to extrapolate beyond the test temperature range. Therefore, T KP equals T K unless above or below the temperatures described here. For temperatures below 328.15K, T KP equals 328.15K. For temperatures above 358.15 K, T KP equals 358.15 K. The passivation equation predicts faster passivation both above and below this range, which is not justifiable without additional testing.

A comparison of the VEGP chemical model with the WCAP-16530-NP-A model is shown in Figure 3.o.2.9-10 and Figure 3.o.2.9-11. Additionally, the aluminum release rate equations with the above constraints were verified for use by VEGP by modeling several integrated autoclave tests. Test 40-01 used VEGP specific materials in testing to replicate the post-LOCA containment debris amounts reported for VEGP Units 1 and 2. Given that the relevant VEGP specific design inputs are a maximum sump pH of 7.78 and a post-LOCA debris type of E-Glass (i.e., no Calcium Silicate, Silica, or Mineral Wool insulation), tests 39-01, 42 01, and 44-01 also use test inputs similar to that of VEGP [

                  ]a,c These tests were run at a range of temperatures from [
                          ]a,c which bounds the VEGP maximum post-LOCA containment sump pool and containment temperature profiles. Therefore, tests 39-01 (in-bag and out-of-bag), 40-01, 42-01, and 44-01 were simulated using the WCAP-16530-NP-A methodology with the refined aluminum release equation. The critical parameters for the integrated autoclave tests 39-01, 40-01, 42-01, and 44-01 are summarized in Table 3.o.2.9-1 and Figure 3.o.2.9-1.

ES-156

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) a,c Figure 3.o.2.9-1: Integrated Autoclave Test Temperature Profiles Table 3.o.2.9-1: Critical Integrated Autoclave Test Parameters a,c E5-157

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Vogtle's maximum pH (7.78) and the pH of test 44-01 [ ]a,c are within the pH range for the UNM aluminum release equations of 6.84 through 7.84. Therefore, test 44-01 serves as the primary validation of the UNM aluminum release equations in an integrated chemical environment. The aluminum concentration results for test 40-01 and the simulation is provided in Figure 3.o.2.9-2. a,c Figure 3.o.2.9-2: Test 44-01 (pH = 7.52) Aluminum Concentrations Because the maximum pH of Test 44-01 [ ]a,c is below the VEGP maximum pH (7.78), integrated autoclave tests 39-01 [ ]a,c 40-01 [

            ]a,c and 42-01 [       ]a,c were also simulated to validate the UNM aluminum release equations at bounding pH values. However, these tests are above the pH range for the single effects tests (6.84 through 7.84) used to develop the UNM aluminum release equations. The aluminum concentration results for these tests and their simulations are provided in Figures 3.o.2.9-3 through 3.o.2.9-5.

ES-158

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) a,c Figure 3.o.2.9-3: Test 39-01 (pH = 8.00) Aluminum Concentrations a,c Figure 3.o.2.9-4: Test 40-01 (pH = 8.03) Aluminum Concentrations ES-159

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) a,c Figure 3.o.2.9-5: Test 42-01 (pH = 8.03) Aluminum Concentrations In addition to the aluminum results, the integrated autoclave test calcium release results were also compared with the simulation in Figures 3.o.2.9-6 through 3.o.2.9-9 to demonstrate the overall conservatism of the VEGP chemical model. ES-160

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) a,c Figure 3.o.2.9-6: Test 39-01 (pH = 8) Calcium Concentrations a,c I : Figure 3.o.2.9-7: Test 40-01 (pH = 8.03) Calcium Concentrations ES-161

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) a,c Figure 3.o.2.9-8: Test 42-01 (pH = 8.03) Calcium Concentrations a,c Figure 3.o.2.9-9: Test 44-01 (pH = 7.52) Calcium Concentrations E5-162

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) The aluminum release results indicate that the aluminum release rate as a function of time in a TSP buffered solution decreases to approximately zero within 24 hours, as is predicted by the aluminum release equations. For integrated autoclave tests 40-01, 42-01, and 44-01, the measured aluminum concentrations either follow or are bounded by the trends predicted by the UNM aluminum release equations, which justifies the UNM methodology with the extended pH (5 to 7.84) and temperature (104.8 degrees F to 266.5 degrees F) constraints as acceptable. Note that the aluminum concentration for Test 39-01 is under-predicted. The presence of zinc in solution can reduce the rate of aluminum corrosion. A reduction of approximately 2/3 of the released aluminum concentration was observed in the Howe et al. bench tests when zinc coupons were present. The Howe et al. equations do not credit zinc inhibition, which was shown by bench testing to result in an over-prediction of aluminum release under the maximum pH tested of 7.84. However, integrated autoclave tests 39-01, 40-01, and 42-01 are above the pH range for the single effects tests. Of these three integrated autoclave tests, Test 39-01 contained the least amount of galvanized steel I ,. and contained less than 38% (0.117 ft2/0.308 ft 2 ) of the galvanized steel ' ' surface area as Test 40-01, which used VEGP-specific material quantities. Furthermore, the aluminum release result for Test 40-01 is accurately predicted, and Test 42-01 is slightly over-predicted with 238% of the VEGP-specific galvanized steel quantity (0.732 ft2/0.308 ft 2 ). Although these results demonstrate that the aluminum release equations accurately predict aluminum concentrations at elevated pH when VEGP-specific or greater zinc quantities are present, the maximum acceptable pH is not extended above the value of 7.84 as set by Howe, et al. Additionally, as shown in Figures 3.o.2.9-6 through 3.o.2.9-9 calcium concentrations are significantly over-predicted by the WCAP-16530-NP-A model. As discussed in Section 3.o.2.7.i, calcium phosphate is conservatively assumed to form immediately as calcium is released. Finally, as discussed in Section 3.f.10, the full load calcium phosphate head loss is assumed as soon as calcium phosphate starts to accumulate on the strainer. Because the VEGP chemical model results in over-predicted quantities of aluminum and calcium precipitates at VEGP-specific conditions, the overall methodology is conservative for use to determine the precipitate loading for strainer head loss. Solubility of Aluminum The aluminum solubility limit is determined using Equation 3.o.2.9-4, developed by ANL. 26980

• 10(pH+LlpH)-14.4+0.0243T if T < 175 op c - , - (Equation 3.o.2.9-4)

Al.sol - { 26980 . lO(pH+LlpH)-10.41+0.00148T 1 if T > 175 op E5-163

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Nomenclature: L\pH =pH change due to radiolysis acids T =solution temperature, degrees F The aluminum solubility limit equation was used to determine the temperature and timing of aluminum precipitation and to determine the aluminum concentration in solution for use in the aluminum release equations for concrete and insulation. When precipitation is predicted by this equation, the full amount of aluminum released is assumed to precipitate as SAS. The aluminum solubility limit equation was not used to reduce the predicted quantity of precipitate by crediting the amount remaining in solution. Aluminum solubility testing developed by ANL was completed in boric acid/NaOH buffered solutions. As shown in Table 3.o.2.9-2, the method that Vogtle utilizes to predict aluminum precipitation temperature yields much higher temperatures than the filtration tests that did not detect precipitation. This demonstrates the applicability of the ANL equation in a boric acid/TSP buffered solution. Since the tests were performed for a 24-hour duration, the maximum amount of time allowed in the VEGP chemical model for aluminum precipitation to occur was capped at 24 hours. Table 3.o.2.9-2: Aluminum Precipitation Test Results a,c Comparison of VEGP Chemical Model with WCAP-16530-NP-A Enclosure 3 Section 14.4 includes a sensitivity study comparing the VEGP release model with phosphate inhibition credited (UNM Aluminum Metal Release Equation) to the WCAP-16530-NP-A chemical model without this refinement (WCAP-16530 Equation). Additionally, the bounding VEGP case for maximum precipitate generation (based on the hand calculation) was conducted using both the VEGP release model and the WCAP-16530-NP-A model without the refined equations for phosphate inhibition. Note that to address NRC concerns on the use of the WCAP-16530-NP-A for aluminum ES-164

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) metal release rates early in the event, the aluminum metal release rate was doubled for the initial 15 days. The maximum precipitation case uses the pH, temperature, aluminum metal, and concrete inputs described in the VEGP response to 3.o.2.3. The maximum sump pool mass of 6,675,858 lbm was used to maximize the dissolution of aluminum and calcium from concrete and insulation materials. The containment sprays were assumed to remain active for 30 days with a minimum RWST injection duration of 27.9 minutes to maximize aluminum release from sprayed aluminum surfaces. Lastly, the maximum generated quantity of Nukon insulation, 2,229.2 ft3, was used. Note that these inputs apply to this comparison completed in the hand calculation, which were selected to yield bounding results for all break scenarios as opposed to the break-specific and time dependent inputs used in the NARWHAL CFP calculation. Figure 3.o.2.9-10 shows a comparison of the aluminum concentration results for the two models over the full 30-day window, and Figure 3.o.2.9-11 shows the results for the first 24 hours. The UNM curve shows the results with aluminum passivation and the 2xWCAP-16530-NP-A curve shows the results without this refinement. 5 _,.... ---*- **-* *- *-*-. -**-* ---*-**- - . 4.5 4 ................. . ..... -~-.-. *-**** ......, l _3.5 E c.

