L-2022-121, Exemption Request, License Amendment Request and Revised Response in Support of a Risk-informed Resolution of Generic Letter 2004-02

Exemption Request, License Amendment Request and Revised Response in Support of a Risk-informed Resolution of Generic Letter 2004-02

ML22210A086
Person / Time
Site:	Point Beach
Issue date:	07/29/2022
From:	Strand D Point Beach
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
L-2022-121
Download: ML22210A086 (666)

Text

L-2022-121 10 CFR 50.12 10 CFR 50.90 July 29, 2022 GL 2004-02 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington D C 20555-0001 RE: Point Beach Nuclear Plant, Units 1 and 2 Docket Nos. 50-266 and 50-301 Renewed Facility Operating Licenses DPR-24 and DPR-27 Exemption Request, License Amendment Request and Revised Response in Support of a Risk-informed Resolution of Generic Letter 2004-02

References:

1. NRC Generic Letter 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors, September 13, 2004

2. NextEra Energy Point Beach, LLC letter NRC 2017-0045, Updated Final Response to NRC Generic Letter 2004-02, December 29, 2017 (ADAMS Accession No. ML17363A253)

3. Revised Figure 3.a.1-1 for Turkey Point and Point Beach Updated Final Response to GL 2004-02 (ADAMS Accession No. ML18295A198)

4. NRC Letter Point Beach Nuclear Plant, Units 1 and 2; Seabrook Station, Unit No. 1; St. Lucie Plant, Units 1 and 2; and Turkey Point Nuclear Generating Units 3 and 4, Audit Report Regarding Generic Letter 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors Closure Methodology (EPID 2017-Lrc-0000) dated December 2, 2019 (ADAMS Accession No. ML19217A003)

Pursuant to 10 CFR 50.12, NextEra Energy Point Beach, LLC (NextEra) hereby requests an exemption, from the requirements of 10 CFR 50.46(a)(1), for Point Beach Nuclear Plant Units 1 and 2 (Point Beach),

respectively. The requested exemption would allow the use of risk-informed methods to evaluate the long-term core cooling (LTCC) effects of debris generation resulting from a postulated loss of cooling accident (LOCA) in order to address the safety issues described in Generic Letter (GL) 2004-02 (Reference 1). GL 2004-02 requested licensees evaluate these issues and demonstrate compliance with the emergency core cooling system (ECCS) performance criteria of 10 CFR 50.46 using NRC-approved methodologies.

Additionally, pursuant to 10 CFR 50.90, NextEra hereby requests license amendments to Renewed Facility Operating Licenses DPR-24 and DPR-27 for Point Beach Nuclear Plant Units 1 and 2, respectively. The proposed amendments would revise the licensing basis described in the Point Beach Updated Final Safety Analysis Report (UFSAR) to include a risk-informed method of evaluating the effects of LOCA generated debris on LTCC.

Additionally, NextEra hereby provides on behalf of Point Beach its revised response to GL 2004-02 based on a risk-informed approach to the safety issues described therein. By Reference 2, as supplemented by Reference 3, NextEra submitted a response to GL 2004-02 which rescinded the Point Beach Generic Safety lssue (GSI) 191 and GL 2004-02 related commitments described in previous correspondence and provided a technical basis for the resolution of GSl-191 and thereby closure of GL 2004-02. In Reference 4, following a January 2019 audit of NextEras GSI-191/GL 2004-02 response at NextEras Juno Beach facility, the NRC requested additional information deemed necessary to reach conclusions regarding the functionality NextEra Energy Point Beach, LLC 6610 Nuclear Road, Two Rivers, WI 54241

Point Beach Nuclear Plant, Units 1 and 2 L-2022-121 Docket Nos. 50-266 and 50-301 Page 2 of 2 of the Point Beach emergency core cooling (ECCS) and containment spray (CSS) systems in the context of GSI-191 and GL 2004-02. In public meetings held May 18, 2021 (ADAMS Accession No. ML21134A023) and December 9, 2021 (ADAMS Accession No. ML21349A207), NextEra notified the NRC of its intent to pursue a risk-informed resolution to GL 2004-02 for Point Beach and provided preliminary results of the risk-informed analysis, respectively. The finalized results of this effort are provided herein, along with the response to the NRCs questions raised in Reference 4, in support of GL 2004-02 closure for Point Beach.