      .e

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                 ,-----:------:-------------;------ ------1
                                      ;                  I

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0.5 0 7200 14400 21600 28800 36000 43200 Time(min)

                                                     - -       UNM        - 2 x WCAP*16530*NP*A Figure 3.o.2.9-10: Maximum Aluminum Release Cases (30 days)

ES-165

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) 2 1.75 1.5 - *---- *-*:*"" .......- ...- i ..-- *--*-*-*-,----- ---.

        "Ec.                                                  - - ~ - - - - i - - - - -~' - - - - - - - -
        ..8: 1.25 I:
                                                                                       *****- +-*-**-*------ ----r *----*~- *-

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                                                              .... -- . .-..1 ..

I 0.5 ' 0.25 *-* -**-T*--* ' -----*-***< --- ***-- -*-- r*- --***---* 0 0 180 360 540 720 900 1080 1260 1440 Time(min)

                                 - - UNM               - 2 x WCAP-16530-NP-A Figure 3.o.2.9-11: Maximum Aluminum Release Cases (24 hours)

The use of the aluminum passivation equations clearly decreases the release of aluminum predicted over 30 days. However, as would be expected, the initial release rate with passivation credited (UNM) follows the model without passivation credited (2xWCAP-16530-NP-A) closely over the initial two hours before diverging as the aluminum metal surface area available for release passivates. ii. For crediting inhibition of aluminum that is not submerged, licensees should provide the substantiation for the following: (1) the threshold concentration of silica or phosphate needed to passivate aluminum, (2) the time needed to reach a phosphate or silicate level in the pool that would result in aluminum passivation, and (3) the amount of containment spray time (following the achieved threshold of chemicals) before aluminum that is sprayed is assumed to be passivated. VEGP Response: See the Response to 3.o.2.9.i. ES-166

                                       .Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) iii. For any attempts to credit solubility (including performing integrated testing),

licensees should provide the technical basis that supports extrapolating solubility test data to plant-specific conditions. In addition, licensees should indicate why the overall chemical effects evaluation remains conservative when crediting solubility given that small amount of chemical precipitate can produce significant increases in head loss. VEGP Response: See the Response to 3.o.2.9.i. iv. Licensees should list the type (e.g., AIOOH) and amount of predicted plant-specific precipitates. VEGP Response: See the Response to 3.o.2.7.ii.

10. Precipitate Generation (Decision Point): State whether precipitates are formed by chemical injection into a flowing test loop or whether the precipitates are formed in a separate mixing tank.

VEGP Response: As discussed in the Response to 3.o.2.12, VEGP pre-mixed surrogate chemical precipitates in a separate mixing tank for chemical head loss testing. The direct chemical injection method was not used in head loss testing.

11. Chemical Injection into the Loop:

i. Licensees should provide the one-hour settled volume (e.g., 80 ml of 100 ml solution remained cloudy) for precipitate prepared with the same sequence as with the plant-specific, in-situ chemical injection.

VEGP Response: As discussed in the response to item 3.o.2.12, VEGP pre-mixed surrogate chemical precipitates in a separate mixing tank for chemical head loss testing. The direct chemical injection method was not used in head loss testing. See the Figure 1 flow chart in "NRC Staff Review Guidance Regarding Generic Letter 2004-02 Closure in the Area of Plant-Specific Chemical Effect Evaluations" from March 2008 (Reference 106). ES-167

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) ii. For plant-specific testing, the licensee should provide the amount of injected chemicals (e.g., aluminum), the percentage that precipitates, and the percentage that remains dissolved during testing. VEGP Response: See the Response to 3.o.11.i. iii. Licensees should indicate the amount of precipitate that was added to the test for the head loss of record (i.e., 100 percent, 140 percent of the amount calculated for the plant). VEGP Response: See the Response to 3.o.11.i.

12. Pre-Mix in Tank: Licensees should discuss any exceptions taken to the procedure recommended for surrogate precipitate formation in WCAP-16530-N P-A.

VEGP Response: The chemical head loss tests employed the pre-mixed chemical surrogate

• methodology. The WCAP-16530-NP-A precipitate formation methodology for SAS and calcium phosphate was followed with no exceptions. The 1-hour settling volume for each batch of chemical precipitates was determined at the time that the batch was produced and was required to be 6 ml or greater. The chemical precipitate settling was also required to be measured within 24 hours of the time the surrogate was to be used, and the 1-hour settled volume was required to be 6 ml (SAS and AIOOH) or greater and within 1 .5 ml of the freshly prepared surrogate (Reference 73). Chemical precipitates that failed the criteria of being 6 ml or greater (initial test or re-test) and within 1.5 ml of the freshly prepared surrogate criteria were not used in testing.

13. Technical Approach to Debris Transport (Decision Point): State whether near-field settlement is credited or not.

VEGP Response: VEGP chemical effects testing used agitation and turbulence in the test tank to ensure that essentially all debris analyzed to reach the strainer in the plant reached the strainer in. head loss testing.

                                         .    . VEGP did not credit any near field settlement in head loss testing.

ES-168

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary)

14. Integrated Head Loss Test with Near-Field Settlement Credit:

i. Licensees should provide the one-hour or two-hour precipitate settlement values measured within 24 hours of head loss testing.

VEGP Response: VEGP is not crediting near field settlement of chemical precipitate in chemical head loss testing. See the Figure 1 flow chart in "NRC Staff Review Guidance Regarding Generic Letter 2004-02 Closure in the Area of Plant-Specific Chemical Effect Evaluations" from March 2008 (Reference 106). ii. Integrated Head Loss Test with Near-Field Settlement Credit: Licensees should provide a best estimate of the amount of surrogate chemical debris that settles away from the strainer during the test. VEGP Response: See the Response to 3.o.2.14.i.

15. Head Loss Testing Without Near Field Settlement Credit:

i. Licensees should provide an estimate of the amount of debris and precipitate that remain~ on the tank/flume floor at the conclusion of the test and justify why the settlement is acceptable.

VEGP Response: Even though all debris had an opportunity to collect on the surfaces of the test strainer, a portion of the debris added to the test settled on the floor. Post-test photographs show that while nearly all of the Nukon had reached the strainers, approximately 10-15 percent of the (larger/heavier) dirt/dust surrogate and nearly all of the lnteram debris settled on the floor in front of the strainer array (see Attachment A). Additionally, a minor amount (approximately 10-20 lb.) of silicon carbide settled underneath the simulated containment floor, and less than 1.5 lbm of particulate debris was removed along with the water drained from the tank to ensure sufficient volume in the tank for chemical additions. Finally, a small amount of calcium phosphate (less than 0.25 L of the 480.72 L total) was spilled outside the tank such that it was unrecoverable. Because of the measures taken during the test, as described in Response to 3.f.12, to keep debris suspended and transportable to the test strainer, the amount of settled debris described above is considered acceptable. E5-169

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) ii. Licensees should provide the one-hour or two-hour precipitate settlement values measured and the timing of the measurement relative to the start of head loss testing (e.g., within 24 hours). VEGP Response: See the Response to 3.o.2.12.

16. Test Termination Criteria: Licensees should provide the test termination criteria.

VEGP Response: The head loss was considered stable when the differential pressure across the debris bed changed by less than or equal to 1 percent over a 1-hr period. In addition, the rate of head loss increase was required to be significantly decreasing, or the head loss was required to be consistently steady at termination of the test. ES-170 __I

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary)

17. Data Analysis:

i. Licensees should provide a copy of the pressure drop curve(s) as a function of time for the testing of record .