The enclosures to this letter provide NextEras evaluation of the proposed changes. Enclosure 1 provides the request for an exemption from 10 CFR 50.46(a)(1). Enclosure 2 provides the amendment request for the Point Beach licensing basis as summarized in the UFSAR. Attachment 1 to Enclosure 2 provides the marked up UFSAR pages showing the proposed changes. The UFSAR markup pages are provided for information only and will be incorporated in accordance with 10 CFR 50.71(e) upon implementation of the approved license amendments. Enclosure 3 provides the revised, risk-informed response to GL 2004-02 for Point Beach. Enclosure 3 supersedes and replaces NextEras previous responses to GL 2004-02 for Point Beach. Changes to NextEras responses provided in Reference 2 are evidenced by revision bars in the right-hand margin of Enclosure 3. Enclosure 4 provides a summary of the risk-informed evaluations performed for Point Beach. Enclosure 5 provides the defense-in-depth and safety margin evaluation for the revised, risk-informed response to GL 2004-02. There are no proposed changes to the Point Beach Technical Specification (TS) or TS Bases associated with this submittal.

NextEra requests the proposed amendments be processed as a normal license amendment request, and requests approval of the proposed exemption and license amendments within one year of satisfactory acceptance reviews. Once approved, the license amendments will be implemented within 90 days.

NextEra has determined that the proposed exemption request and license amendment request meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9) and would not result in a significant radiological environmental impact. NextEra has additionally determined that the requested license amendments do not involve a significant hazards consideration pursuant to 10 CFR 50.92(c) and that there are no significant environmental impacts associated with the proposed changes.

The Point Beach Onsite Review Group (ORG) has reviewed the enclosed exemption request and license amendment request. In accordance with 10 CFR 50.91(b)(1), a copy of this submittal is being forwarded to the designee for the State of Wisconsin.

This letter contains no new regulatory commitments.

Should you have any questions regarding this submittal, please contact Mr. Michael Davis, Fleet Licensing Manager, at 319-851-7032.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on the ____ 29 day of July 2022.

Sincerely, Dianne Strand General Manager, Regulatory Affairs cc: USNRC Regional Administrator, Region III Project Manager, USNRC, Point Beach Nuclear Plant Resident Inspector, USNRC, Point Beach Nuclear Plant Public Service Commission of Wisconsin Enclosures (5)

Point Beach Nuclear Plant Licensing Submittal for a Risk-Informed Resolution of Generic Letter 2004-02 Enclosure 1 Request for Exemption Table of Contents 1.0 GENERAL ............................................................................................................ 2 1.1 Introduction ................................................................................................ 2 1.2 Background and Overview ......................................................................... 3 2.0 EXEMPTION REQUEST ...................................................................................... 4 3.0 REGULATORY REQUIREMENTS INVOLVED .................................................... 6 4.0 BASIS FOR THE EXEMPTION REQUEST .......................................................... 7 4.1 Applicability of 10 CFR 50.12(a)(1) ............................................................ 8 4.2 Applicability of 10 CFR 50.12(a)(2) ............................................................ 9 4.3 Environmental Consideration ................................................................... 11 5.0 TECHNICAL JUSTIFICATION FOR THE EXEMPTION ..................................... 15

6.0 CONCLUSION

.................................................................................................... 15

7.0 REFERENCES

................................................................................................... 15

Enclosure 1 Request for Exemption 1.0 GENERAL 1.1 Introduction This enclosure provides NextEra Energy Point Beach, LLCs (NextEra) request for exemption under Title 10 of the Code of Federal Regulations (CFR) Section 50.12 (10 CFR 50.12) from certain requirements in 10 CFR 50.46 to support Point Beach Nuclear Plant (PBN) risk-informed approach to respond to the United States Nuclear Regulatory Commission (NRC) Generic Letter (GL) 2004-02 (Reference 1). This exemption request complements a license amendment request (LAR) provided in of this submittal for adopting a risk-informed methodology for addressing GL 2004-02. Changes to the PBN Final Safety Analysis Report (FSAR) are marked up in Attachment 1 to Enclosure 2 in this submittal. Enclosure 3 contains the updated GL 2004-02 response. Enclosure 4 provides an overview of the risk-informed methodology and discusses the risk quantification, and Attachment 1 to Enclosure 4 discusses maintaining resolution of GL 2004-02 for future operation at PBN. Enclosure 5 discusses defense-in-depth and safety margins.

The specific exemption request pertains to requirements associated with the emergency core cooling system (ECCS) function for core cooling following a postulated loss of cooling accident (LOCA). The scope and key elements of the requested exemption are described in Section 2.0.

Approval of the exemption will allow the use of a risk-informed method to account for the probabilities and uncertainties associated with mitigation of the effects of debris following postulated LOCAs. The method evaluates concerns raised by GSI-191 related to the effects of post-accident debris on the containment sump recirculation strainers and reactor core blockage due to debris in the recirculating fluid. To confirm acceptable sump design, the risk associated with loss of core cooling due to the effects of debris is evaluated. The risk-informed approach is consistent with the guidance in Regulatory Guide (RG) 1.174 (Reference 2).

The PBN approach is the risk-informed part of an overall graded approach that is based on the amount of debris in the plant, as discussed in SECY-12-0093 (Reference 3). The PBN risk-informed approach addresses the five key principles in RG 1.174 (Reference 2). The resulting risk metrics, in terms of core damage frequency (CDF), large early release frequency (LERF), and changes in CDF (CDF) and LERF (LERF) due to debris effects, are used to determine whether plant modifications are warranted to ensure acceptable sump performance. The PBN risk quantification in Enclosure 4 of this submittal shows that CDF and LERF are below the threshold for RG 1.174 Region Ill, Very Small Changes, without further plant or procedure modifications. Therefore, the risk-informed approach provides an equivalent level of assurance for sump performance without incurring significant cost and occupational dose associated with removing, replacing, or reinforcing insulation in containment. Approval of the requested exemption will support the application of the risk-informed approach.

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Enclosure 1 Request for Exemption 1.2 Background and Overview GSI-191 identifies the possibility that debris generated during a LOCA could clog the containment sump strainers in pressurized water reactors (PWRs) and result in loss of net positive suction head (NPSH) for the ECCS and containment spray system (CSS) pumps, impeding the flow of water from the sump. GL 2004-02 requested licensees to address GSI-191 by demonstrating compliance with the 10 CFR 50.46 ECCS acceptance criteria (Reference 1). In addition, GL 2004-02 required licensees to address downstream effects. As stated in GL 2004-02, licensees were requested to perform analyses using an NRC-approved methodology and to ensure successful operation of the ECCS and CSS during design-basis accidents (DBAs) that require containment sump recirculation (Reference 1):

Although not traditionally considered as a component of the 10 CFR 50.46 ECCS evaluation model, the calculation of sump performance is necessary to determine if the sump and the ECCS are predicted to provide enough flow to ensure long-term cooling.

Based on the new information identified during the efforts to resolve GSI-191, the staff has determined that the previous guidance used to develop current licensing basis analyses does not adequately and completely model sump screen debris blockage and related effects. As a result, due to the deficiencies in the previous guidance, an analytical error could be introduced which results in ECCS and CSS performance that does not conform to the existing applicable regulatory requirements outlined in this generic letter. Therefore, the staff is revising the guidance for determining the susceptibility of PWR recirculation sump screens to the adverse effects of debris blockage during design basis accidents requiring recirculation operation of the ECCS or CSS. In light of this revised staff guidance, it is appropriate to request that addressees perform new, more realistic analyses and submit information to confirm the functionality of the ECCS and CSS during design basis accidents requiring recirculation operations.