VEGP Response: The pressure drop curves for the full load test are provided as Figures 3.o.2.17-1 through 3.o.2.17-4 below. Day 1

• Head Loss & Approach Velocity Profile

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I- Head Loss -*-* Flow Adjustment - - - *Drain Down ---- Tank Fill - Velocity I Figure 3.o.2.17-1: Full Load Test Differential Pressure and Velocity vs. Time - Day 1 ES-171

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Day 2

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• Flow Adjustment - Velocity I Figure 3.o.2.17-3: Full Load Test Differential Pressure and Velocity vs . Time - Day 3 E5-172

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Day 4 - Head Loss & Approach Velocity Profile 16.0 ~----~-----~~~---~--~~-~~~ 0 . 018 14.0 120 l!l 10.0 "3:

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2.0 0.002 0.0 +----~-'----~-------r-~--~~--~r-'---'--~-+ 0.000 11:50:00 PM 3:26:00AM 7:02:00 AM 10:38:00 AM 2:14:00 PM 5:50:00 PM 9:26:00 PM Time (H:MM:SS AM/PM) I- Head Loss - * - -Flow Adjustment - Velocity I Figure 3.o.2.17-4: Full Load Test Differential Pressure and Velocity vs. Time - Day 4 ii. Licensees shou ld explain any extrapolation methods used for data analysis . VEGP Response: See the Response to 3.f.10.

18. Integral Generation (Alien) : Licensees should explain why the test parameters (e.g. , temperature , pH) provide for a conservative chemical effects test VEGP Response :

VEGP is using the separate chemical effects approach to determine the chemical source term . Th is section is not applicable to the VEGP chemical effects analysis. E5-173

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) *

19. Tank Scaling I Bed Formation:

i. Explain how scaling factors for the test facilities are representative or conservative relative to plant-specific values.

VEGP Response: VEGP is using the separate chemical effects approach to determine the chemical source term. This section is not applicable to the VEGP chemical effects analysis. ii. Explain how bed formation is representative of that expected for the size of materials and debris that is formed in the plant specific evaluation. VEGP Response: VEGP is using the separate chemical effects approach to determine the chemical source term. This section is not applicable to the VEGP chemical effects analysis.

20. Tank Transport: Explain how the transport of chemicals and debris in the testing facility is representative or conservative with regard to the expected flow and transport in the plant-specific conditions.

VEGP Response: VEGP is using the separate chemical effects approach to determine the chemical source term. This section is not applicable to the VEGP chemical effects analysis.

21. 30-Day Integrated Head Loss Test: Licensees should provide the plant-specific test conditions and the basis for why these test conditions and test results provide for a conservative chemical effects evaluation.

VEGP Response: VEGP is using the separate chemical effects approach to determine the chemical source term. This section is not applicable to the VEGP chemical effects analysis. ES-174 J

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary)

22. Data Analysis Bump Up Factor: Licensees should provide the details and the technical basis that show why the bump-up factor from the particular debris bed in the test is appropriate for application to other debris beds.

VEGP Response: VEGP is using the separate chemical effects approach to determine the chemical source term. This section is not applicable to the VEGP chemical effects analysis.

p. Licensing Basis The objective of the licensing basis section is to provide information regarding any changes to the plant licensing basis due to the sump evaluation or plant modifications.

1. Provide the information requested in GL 2004-02 Requested Information Item 2(e) regarding changes to the plant-licensing basis. The effective date for changes to the licensing basis should be specified. This date should correspond to that specified in the 10 CFR 50.59 evaluation for the change to the licensing basis.

GL 2004-02 Requested Information Item 2(e) A general description of and planned schedule for any changes to the plant licensing bases resulting from any analysis or plant modifications made to ensure compliance with the regulatory requirements listed in the Applicable Regulatory Requirements section of this GL. Any licensing actions or exemption requests needed to support changes to the plant licensing basis should be included. Response to 3.p.1: VEGP is following the "STP Piloted Risk-Informed Approach for GSl-191," (References 44 and 45). The proposed change replacing the current deterministic methodology with a risk-informed methodology requires changes to the descriptions of how VEGP meets 10 CFR 50.46(a)(1 ), GDC 35, GDC 38, and GDC 41. Those changes require exemptions to certain requirements of 10 CFR 50.46(a)(1 ), GDC 35, GDC 38, and GDC 41, and the requests for the exemptions are provided in the future LAR. VEGP's risk-informed approach to assess the effects of LOCA debris replaces the existing deterministic approach described in the VEGP licensing basis. This, in turn, requires an amendment to the VEGP Unit 1 and Unit 2 operating licenses to incorporate the revised methodology per the requirements of 10 CFR 50.59. This proposed amendment to the operating license is included in the future LAR. ES-175

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) 4.0 NRC Request for Additional Information: VEGP received two requests for additional information from the NRC: NL-08-1497 and NL-08-1829 (References 101 and 87, respectively). The first RAI, issued September 17, 2008, addressed critical issues with test protocols used in chemical effects testing performed at the VUEZ facility by ALION Science and Technology. SNC responded to the VUEZ-specific RAls with the issuance of NL-08-1583 (Reference 102), November 7, 2008. SNC determined the need to consider an alternate approach to demonstrating adequate performance of the emergency containment sump strainers after careful consideration of the NRC's concerns with the VUEZ test protocol. Therefore, the VUEZ-specific RAls are no longer applicable and do not require a response, as recorded in NL-08-1583. The second RAI, NL-08-1829, was issued December 2, 2008, containing 29 requests based on the review of all four VEGP Supplemental Responses to GL 2004-02: NL-07-1777, NL-08-0670, NL-08-1155, and NL-08-1228 (References 95, 96, 98, and 100, respectively). The final SNC responses to RAls 1 through 29 are referenced in the table below. Additionally, SNC provided its intended path forward for the resolution of GSl-191 letter to the NRC, NL-13-0953(Reference105). By letter dated November 14, 2013, the NRC sent SNC an RAI. The RAI and SNC's response are included below as provided in SNC letter to the NRC, NL-13-2544 (Reference 110). SNC has since revised the ERGs as discussed in the response below. NRC RAI SN C's May 16, 2013, letter did not identify that SNC had implemented, or identified for future implementation, any mitigative measures to deal with the potential for in-vessel blockage. SNC stated that it was evaluating Westinghouse recommendations for mitigative measures for in-vessel blockage and that appropriate procedure changes and operator training would be completed, as deemed necessary, following the evaluation. Please provide the mitigative measures chosen for the VEGP to deal with in-vessel blockage, should it occur. SNC Response to NRC RAI SNC will make changes to the Vogtle Units 1 and 2 Emergency Response Guidelines (ERGs). Specifically, after transferring to cold leg recirculation, Vogtle will monitor core exit temperatures, injection flow, and reactor vessel level indicating system (RVLIS) indications to identify any abnormal indications. Should abnormal indications exist, realignment to hot leg recirculation and back flowing through the reactor vessel will be necessary. ES-176 L

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) RAI Complete RAls as Provided in NL-08-1829 (Reference 87) Response No. 1 Please provide a description of the jacketing/banding systems 17.0D ZOI is used used to encapsulate Nukon insulation at Vogtle (e.g., on piping, for all Nukon. steam generators, reactor coolant pumps) and during WCAP-16710-P jacketing/banding system qualification testing. The information is no longer should include the jacket materials used in the testing, geometries credited. and sizes of the targets and jet nozzle, and materials used for Because we are jackets installed in the plant. Please provide information that no longer compares the mechanical configuration and sizes of the test crediting the targets and jets, and the potential targets and two-phase jets in the jacketing in the plant. Pleases evaluate how any differences in jet/target sizing ZOI, a response and jet impingement angle affect the ability of the insulation system to this portion of to resist damage from jet impingement. In doing so, please provide the question is a justification for applying debris generation test data obtained for not provided. the Nukon jacketing systems employed at the Wolf Creek and Callaway plants to the jacketing systems used at Vogtle and See the demonstrating that the Vogtle jacketing systems are as resistant to Response to destruction as the jacketing systems tested. In responding to this 3.c.1. question, please address the potential varied jacketing systems for various components of the reactor coolant system, which are within the LOCA ZOls (e.g., piping, coolant pumps, and steam generators). 2 Please specify the ZOI radius used to calculate the quantity of See the lnteram fire barrier debris that could be generated. Please provide Response to the characteristics of the lnteram fire barrier material including the 3.b.1, 3.c.1, and type of lnteram installed and its anticipated debris characteristics. 3.f.4. Please provide information on how the material was prepared for inclusion in head loss testing or provide information on the surrogate material used and its properties. Please provide assumptions made regarding the physical properties of LOCA damaged lnteram fire barrier material and the bases for how any surrogate material used in testing conservatively model these properties. 3 Please provide the following additional information needed to See the support the assumption of 15% erosion of fibrous debris pieces in Response to the containment pool: 3.e.1.

a. The similarity of the flow conditions (velocity and turbulence),

chemical conditions, and fibrous material present in the erosion Note that a tests to the analogous conditions applicable to the expected smaller fraction plant conditions, and for erosion of

b. The durations of the erosion tests and how the test results were fibrous debris extrapolated to the sump performance mission time. pieces in the containment pool is used based on testing.