In addition, GL 2004-02 identified the following regulatory requirement (Reference 1):

NRC regulations in Title 10, of the Code of Federal Regulations Section 50.46, 10 CFR 50.46, require that the ECCS has the capability to provide long-term cooling of the reactor core following a LOCA. That is, the ECCS must be able to remove decay heat, so that the core temperature is maintained at an acceptably low value for the extended period of time required by the long-lived radioactivity remaining in the core.

Compensatory and mitigative measures have been implemented in response to Bulletin 2003-01 (Reference 4) and GL 2004-02 (Reference 1) to address the potential for sump strainer clogging and related GSI-191 concerns. This included installation of larger containment sump strainers that greatly reduce the potential for loss of NPSH for the RHR and CSS pumps.

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Enclosure 1 Request for Exemption The Commission issued Staff Requirements Memorandum (SRM) SECY-10-0113 directing the staff to consider alternative options for resolving GL 2004-02 (Reference 5). Subsequently, SECY-12-0093 outlined a few different options that PWR licensees can use to address GL 2004-02, including deterministic and risk-informed approaches (References 3 and 6). In a letter to the NRC on May 16, 2013 (Reference 7), PBN originally selected Option 3 in SECY-12-0093, using a deterministic approach for all aspects of GSI-191, except for in-vessel downstream effects, where a risk-informed approach will be used. Following NextEras GL 2004-02 response for PBN in 2017 (Reference 11), PBN subsequently selected to use Option 2, the full risk-informed resolution path, as communicated to the NRC in a public meeting on May 18, 2021.

Based on the guidance in RG 1.174, the risk-informed approach requires an exemption from certain requirements of 10 CFR 50.46 in accordance with 10 CFR 50.12.

2.0 EXEMPTION REQUEST Pursuant to 10 CFR 50.12, NextEra is submitting this request for exemption from certain requirements of 10 CFR 50.46(a)(1), other properties, as it relates to using specific deterministic methodology to evaluate the effects of debris on long-term core cooling.

10 CFR 50.46(a)(1) is shown below with the other properties portion for which exemption is requested in bold.

(a)(1)(i) Each boiling or pressurized light-water nuclear power reactor fueled with uranium oxide pellets within cylindrical zircaloy or ZIRLO cladding must be provided with an emergency core cooling system (ECCS) that must be designed so that its calculated cooling performance following postulated loss-of-coolant accidents conforms to the criteria set forth in paragraph (b) of this section. ECCS cooling performance must be calculated in accordance with an acceptable evaluation model and must be calculated for a number of postulated loss-of-coolant accidents of different sizes, locations, and other properties sufficient to provide assurance that the most severe postulated loss-of-coolant accidents are calculated.

Except as provided in paragraph (a)(1)(ii) of this section, the evaluation model must include sufficient supporting justification to show that the analytical technique realistically describes the behavior of the reactor system during a loss-of-coolant accident. Comparisons to applicable experimental data must be made and uncertainties in the analysis method and inputs must be identified and assessed so that the uncertainty in the calculated results can be estimated. This uncertainty must be accounted for, so that, when the calculated ECCS cooling performance is compared to the criteria set forth in paragraph (b) of this section, there is a high level of probability that the criteria would not be exceeded. Appendix K, Part II Required Documentation, sets forth the documentation requirements for each evaluation model. This section does not apply to a nuclear power reactor facility for which the certifications required under § 50.82(a)(1) have been submitted.

(ii) Alternatively, an ECCS evaluation model may be developed in conformance with the required and acceptable features of appendix K ECCS Evaluation Models.

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Enclosure 1 Request for Exemption The scope of the exemption applies to all debris effects addressed in the risk-informed element of the PBN methodology described in Enclosure 4. The debris effects are associated with those breaks that potentially generate and transport debris that exceeds the tested/analyzed debris limits. NextEra is requesting exemption for these breaks to allow evaluation of the debris effects using a risk-informed methodology. The key elements of the exemption request are listed as follows.

1. The exemption will apply only to the effects of debris as described in Enclosure 4.

2. The exemption will apply to any breaks that can generate and transport debris quantities that are not bounded by PBN-specific analyzed limits, provided that the CDF and LERF remain in RG 1.174 Region III (Reference 2).

This exemption request is complemented by the accompanying LAR in Enclosure 2, which seeks NRC approval to amend the PBN licensing basis describing acceptable design of the containment sump. The risk-informed method provides a high probability of assurance for acceptable sump performance and debris mitigation as assumed in the ECCS evaluation model.

The PBN risk-informed approach to respond to GL 2004-02 is consistent with the NRC safety evaluation (SE) for NEI 04-07 that discussed the modeling of sump performance as follows (Reference 9):

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

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

This exemption request is consistent with the provisions of the proposed ECCS rule change. The following statement, found on Page 85 of the proposed 10 CFR 50.46c final rule change package attached to SECY-16-0033 (Reference 10), applies to the proposed 10 CFR 50.46c(d), which will replace the current 10 CFR 50.46(a)(1):

Demonstration of consideration of such factors may also be achieved through analytical models that adequately represent the empirical data obtained regarding debris deposition. The final rule alternatively allows the use of risk-informed approaches to evaluate the effects of debris on localized coolant flow and delivery of coolant to the core during the long-term cooling (post-accident recovery) period.

The proposed ECCS rule change will allow use of a risk-informed approach, addressed in 10 CFR 50.46c(e), in lieu of a deterministic evaluation. Similar to the proposed new E1-5

Enclosure 1 Request for Exemption rule change, PBN's risk-informed approach is an alternative to the current deterministic evaluation required by 10 CFR 50.46(a)(1). PBN requires exemption from 10 CFR 50.46(a)(1) other properties since there currently is no risk-informed evaluation alternative. PBN requests an exemption from those deterministic requirements in order to enable the use of a risk-informed method to demonstrate acceptable sump performance and debris mitigation, and to validate assumptions in the ECCS evaluation model.

3.0 REGULATORY REQUIREMENTS INVOLVED By regulatory precedent, licensees are required to demonstrate compliance with the relevant regulations by the use of a bounding calculation or other deterministic method.

NextEra seeks exemption to the extent that 10 CFR 50.46(a)(1) other properties requires deterministic calculations or other analyses to address the concerns raised by GSI-191 related to acceptable plant performance during the recirculation mode following a LOCA. The proposed changes to the licensing basis, submitted for NRC approval through a LAR (see Enclosure 2), address GL 2004-02 for PBN on the basis that the associated risk is shown to meet the acceptance guidelines in RG 1.174 and adequate defense-in-depth measures and safety margins are demonstrated.

This exemption request is to allow the use of a risk-informed method to demonstrate acceptable mitigation of the effects of debris following postulated LOCAs. Prior to the risk-informed approach, deterministic methods were used to evaluate the effects of accident-generated and transported debris in order to meet the current licensing basis assumptions. However, these evaluations did not address debris effects fully for the as-built, as-operated plant conditions. The risk-informed approach evaluates the debris effects as part of the assessment of the residual risk associated with GSI-191 concerns.

The licensing basis for 10 CFR 50.46(a)(1) can be amended because the ECCS design has been demonstrated to be acceptable by meeting the risk guidelines in RG 1.174.

The exemption request to support closure of GL 2004-02 for PBN is intended to address ECCS cooling performance design as presented in 10 CFR 50.46(a)(1) as it relates to imposing the deterministic requirements in other properties. For the purposes of demonstrating the balance of the acceptance criteria of 10 CFR 50.46, the design and licensing basis descriptions of accidents requiring ECCS operation remain unchanged, as documented in PBN FSAR Chapters 6 and 14, including analysis methods, assumptions, and results. The performance evaluations for accidents requiring ECCS operation described in FSAR Chapters 6 and 14 are based on the Appendix K large-break LOCA (LBLOCA) analysis and demonstrate that, for breaks up to and including a LBLOCA, the ECCS will limit the clad temperature to below the acceptance criterion of 10 CFR 50.46 and the core will remain in place and substantially intact with its essential heat transfer geometry preserved.