E5-177

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) RAI Complete RAls as Provided in NL-08-1829 (Reference 87) Response No. 4 On pages E1-19 and E1 -20 of the supplemental response dated No longer February 28, 2008, it is indicated that, based on the fact that less applicable than 25% of the strainer perimeter area is in excess of the curb lift because the curb velocity metric, 25% of small debris pieces are assumed to is not being surmount the curb/plenum on which the strainer modules rest. credited as a However, based on the diagram of containment provided on page debris interceptor. E1-9 of the same supplemental response, the staff expects that it Any debris that is likely that flow and debris will preferentially approach the sump transported I from openings in the shield wall.. As a result, the fraction of debris during f* approaching the sump in the higher velocity flow channel could recirculation was I significantly exceed 25%. In light of the considerations such as assumed to reach this, please provide a technical justification for the assumption that the strainers. I only 25% of small pieces of fibrous debris can surmount the curb/plenum and reach the sump strainers. See the Response to 3.e.4 and 3.e.6. 5 The supplemental response states that the head loss test results See the were scaled to the full-sized strainer system based on Response to temperature, velocity, and bed thickness differences. Without 3.f.10. additional information on the methodology used to make these extrapolations, it is not possible to determine whether they were performed conservatively or prototypically. It appears that the head loss test result of 6.84 ft was extrapolated to 8.126 ft. Please provide the details of all extrapolations performed for the head loss test data. Please include the raw test data and conditions, and the final head loss value and the conditions to which it was extrapolated. Please include any differences in temperature, velocity, bed thickness between the head loss testing and anticipated plant conditions. ES-178 J

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) RAI Complete RAls as Provided in NL-08-1829 (Reference 87) Response No. 6 The supplemental response stated that the submergence value for See the the SBLOCA was not calculated. The stated [reason] for this was Response to 3.f.2 that an SBLOCA would create less debris and therefore result in a and 3.f.3. less challenging head loss. This is true for some portions of the evaluation. However, the vortexing evaluation is often most limiting when there is no debris on the strainer. In addition, if the strainer is not fully submerged, the acceptance criteria for maximum head loss may be limited by the strainer height (50% of the strainer height per RG 1.82, Rev. 3) instead of the pump NPSH margin. This would be a reduction to about 25% of the tested strainer head loss. Also, if the strainer is not submerged, air ingestion would have to be evaluated more rigorously. Un-submerged strainer area cannot be credited to accumulate debris, so other areas of the strainer would have to absorb the debris that cannot be collected on the uncovered portion of the strainer. Due to break location, the SBLOCA level may not include some RCS inventory and also may not include all or part of the accumulator volume. Please provide the minimum submergence for an SBLOCA. If the strainer is not fully submerged for this event, please provide appropriate evaluations for air ingestion and strainer head loss, including acceptability based on the guidance in RG 1.82 (or other appropriate methodology). 7 Related to the RAI above on the response of the plant to cases See the where the strainers may not be fully submerged, have various Response to scenarios such as an SBLOCA with the failure of one train of EGGS 3.g.6 and no CS actuation been considered? For this case all debris would transport to a single EGGS strainer that may not be fully submerged. A thin bed with the bulk of the particulate debris could form on the operating strainer surface. Please provide an evaluation that demonstrates that adequate NPSH margin will be provided to the EGGS pumps (reference Regulatory Guide 1.82, Revision 3, Section 1.3.4.4). 8 It was implied that the debris was added to the testing prior to Debris was added starting the recirculation pump. Please provide justification that after starting the this test sequence provides prototypical or conservative test reci rcu latio n conditions. pump. See the Response to 3.f.4. ES-179

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) RAI Complete RAls as Provided in NL-08-1829 (Reference 87) Response No. 9 Scaling was based on the circumscribed area for the test and plant Scaling was strainers. Normally scaling is based on the screen area. For the based on screen module testing the scaling factors based on circumscribed and area for test and screen area appear to be the same. The scaling factors for the plant strainers. sector testing could not be determined by the staff. Please provide information that justifies the use of the circumscribed area of the See the test and plant strainers for scaling of the sector tests. Response to 3.f.4 and 3.f.10 10 Without information on how debris was prepared and introduced See the into the thin bed tests it cannot be concluded that the thin bed Response to 3.f.4 testing was valid. It is possible that a properly conducted thin bed and 3.f.6. test would result in higher head loss than the full load test that was stated to be the limiting head loss condition for Vogtle. Please provide information that justifies that the sector testing conducted to determine the strainer's ability to deal with a thin bed was conducted under conditions that would conservatively model the debris bed. Please reference the staff Head Loss and Vortexing Guidance for thin bed testing considerations (ADAMS Accession No. ML080230038). 11 The supplemental response stated that air ingestion was evaluated See the at the top of the module. The results of the vortexing evaluation Response to were not provided. Please provide the results of the air 3.f.3. ingestion/vortexting evaluation including the plant conditions assumed. 12 No documentation of fiber size distribution used for testing See the compared to the fiber size distribution predicted to arrive at the Response to strainer was provided. The supplemental response stated that 3.f.4. only fine and small pieces of fiber would be created by the break. The size distribution of the fibrous debris used in testing was not provided. In general shredded fiber does not imply that all fine fibers are created. For thin bed testing, only fine fibrous debris should be added to the test flume until all fibrous fines predicted to be created are added. Please provide information regarding the size distribution of fibrous debris used in various tests and how these size distributions compare to the transport evaluation predictions. E5-180

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) RAI Complete RAls as Provided in NL-08-1829 (Reference 87) Response No. 13 The documentation of fibrous concentration during addition and Tank test was methods of addition to the flume were unclear. Documentation utilized which did should be provided showing that the concentration of debris during not allow settling addition was controlled so that non-prototypical agglomeration of of debris. the debris would not occur. Please provide information that justifies that the debris introduction methods used during testing See the did not result in non-prototypical settling or agglomeration of Response to debris. Also, please include the amounts of debris added during 3.f.4. each addition, the actual size distribution of the debris, and the debris types. 14 Documentation of the amount of debris that settled in the agitated See the and non-agitated areas of the test tanks was not provided. Please Response to 3.f.1 provide the amount of debris that settled in the agitated and non- and 3.f.4. agitated areas of the test tank for each test. 15 There is no discussion of extrapolation of head loss test results to See the ECCS mission times, nor discussion of test termination criteria and Response to subsequent extrapolation. Please provide information that shows 3.f.10. that the head loss testing was run to a maximum value, or that an extrapolation was performed to obtain the head loss at the end of the strainer mission time. Please provide sufficient time dependent test data so that the termination criteria and any extrapolations conducted can be verified. Please provide a graph of the head loss over time for the limiting module and sector tests. Please specify the sector test that created the limiting head loss. 16 The flashing evaluation did not describe the margin to flashing See the through the strainer. The supplemental response stated that Response to overpressure is credited, but the amount of overpressure required 3.f.14. was not provided, nor was the available margin. The total head loss (without chemicals) is about 8 ft with a submergence of about 3 inches (LBLOCA). Please provide the minimum margin to flashing across the strainer throughout the strainer mission time. Please provide the assumptions used to determine this value. ES-181

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) RAI Complete RAls as Provided in NL-08-1829 (Reference 87) Response No. 17 The supplemental response stated that the vortex testing was See the conducted at a submergence that may have been slightly greater Response to (non-conservative) than that expected under LBLOCA conditions 3.f.3. (3.4 in vs. 3.675 +/- 0.5 in). Another section of the supplemental response stated that the testing was conducted with a representative or conservatively lowered water level, but no other details were provided. No vortex testing appears to have been conducted for SBLOCA conditions. No details on the test flow I I rates for vortex evaluations were provided. Vortex testing should be conducted with the minimum potential submergence and the maximum potential scaled flow rate through the strainer. Please provide an evaluation of vortex formation for the minimum level at which the strainer is required to operate (likely an SBLOCA condition). Please verify that the flow rates used in the vortex testing were conservative including the potential for higher flow rates in some sections of the strainer (generally those hydraulically closer to the pump suction). Please verify that the level that was tested for the LBLOCA case was in fact conservatively low. Please provide the submergence value for LBLOCA testing. Alternatively, please provide an updated evaluation considering all of these considerations. 18 The clean strainer head loss (CSHL) calculation methodology was See the not provided. It was not clear how the CSHL was divided between Response to strainer module head loss and piping head loss. Please provide 3.f.9. the methodology used to determine CSHL. Please provide information indicating that each section of the strainer, plenum, or piping was included in the calculation, the head loss value for each section, and the method used to determine the head loss for each strainer section. Please include any assumptions made for each portion of the calculation. 19 The supplemental response stated that for the sector tests debris Sector test is no was maintained in suspension using stirring. No information was longer utilized. provided to show that the stirring did not drive non-prototypical debris onto the bed nor prevent debris from collecting naturally on See the the strainer during these tests. For the module tests, from the Response to provided diagrams it appeared that the stirrer was far enough from 3.f.4. the strainer to prevent non-prototypical bed formation. Please provide information that justifies that the debris beds were not disrupted by the stirring and that the stirring did not result in non-prototypical debris accumulation on the strainer (accumulation of larger sizes of debris than would be expected). ES-182 J