The requirements of 10 CFR 50.46(a)(1) remain applicable to the model of record that meets the required features of Appendix K. Approval of the requested exemption does not impact the current ECCS evaluation. The ECCS evaluation model remains the licensing basis for demonstrating that the ECCS calculated cooling performance following postulated LOCAs meets the acceptance criteria.

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Enclosure 1 Request for Exemption The PBN risk-informed approach determines strainer and core blockage conditional failure probabilities and uses inputs from the PBN Probabilistic Risk Assessment (PRA) model to determine the CDF and LERF associated with debris-related failures, as described in Enclosure 4 of this submittal. The results show that PBN meets the Region III acceptance guidelines defined in RG 1.174 (Reference 2). The exemption request is limited to the requirement for demonstrating ECCS cooling performance design as required by 10 CFR 50.46(a)(1) as it pertains to the requirements for deterministic analyses described in other properties. It is not intended to be applicable to other requirements provided in 10 CFR 50.46 or Appendix K to 10 CFR 50.

As noted in Section 1.2, the NRC staff considers the modeling of sump performance to be an input to the ECCS evaluation model, and therefore the requirements of 10 CFR 50.46 are applicable. Consistent with this, the requirements and attributes for the proposed PBN risk-informed method include a full spectrum of postulated breaks, up to and including double ended guillotine breaks (DEGBs) on the largest reactor coolant system (RCS) pipes in containment, as described in Enclosures 3 and 4 of this submittal.

Engineering analyses and evaluations used to perform plant-specific prototypical testing consider a wide range of effects, including those addressed in NEI 04-07 (Reference 8) and its associated NRC SE (Reference 9) for evaluation of sump performance. The requested exemption does not affect any of the 10 CFR 50.46(a)(1) or Appendix K requirements for an acceptable ECCS evaluation model and does not change the ECCS acceptance criteria in 50.46(b) as it applies to the calculated results. Application of the exemption request allows the use of a risk-informed approach to evaluate the effects of debris. The results of the risk-informed method demonstrate that the risk associated with failures caused by LOCA-generated debris meets the acceptance guidelines of RG 1.174 (Reference 2). The current licensing basis for addressing the adequacy of the ECCS to meet the criteria of 10 CFR 50.46, including the Appendix K large-break LOCA analysis and the associated Chapter 14 accident analysis for LOCA, remains in place.

4.0 BASIS FOR THE EXEMPTION REQUEST Under 10 CFR 50.12, a licensee may request and the NRC may grant exemptions from the requirements of 10 CFR 50 that are authorized by law, will not present an undue risk to the public health and safety, are consistent with the common defense and security, and when special circumstances are present (see discussion in Section 4.1).

The exemption request meets a key principle of RG 1.174, which states, The proposed licensing basis change meets the current regulations unless it is explicitly related to a requested exemption (Reference 2). This exemption request is provided in conjunction with the proposed changes provided in the LAR (see Enclosure 2).

As required by 10 CFR 50.12(a)(2), the Commission will not consider granting an exemption unless special circumstances are present. Special circumstances are present whenever one of the items (i through vi) listed under 10 CFR 50.12(a)(2) are applicable (see discussion in Section 4.2).

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Enclosure 1 Request for Exemption NextEra has evaluated the requested exemption against the conditions specified in 10 CFR 50.12(a) and determined that this requested exemption meets the requirements for granting an exemption from the regulation, and that special circumstances are present. The information supporting the determination is provided below.