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) RAI Complete RAls as Provided in NL-08-1829 (Reference 87) Response No. 20 During module tests stirring was used outside the area of the See the strainer. The supplemental response stated that the flow in the Response to test flume conservatively represented the plant flow patterns in the 3.f.4. area of the strainer to ensure that non-prototypical settling would not occur. No details were provided on how the plant and flume flow rates were modeled to assure that flow and turbulence would be prototypical. In general flow patterns in the plant are affected by, for example, upstream conditions, drainage into the sump pool, flow rates from various locations, upstream obstructions, and obstructions near the strainer. The boundary conditions in the models for determining typical plant flow patterns should be prototypical or conservative. Please provide information that justifies that the flow rates and patterns in the test flume for the module tests were prototypical or conservative with respect to plant conditions. 21 Please provide the basis for the statement in the supplemental RHR and CS response that the debris head loss for the RHR strainers bounds strainer head the head loss for the containment spray strainers. losses are evaluated independently. See the Response to 3.f.10. 22 Because of the large volume of debris and the relatively low See the submergence of the strainer it is possible for debris to collect on Response to top of the strainer and provide a pathway for air ingestion. This 3.f.3. was not discussed in the supplemental response. Air ingestion could result from a damming effect, or, if head loss exceeds submergence and holes form in the debris bed, these holes could allow air to be ingested through the debris bed. Please provide an evaluation of the potential for debris to collect on top of the strainer and provide a pathway for air ingestion into the strainer. 23 It was unclear how varying the debris loading affected the results See the in all of the head loss testing. Please provide the debris amounts Response to added to each test, the resulting theoretical bed thicknesses, and 3.f.4, 3.f.6, 3.f.7, the maximum head loss determined for each test. and 3.f.10. E5-183 I ___________J

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) RAI Complete RAls as Provided in NL-08-1829 (Reference 87) Response No. 24 Please provide the containment sump/pool level both soon after See the the realignment to containment spray recirculation as well as at the Response to 3.f.2 time post-accident when all assumed water contributions and and 3.g.1. diversions/hold-ups have completely taken effect (except for subsequent sump/pool water thermal contraction). See the The differences in the assumptions and results for these two cases Response to should be clearly explained, as should the times when the short- 3.g.2 and 3.g.9. term and long-term results are applicable. The strainer submergence should be provided for both cases. 25 The second portion of item 3.k of the revised content guide for the See the GL2004-02 supplemental responses requests that the licensee Response to "summarize the structural qualification results and design margins 3.k.2. for the various components of the sump strainer structural assembly." Please provide the actual and allowable stresses and show the design margins for the 16 bolt locations of the strainer base frame (in addition to the reaction forces already provided in Table 3.k.2-8 of the supplemental response). 26 Item 3.k.3 of the revised content guide for the GL2004-02 See the supplemental responses requests that the licensee "summarize Response to the evaluations performed for dynamic effects such as pipe whip, 3.k.3. jet impingement, and missile impacts associated with high-energy line breaks (as applicable)." In addition to the information provided in your September 2005 and February 2008 responses, please submit a detailed summary along with any additional supporting information regarding your assessment that the strainers are not subject to the aforementioned dynamic effects. 27 Please provide additional basis for concluding that the refueling See the cavity drains would not become blocked with debris. Please Response to 3.1.4. identify the potential types and characteristics of debris that could reach these drains. In particular, could large pieces of debris be blown into the upper containment by pipe breaks occurring in the lower containment, and subsequently fall into the cavity? In the case that partial/total blockage of the drains might occur, what would be the impact to minimum sump water level and ECCS and CS pump NPSH? Are there any potential flow restrictions in the two 12-inch refueling cavity drain lines (e.g., valves, meshing or gratings), and if so, how are these potential restrictions addressed so as ensure that these lines are not blocked during a LOCA? ES-184 __J

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) RAI Complete RAls as Provided in NL-08-1829 (Reference 87) Response No. 28 The NRC staff considers in-vessel downstream effects to not be See the fully addressed at Vogtle Units 1 and 2, as well as at other PWRs. Response to The supplemental response refers to draft WCAP-16793-NP, 3.n.1.

         "Evaluation of Long-Term Cooling Considering Particulate, Fibrous, and Chemical Debris in the Recirculating Fluid." TheNRC staff has not issued a final safety evaluation (SE) for WCAP-16793-NP. The licensee may demonstrate that in-vessel downstream effects issues are resolved for Vogtle Units 1 and 2 by showing that the Vogtle plant conditions are bounded by the final WCAP-16793-NP and the corresponding final NRC staff SE, and by addressing the conditions and limitations in the final SE. The licensee may alternatively resolve this item by demonstrating, without reference to WCAP-16793-NP or the staff SE, that in-vessel downstream effects have been addressed at Vogtle Units 1 and 2. In any event, the licensee should report how it has addressed the in-vessel downstream effects issue within 90 days of issuance of the final NRC staff SE on WCAP-16793-NP. The NRC staff is developing a Regulatory Issue Summary to inform the industry of the staffs expectations and plans regarding resolution of this remaining aspect of GSl-191.

29 The NRC Staff understands that SNC has changed its test See the approach to evaluate chemical effects. Please submit the revised Response to chemical effects test results and analyses to the NRC when they 3.o.1. become available. 5.0

References:

1. NRC Generic Letter (GL) 2004-02, "Potential Impact of Debris Blockage on Emergency Recirculation during Design Basis Accidents for Pressurized-Water Reactors," dated September 13, 2004

2. NEI Guidance Report NEI 04-07, Revision 0, "Pressurized Water Reactor Sump Performance Evaluation Methodology 'Volume 1 - Pressurized Water Reactor Sump Performance Evaluation Methodology'," December 2004

3. NEI Guidance Report NEI 04-07-, Revision 0, "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'," December 2004

4. Not Used

5. NRC Bulletin 2003-01, "Requests for Additional Information, Bulletin 2003-01, Potential Impact of Debris Blockage on Emergency Sump Recirculation at Pressurized Water Reactors" for VEGP Electric Generating Plant, Units 1 and 2, Docket Nos. 50-424 and 50-425

6. 60 Day Response to NRC Bulletin 2003-01 SNC-to-NRC NL-03-1514 dated 8/07/2003 ES-185

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Combined SNC response for Joseph M. Farley Nuclear Plant (FNP) and Vogtle Electric Generating Plant (VEGP) as required by NRC Bulletin 2003-01, "Potential Impact of Debris Blockage on Emergency Sump Recirculation at Pressurized Water Reactors" (ML032240030)

7. Response to a Request for Additional Information on NRC Bulletin 2003-01 NRC-to-SNC (NL-04-2013) dated 10/29/2004 Combined SNC response for Joseph M. Farley Nuclear Plant (FNP) and Vogtle Electric Generating Plant (VEGP) as required by NRC Bulletin 2003-01, "Potential Impact of Debris Blockage on Emergency Sump Recirculation at Pressurized Water Reactors"

8. Revised Response to a Request for Additional Information on NRC Bulletin 2003-01 NRC-to-SNC (NL-05-1207) dated 7/22/2005 Combined SNC response for Joseph M. Farley Nuclear Plant (FNP) and Vogtle Electric Generating Plant (VEGP) as required by NRC Bulletin 2003-01, "Potential Impact of Debris Blockage on Emergency Sump Recirculation at Pressurized Water Reactors - Revision 1"

9. NRC-to-SNC (NL-05-1633) dated 8/26/2005 Vogtle Electric Generating Plant, Units 1 and 2 - Response to NRC Bulletin 2003-01, "Potential Impact of Debris Blockage on Emergency Sump Recirculation at Pressurized Water Reactors" (TAC Nos. MB9625 and MB9626)