4.1 Applicability of 10 CFR 50.12(a)(1)

Pursuant to 10 CFR 50.12, Specific exemptions, the NRC may grant exemptions from the requirements of this part provided the following three conditions are met as required by 10 CFR 50.12(a)(1):

1. The exemption is authorized by law.

10 CFR 50.46(a)(1) was promulgated under and is consistent with the NRCs authority under the Atomic Energy Act of 1954 Section 161, as amended. Therefore, the NRC is authorized to grant 10 CFR 50.46(a)(1) exemptions if doing so would not violate the requirements of law. This exemption is authorized by law as 10 CFR 50.12 provides the NRC authority to grant exemptions from 10 CFR 50 requirements with provision of proper justification. Approval of the exemption from 10 CFR 50.46(a)(1), other properties, would not conflict with any provisions of the Atomic Energy Act of 1954, as amended, any of the Commissions regulations, or any other law.

2. The exemption does not present an undue risk to the public health and safety.

The purpose of 10 CFR 50.46 is to establish acceptance criteria for ECCS performance to provide a high confidence that the system will perform its required functions. The requested exemption does not involve any modifications to the plant that could introduce a new accident precursor or affect the probability of postulated accidents, and therefore the probability of postulated initiating events is not increased. The PRA and engineering analysis demonstrate that the calculated risk is very small (Enclosure 4) and consistent with the intent of the Commissions safety goal policy statement, which defines an acceptable level of risk that is a small fraction of other risks to which the public is exposed.

As discussed in previous 10 CFR 50.46 rulemaking, the probability of a LBLOCA is sufficiently low. Application of a risk-informed approach shows a high probability with low uncertainty that the ECCS will meet 10 CFR 50.46 requirements (Enclosure 4),

rather than using deterministic methods to achieve a similar understanding. This is applicable to evaluating acceptable containment sump design in support of ECCS and CSS recirculation modes.

The proposed change is to apply a risk-informed method rather than a deterministic method to establish a high probability of success for performance of ECCS in accordance with the requirements in 10 CFR 50.46(a)(1). The risk-informed approach involves a complete evaluation of the spectrum of LOCAs up to and E1-8

Enclosure 1 Request for Exemption including DEGBs on the largest pipe in the reactor coolant system, as described in Enclosure 4 of this submittal.

The risk-informed approach analyzes LOCAs, regardless of break size, using similar methods, assumptions, and criteria in order to quantify the uncertainties and overall risk metrics (Enclosure 4). This ensures that LBLOCAs with a low probability of occurrence and smaller break LOCAs with higher probability of occurrence are both considered in the results. Because the design-basis requirement for consideration of a DEGB of the largest pipe in the reactor coolant system is retained, the existing defense-in-depth and safety margin established for the design of the facility are not reduced.

This exemption only affects 10 CFR 50.46(a)(1), other properties, requirements that a licensee is able to demonstrate, using a bounding calculation or other deterministic method, that the ECCS and CSS are capable of functioning during a design basis event. This exemption does not impact the adequacy of the acceptance criteria for fuel cladding performance, which is important to maintain adequate safety margins.

3. The exemption is consistent with the common defense and security.

This exemption involves a change to the licensing basis for the plant that has no relation to the control of licensed material or any security requirements that apply to PBN. Therefore, the exemption is consistent with the common defense and security.

4.2 Applicability of 10 CFR 50.12(a)(2)

This section discusses the presence of special circumstances as related to 10 CFR 50.12(a). 10 CFR 50.12(a)(2) states that the NRC will not consider granting an exemption to the regulations unless special circumstances are present. Special circumstances are present whenever one of the listed items (i through vi) under 10 CFR 50.12(a)(2) are applicable.

Such special circumstances are present in this instance to warrant exemption from the requirements in 10 CFR 50.46(a)(1) other properties, which use deterministic calculation methods as the design basis for acceptable sump performance to validate the results of the ECCS evaluation model. Approval of this exemption request would allow the use of a risk-informed method to amend the design basis for acceptable performance of the containment emergency sump, as a validation of inputs in the ECCS evaluation model, and in support of the existing licensing bases for compliance with 10 CFR 50.46.