10. 90 day response to GL 2004-02

• SNC-to-NRC NL-05-0290 dated 2/25/2005 (ML052430746)

Joseph M. Farley Nuclear Plant, Vogtle Electric Generating Plant, Response to NRC Generic Letter 2004-02, "Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors"

11. Response (VEGP and FNP) to GL 2004-02 SNC-to-NRC NL-05-1264 dated 8/31/2005 Combined SNC response for Joseph M. Farley Nuclear Plant (FNP) and Vogtle Electric Generating Plant (VEGP) as required by NRC Generic Letter 2004-02, "Potential Impact of Debris Blockage on Emergency Recirculation during Design Basis Accidents at Pressurized Water Reactors"

12. NRC Request for Additional Information NRC-to-SNC (NL-06-0279) dated 2/9/2006 Vogtle Electric Generating Plant, Units 1 And 2, Request For Additional Information Re: Response To Generic Letter 2004-02, "Potential Impact Of Debris Blockage On Emergency Recirculation During Design-Basis Accidents At Pressurized Water Reactors" (TAC Nos. MC4727 and MC4728)

13. NRC-to-SNC (NL-06-0753) dated 3/28/2006 (ML060870274)

Alternative Approach for Responding to the Nuclear Regulatory Commission Request for Additional Information Letter Re: Generic Letter 2004-02,

14. VEGP 1st extension request to complete CAs (Unit 1 downstream effects) for GL 2004-02 SNC-to-NRC (NL-06-1275) dated 6/22/06 (ML061730462)

Vogtle Electric Generating Plant - Units 1 and 2 Request for Extension for Completing Corrective Actions for Generic Letter 2004-02, "Potential Impact of ES-186

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors"

15. SNC-to-NRC (NL-06-1483) dated 7/28/2006 Response to NRC RAI (6/30/06 phone call )on SNC Request for Extension for Completing Corrective Actions for Generic Letter 2004-02, "Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors"

16. NRC-to-SNC (NL-06-2055) dated 9/7/2006 (ML062500269)

Vogtle Electric Generating Plant, Unit 1, Approval of Generic Letter 2004-02 Extension Request (SNC request dated 6/22/2006)

17. NRC-to-NEI (NL-06-2686) dated 11/14/2006 Nuclear Regulatory Communication Request for Additional Information to Pressurized Water Reactor Licensees Regarding Reponses to Generic Letter 2004-02

18. NRC-to-All Licenses (NL-07-0090) dated 1/4/2007 Alternative Approach for Responding to the NRC request for Additional Information Letter Regarding GL 2004-02

19. SNC-to-NRC (NL-07-1969) dated 12/7/2007 Vogtle Electric Generating Plant Units 1 and 2 Generic Letter 2004-02 Response Extension Request for completion of Chemical Effects testing and analysis, Downstream Effects analysis for Components - Systems, and Fuel -

Vessel

20. NRC-to-SNC (NL-07-2367) dated 12/19/2007 Vogtle Electric Generating Plant, Units 1 and 2 -Generic Letter 2004-02 "Potential Impact of Debris Blockage on Emergency Recirculation during Design Basis Accidents at Pressurized-Water Reactors," Extension Request Approval (to May 31, 2008)

21. WCAP-16406-P-A Revision 1.0, "Evaluation of Downstream Sump Debris Effects in Support of GSl-191" March 2008

22. WCAP-16793-NP-A Revision 2.0, "Evaluation of Long-Term Cooling Considering Particulate, Fibrous and Chemical Debris in the Recirculating Fluid," July 2013

23. WCAP-16568-P Revision 0.0, "Jet Impingement Testing to Determine the Zone of Influence (ZOI) for OBA-Qualified I Acceptable Coatings"

24. Not used

25. Regulatory Guide 1.82, Revision 3, "Water Sources for Long Term Recirculation Cooling Following a Loss of Coolant Accident," November 2003

26. NUREG/CR-0800, Revision 1, "U.S. Nuclear Regulatory Commission Standard Review Plan," Section 3.6.2, "Determination of Rupture Locations and Dynamic Effects Associated with the Postulated Rupture of Piping," July 1981

27. NUREG/CR-2791, "Methodology for Evaluation of Insulation Debris Effects, Containment Emergency Sump Performance Unresolved Safety Issue A-43,"

Issued September 1982

28. NUREG/CR-3616, Transport and Screen Blockage Characteristics of Reflective Metallic Insulation Materials," January 1984

29. NUREG/CR-6224, "Parametric Study of the Potential for BWR ECCS Strainer Blockage Due to LOCA Generated Debris, Final Report," Issued October 1995 ES-187

                                        . Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary)

30. NUREG/CR-6369, "Drywell Debris Transport Study, Final Report," Volume 1, Issued September 1999

31. NUREG/CR-6369, "Drywell Debris Transport Study: Experimental Work, Final Report," Volume 2, Issued September 1999

32. NUREG/CR-6369, "Drywell Debris Transport Study: Computational Work, Final Report," Volume 3, Issued September 1999

33. NUREG/CR-6762, Volume 1, "GSl-191 Technical Assessment: Parametric Evaluations for Pressurized Water Reactor Recirculation Sump Performance,"

Issued August 2002

34. NUREG/CR-6762, Volume 2, "GSl-191 Technical Assessment: Summary and Analysis of U.S. Pressurized Water Reactor Industry Survey Responses and Responses to GL 97-04," Issued August 2002

35. NUREG/CR-6762, Volume 3, "GSl-191 Technical Assessment: Development of Debris Generation Quantities in Support of the Parametric Evaluation," Issued August 2002

36. NUREG/CR-6762, Volume 4, "GSl-191 Technical Assessment: Development of Debris Transport Fractions in Support of the Parametric Evaluation," Issued August 2002

37. NUREG/CR-6772, "GSl-191: Separate Effects Characterization of Debris Transport in Water," Issued August 2002

38. NUREG/CR-6773, "GSl-191: Integrated Debris-Transport Tests in Water Using Simulated Containment Floor Geometries," Issued December 2002

39. NUREG/CR-6808, "Knowledge Base for the Effect of Debris on Pressurized Water Reactor Emergency Core Cooling Sump Performance," Issued February 2003

40. NUREG/CR-6916, "Hydraulic Transport of Coating Debris, A Subtask of GSl-191,"

Issued December 2006

41. NEI Document 02-01, Revision 1, "Condition Assessment Guidelines: Debris Sources Inside PWR Containments"

42. Westinghouse Technical Bulletin, TB-06-15, "Unqualified Service Level 1 Coatings on Equipment in Containment," Dated September 28, 2006

43. C.D.I. Report 96-06, Revision A, "Air Jet Impact Testing of Fibrous and Reflective Metallic Insulation," included in Volume 3 of General Electric Document NED0-32686-A, "Utility Resolution Guide for ECCS Suction Strainer Blockage"

44. STPNOC Letter NOC-AE-13003043 to NRC, "Supplement 1 to Revised STP Pilot Submittal and Requests for Exemptions and Licensing Amendment for a Risk-Informed Approach to Resolving Generic Safety Issue (GSl)-191,"

November 13, 2013 (ML13323A183)

45. STPNOC Letter NOC-AE-15003241 to NRC, "Supplement 2 to STP Pilot Submittal and Requests for Exemptions and License Amendment for a Risk-Informed Approach to Address Generic Safety Issue (GSl)-191 and Respond to Generic Letter (GL) 2004-02 (TAC NOS. MF2400-MF2409)," August 20, 2015 (ML15246A126)

46. NEI Document (ML120481057), Revision 1, "ZOI Fibrous Debris Preparation:

Processing, Storage and Handling," January 2012

47. NRC Bulletin 2003-01, "Potential Impact of Debris Blockage on Emergency Sump Recirculation at Pressurized Water Reactors," dated June 9, 2003 ES-188

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary)

48. NRCB 93-02, NRC Bulletin 93-02, "Debris Plugging of Emergency Core Cooling Suction Strainers," May 11, 1993

49. NRCB 96-03, NRC Bulletin 96-03, "Potential Plugging of Emergency Core Cooling Suction Strainers by Debris in Boiling-Water Reactors," May 6, 1996

50. NUREG/CR-1829 Volume I, "Estimating LOCA Frequencies Through the Elicitation Process," 2008

51. NUREG/CR-2982, Revision 1, "Buoyancy, Transport, and Head Loss of Fibrous Reactor Insulation," July 1983

52. NUREG/CR-5640, "Overview and Comparison of US Commercial Nuclear Power Plants," September 1990

53. NUREG/CR-6367, "Experimental Study of Head Loss and Filtration for LOCA Debris," February 1996

54. NUREG/CR-6808, "Knowledge Base for the Effect of Debris on Pressurized Water Reactor Emergency Core Cooling Sump Performance," November 2005

55. NUREG/CR-6874, GSl-191: Experimental Studies of Loss-of-Coolant-Accident-Generated Debris Accumulation and Head Loss with Emphasis on the Effects of Calcium Silicate Insulation, May 2005

56. NUREG/CR-6877, Characterization and Head-Loss Testing of Latent Debris from Pressurized-Water-Reactor Containment Buildings, July 2005

57. NUREG/CR-6917, Experimental Measurements of Pressure Drop across Sump Screen Debris Beds in Support of Generic Safety Issue 191, February 2007

58. NUREG/CR-6988, Final Report- Evaluation of Chemical Effects Phenomena in Post-LOCA Coolant, March 2009: Revision 0

59. NUREG/CR-7172, "Knowledge Base Report on Emergency Core Cooling Sump Performance in Operating Light Water Reactors," January 2014

60. NUREG/CR-0869, USI A-43 Regulatory Analysis, Revision 1: October 1985

61. NUREG/CR-0897, Technical Findings Related to Unresolved Safety Issue A-43, Revision 1: October 1985

62. Not used

63. NUREG/CR-1862, Development of a Pressure Drop Calculation Method for Debris-Covered Sump Screens in Support of Generic Safety Issue 191, February 2007

64. NUREG/CR-1918, "Phenomena Identification and Ranking Table Evaluation of Chemical Effects Associated with Generic Safety Issue 191," February 2009

65. PWROG, OG-07-419, Transmittal of LOCADM Software in Support of WCAP-16793-NP, "Evaluation of Long-Term Cooling Associated with Sump Debris Effects" (PA-SEE-0312), September 2007

66. PWROG, OG-07-534, Transmittal of Additional Guidance for Modeling Post-LOCA Core Deposition with LOCADM Document forWCAP-16793-NP (PA-SEE-0312),

December 2007

67. PWROG, OG-08-64, Transmittal of LTR-SEE-1-08-30, "Additional Guidance for LOCADM for Modification to Aluminum Release" for Westinghouse Topical Report WCAP-16793-NP, "Evaluation of Long-Term Cooling Considering Particulate, Fibrous and Chemical Debris in the Recirculating Fluid" (PA-SEE-0312),

February 2008 ES-189

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary)

68. PWROG, OG-10-253, PWROG Response to Request for Additional Information Regarding PWROG Topical Report WCAP-16793-NP, Revision 1, "Evaluation of Long-Term Cooling Considering Particulate, Fibrous and Chemical Debris in the Recirculating Fluid" (PA-SEE-0312), August 2010

69. Regulatory Guide 1.174, Revision 2, "An Approach for Using Probabilistic Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," May 2011

70. SRM-SECY-12-0093, "Closure Options for Generic Safety Issue - 191, Assessment of Debris Accumulation on PWR Sump Performance,"

December 14, 2012

71. SECY-83-472, Information Report from W.J. Dircks to the Commissioners, "Emergency Core Cooling System Analysis Methods," November 17, 1983

72. WOG-06-113, "Submittal ofWCAP-16530-NP, 'Evaluation of Post-Accident Chemical Effects in Containment Sump Fluids to Support GSl-191' for Formal Review," 3/27/2006

73. WCAP-16530-NP-A, "Evaluation of Post-Accident Chemical Effects in Containment Sump Fluids to Support GSl-191," March 2008

74. WCAP-16613-P, Vogtle Electric Generating Plant Measurement Uncertainty Recapture Power Uprate Program Engineering Report, Revision 2, June 2007

75. WCAP-16785-NP, Revision 0, "Evaluation of Additional Inputs to the WCAP-16530-NP Chemical Model," May 2007

76. Not Used

77. Not Used

78. BWR Owners Group, "Utility Resolution Guide for ECCS Suction Strainer Blockage," Volume 3, October 1998

79. PA-SEE-1090(ML14153A013), PWROG Presentation, "GSl-191 Comprehensive Analysis and Test Program Update," NRC Public Meeting: April 2014

80. Letter from William Ruland (NRC) to Anthony Pietrangelo (NEI) (ML080230112),

Revised Guidance for Review of Final Licensee Responses to Generic Letter 2004-02, "Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors," March 28, 2008

81. ML081550043, C. B. Bahn et al., "Technical Letter Report on Evaluation of Long-term Aluminum Solubility in Borated Water Following a LOCA," February 2008

82. NRC Letter to SNC (ML092370630), "Summary of August 13, 2009, Public Conference Call with Southern Nuclear Operating Company, Inc. (SNC), on the Request for Additional Information Pertaining to Generic Letter 2004-02 (TAC NOS. MC4727 and MC2728)," August 31, 2009

83. ML102280594, "Evaluation of Chemical Effects Phenomena Identification and Ranking Table Results," March 2011

84. ML121520429, Nuclear Regulatory Commission, Official Transcript of.

Proceedings, Advisory Committee on Reactor Safeguards Thermal Hydraulic Phenomena Subcommittee Open Session, May 9, 2012

85. SNC Letter NL-04-2321 to NRC, "Joseph M. Farley Nuclear Plant Response to a Request for Additional Information on NRC Bulletin 2003-01 'Potential Impact of Debris Blockage on Emergency Sump Recirculation at Pressurized Water Reactors'," November 30, 2004 ES-190

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary)

86. SNC Letter NL-07-2168 to NRC (ML080150161), "Vogtle Electric Generating Plant License Amendment Request to Revise Technical Specifications (TS) 3.3.2,

        'ESFAS Instrumentation,' and TS 3.5.4, 'Refueling Water Storage Tank (RWST)',"

January 2008

87. NRC Letter NL-08-1829 to SNC, "Vogtle Electric Generating Plant, Units 1 and 2 -

Generic Letter 2004-02, 'Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized Water Reactors,' Request for Additional Information (TAC NOS. MC4727, MC4728)," December 2, 2008 (ML083100142) I ' 88. Pressurized Water Reactor Owners Group (PWROG) Letter OG-07-408, Revision 0, "PWROG Responses to NRC Second Set of Requests for Clarification and Supplemental Information Regarding WCAP-16530," September 2007

89. WOG-06-107, "PWR Owners Group Letter to NRC Regarding Error Corrections to WCAP-16530-NP (PA-SEE-0275)," March 21, 2006

90. SECY-10-0113, "Closure Options for Generic Safety Issue - 191, Assessment of .

Debris Accumulation on PWR Sump Performance," December 23, 2010

91. NEI letter to NRC, "GSl-191 - Current Status and Recommended Actions for Closure," May 4, 2012(ML12142A316)

92. NEI letter to NRC, "GSl-191 - Revised Schedule for Licensee Submittal of Resolution Path," November 15, 2012 (ML12325A072)

93. "NRC Review of Generic Safety lssue-191 Nuclear Energy Institute revised Schedule for Licensee Submittal of Resolution Path," (ML12326A497),

November 21, 2012

94. "Final Safety Evaluation for Pressurized Water Reactor Owners Group Topical Report WCAP-16793-NP, Revision 2, 'Evaluation of Long-Term Cooling Considering Particulate Fibrous and Chemical Debris in the Recirculating Fluid' (TAC NO. ME1234)," April 8, 2013(ML13084A152 and ML13084A154)

95. SNC Letter NL-07-1777 to NRC (ML080640601), "Vogtle Electric Generating Plant Supplemental Response to NRC Generic Letter 2004-02," February 28, 2008

96. SNC Letter NL-08-0670 to NRC (ML081640617), "Vogtle Electric Generating Plant Supplemental Response to NRC Generic Letter 2004-02," May 21, 2008

97. SNC Letter NL-08-0818 to NRC (ML081430616), "Vogtle Electric Generating Plant Generic Letter 2004-02 Response Extension Request," May 22, 2008

98. SNC Letter NL-08-1155 to NRC (ML082170513), "Vogtle Electric Generating Plant Supplemental Response to NRC Generic Letter 2004-02," July 31, 2008

99. SNC Letter NL-08-1195 to NRC (ML082170306), "Vogtle Electric Generating Plant Generic Letter 2004-02 Response Extension Request," July 31, 2008 100. SNC Letter NL-08-1228 to NRC (ML082380890), "Vogtle Electric Generating Plant Supplemental Response to NRC Generic Letter 2004-02," August 22, 2008 101. NRC Letter NL-08-1497 to SNC (ML082560233), "Vogtle Electric Generating Plant, Units 1 and 2 - Request for Additional Information Regarding Generic Letter 2004-02, 'Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors'," September 17, 2008 102. SNC Letter NL-08-1583 to NRC (ML083150262), "Vogtle Electric Generating Plant Generic Letter 2004-02 Extension Request for 'Completion of Chemical Effects and Closeout of GL 2004-02'," November 7, 2008 ES-191

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) 103. SNC Letter NL-09-0159 to NRC (ML090420235), "Vogtle Electric Generating Plant i I Generic Letter Supplemental Response," February 10, 2009 104. SNC Letter NL-09-1839 to NRC (ML093240098), "Vogtle Electric Generating Plant Generic Letter 2004-02 Closeout Status," November 19, 2009 I' 105. SNC Letter NL-13-0953 to NRC (ML13137A130), "Vogtle Electric Generating Plant Generic Letter 2004-02 Closeout Status," May 16, 2013 106. NRC Evaluation Guide (ML080380214), "NRC Staff Review Guidance Regarding Generic Letter 2004-02 Closure in the Area of Plant-Specific Chemical Effect Evaluations," March 2008

• 107. Regulatory Guide 1.82, Revision 4, "Water Sources for Long Term Recirculation Cooling Following a Loss of Coolant Accident," March 2012 108. NRC Letter (ML073110278) "Revised Content Guide for Generic Letter 2004-02 Supplemental Responses," November 2007 109. STPNOC letter NOC-AE-16003401 to the NRC, "Supplement 3 to Revised STP Pilot Submittal and Requests for Exemptions and License Amendment for a Risk-Informed Approach to Address Generic Safety Issue (GSl)-191 and Respond to Generic Letter (GL) 2004-02," October 20, 2016 110. SNC Letter NL-13-2544(ML13351A409) "Vogtle Electric Generating Plant, Units 1 and 2 Response to Request for Additional Information Regarding Closure of Option 2 to Address In-Vessel Mitigative Measures for Potential In-Vessel Blockage," December 13, 2013 111. NRC Evaluation Guide (ML080230038), "NRC Staff Review Guidance Regarding Generic Letter 2004-02 Closure in the Area of Strainer Head Loss and Vortexing,"

March 2008 ES-192

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Vogtle Electric Generating Plant Supplemental Response to NRC Generic Letter 2004-02 Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Attachment 1 General Electric Hitachi (GEH) Proprietary Information with Affidavit E5:A1-1

Global Nuclear Fuel - Americas AFFIDAVIT I, Peter M. Yandow, state as follows: (1) I am the Vice President, NPP/Services Licensing, Regulatory Affairs, GE-Hitachi Nuclear Energy Americas, LLC (GEH), and have been delegated the function of reviewing the information described in paragraph (2) which is sought to be withheld, and have been authorized to apply for its withholding. (2) The information sought to be withheld is contained in Enclosure 1 of GEH's letter, 0006-7789-010, Jack Noonan (GEH) to Jim A. Wade (Southern Nuclear Company), entitled "GEH Proprietary Information in SNC Supplemental Response to NRC Generic Letter 2004-02," March 29, 2017. GEH proprietary information in Enclosure 1, which is entitled "Excerpt of SNC Supplemental Response to NRC Generic Letter 2004 GEH Proprietary Information - Class II (Internal), is identified by a dotted underline inside double square brackets. ((J.h.t~--~-~nt~n£~.j~Jm ..~~J!!DP.l~Y-~J] In each case, the superscript notation {3} refers to Paragraph (3) of this affidavit, which provides the basis for the proprietary determination. (3) In making this application for withholding of proprietary information of which it is the owner or licensee, GEH relies upon the exemption from disclosure set forth in the Freedom of Information Act ("FOIA"), 5 USC Sec. 552(b)(4), and the Trade Secrets Act, 18 USC Sec. 1905, and NRC regulations 10 CFR 9.17(a)(4), and 2.390(a)(4) for "trade secrets" (Exemption 4). The material for which exemption from disclosure is here sought also qualify under the narrower definition of "trade secret", within the meanings assigned to those terms for purposes of FOIA Exemption 4 in, respectively, Critical Mass Energy Project v. Nuclear Regulatory Commission, 975F2d871 (DC Cir. 1992), and Public Citizen Health Research Group v. FDA, 704F2d1280 (DC Cir. 1983). (4) Some examples of categories of information which fit into the definition of proprietary information are:

a. Information that discloses a process, method, or apparatus, including supporting data and analyses, where prevention of its use by GEH's competitors without license from GEH constitutes a competitive economic advantage over other companies;

b. Information which, if used by a competitor, would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing of a similar product;

c. Information which reveals aspects of past, present, or future GEH customer-funded development plans and programs, resulting in potential products to GEH;

d. Information which discloses patentable subject matter for which it may be desirable to obtain patent protection.

0006-7789-010 Affidavit Page 1 of 3 l

The information sought to be withheld is considered to be proprietary for the reasons set forth in paragraphs (4 )a. and (4)b. above. (5) To address 10 CPR 2.390 (b) (4), the information sought to be withheld is being submitted to NRC in confidence. The information is of a sort customarily held in confidence by GEH, and is in fact so held. The information sought to be withheld has, to the best of my knowledge and belief, consistently been held in confidence by GEH, no public disclosure has been made, and it is not available in public sources. All disclosures to third parties including any required transmittals to NRC, have been made, or must be made, pursuant to regulatory provisions or proprietary agreements which provide for maintenance of the information in confidence. Its initial designation as proprietary information, and the subsequent steps taken to prevent its unauthorized disclosure, are as set forth in paragraphs (6) and (7) following. (6) Initial approval of proprietary treatment of a document is made by the manager of the originating component, the person most likely to be acquainted with the value and sensitivity of the information in relation to industry knowledge, or subject to the terms under which it was licensed to GEH. (7) The procedure for approval of external release of such a document typically requires review by the staff manager, project manager, principal scientist or other equivalent authority, by the manager of the cognizant marketing function (or his delegate), and by the Legal Operation, for technical content, competitive effect, and determination of the accuracy of the proprietary designation. Disclosures outside GEH are limited to regulatory bodies, customers, and potential customers, and their agents, suppliers, and licensees, and others with a legitimate need for the information, and then only in accordance with appropriate regulatory provisions or proprietary agreements. (8) The information identified in paragraph (2) is classified as proprietary because it contains detailed results of analytical model and methods of emergency core cooling system and containment spray strainers in Boiling Water Reactors and Pressurized Water Reactors. The development and approval of these models and methods were achieved at a significant cost to GEH. The development of the evaluation process along with the interpretation and application of the analytical results is derived from the extensive experience database that constitutes a major GEH asset. (9) Public disclosure of the information sought to be withheld is likely to cause substantial harm to GEH's competitive position and foreclose or reduce the availability of profit-making opportunities. The information is part of GEH's comprehensive safety and technology base, and its commercial value extends beyond the original development cost. The value of the technology base goes beyond the extensive physical database and analytical methodology and includes development of the expertise to determine and apply 0006-7789-010 Affidavit Page 2 of 3

the appropriate evaluation process. In addition, the technology base includes the value derived from providing analyses done with NRC-approved methods. The research, development, engineering, analytical, and NRC review costs comprise a substantial investment of time and money by GEH. The precise value of the expertise to devise an evaluation process and apply the correct analytical methodology is difficult to quantify, but it clearly is substantial. GEH's competitive advantage will be lost if its competitors are able to use the results of the GEH experience to normalize or verify their own process or if they are able to claim an equivalent understanding by demonstrating that they can arrive at the same or similar conclusions. The value of this information to GEH would be lost if the information were disclosed to the public. Making such information available to competitors without their having been required to undertake a similar expenditure of resources would unfairly provide competitors with a windfall, and deprive GEH of the opportunity to exercise its competitive advantage to seek an adequate return on its large investment in developing and obtaining these very valuable analytical tools. I declare under penalty of perjury that the foregoing is true and correct. Executed on this 29th day of March 2017. Peter M. Yandow Vice President, NPP/Services Licensing GE-Hitachi Nuclear Energy Americas, LLC 3901 Castle Hayne Road, MIC A-65 Wilmington, NC 28401 Peter.Yandow@ge.com 0006-7789-010 Affidavit Page 3 of 3

Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Vogtle Electric Generating Plant Supplemental Response to NRC Generic Letter 2004-02 Enclosure 5 Supplemental Response to NRC Generic Letter 2004-02 (Non-Proprietary) Attachment 2 Westinghouse Electric Corporation (WEC) Proprietary Information with Affidavit E5:A2-1}}