As described below, special circumstances in 10 CFR 50.12(a)(2)(ii) and 10 CFR 50.12(a)(2)(iii) are present as required by 10 CFR 50.12(a)(2) for consideration of the request for exemption.

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Enclosure 1 Request for Exemption 4.2.1 Applicability of 10 CFR 50.12(a)(2)(ii) 10 CFR 50.12(a)(2)(ii) applies:

Application of the regulation in the particular circumstances would not serve the underlying purpose of the rule or is not necessary to achieve the underlying purpose of the rule.

The intent of 10 CFR 50.46(a)(1) is to ensure ECCS cooling performance design requirements imposed by 10 CFR 50.46 are met using a rigorous method that provides a high level of confidence in ECCS performance. This exemption request is consistent with that purpose because use of the proposed risk-informed approach accounts for the effect of debris on the ECCS cooling performance and supports a high probability of successful ECCS performance, based on the risk results meeting the acceptance guidelines of RG 1.174 (Enclosure 4 and Reference 2).

The need for this exemption is based on the requirements in the regulations for using deterministic methods to demonstrate acceptable design. Regulatory requirements are largely based on a deterministic framework, and are established for DBAs, such as the LOCAs, with specific acceptance criteria that must be satisfied. Licensed facilities must be provided with safety systems capable of preventing and mitigating the consequences of DBAs to protect public health and safety. The deterministic regulatory requirements were designed to ensure that these systems are highly reliable. The LOCA analysis was established as part of this deterministic regulatory framework.

In comparison, the risk-informed approach considers nuclear safety in a more comprehensive way by examining the likelihood of a broad spectrum of initiating events and potential challenges, considering a wide range of credible events and assessing the risk based on mitigating system reliability.

An objective of 10 CFR 50.46 is to maintain low risk to the public health and safety through a reliable ECCS, as supported by the containment sump. The supporting analysis demonstrates that using a risk-informed approach to evaluate sump performance is consistent with the Commissions Safety Goals for nuclear power plants and supports ECCS operation with a high degree of reliability. Consequently, the determinstic approach described in 10 CFR 50.46(a)(2)(ii) is not necessary to achieve the underlying purpose and thereby, the special circumstances described in 10 CFR 50.12(a)(2)(ii) apply.

4.2.2 Applicability of 10 CFR 50.12(a)(2)(iii) 10 CFR 50.12(a)(2)(iii) applies:

Compliance would result in undue hardship or other costs that are significantly in excess of those contemplated when the regulation was adopted, or that are significantly in excess of those incurred by others similarly situated.

In order to meet a deterministic threshold value for sump debris loads, the debris sources in containment would need to be significantly reduced. The amount of E1-10

Enclosure 1 Request for Exemption radiological exposure received during the removal and/or modification of insulation from the PBN containment is dependent on the scope of the changes.

Due to uncertainties in radiation levels, contamination levels, and the required modification scope, it is difficult to predict the total occupational dose associated with insulation removal and/or modifications.

There is approximately 2000 ft3 of calcium silicate and asbestos calcium silicate insulation installed in each unit at PBN on the steam and feedwater lines and on piping adjacent to the steam generators and reactor coolant pumps. As part of a study for potential GL 2004-02 resolution alternatives, PBN considered replacement of the calcium silicate and asbestos calcium silicate insulation within the steam generator and reactor coolant pump compartments with reflective metal insulation (RMI). The total dose associated with performing this modification for both units was estimated to be approximately 900 rem. The estimate considered person-hours required to erect and remove scaffolding and the dose associated with removal and replacement of insulation. Additionally, each PBN unit has approximately 120 ft3 of mineral wool insulation on the resistance temperature detector (RTD) piping, which is in a high dose area. The dose for replacing the mineral wool insulation was estimated to be over 200 rem. Finally, the Unit 2 containment also has approximately 1500 ft3 of fiberglass insulation on the steam generators. Replacement of this insulation would incur additional dose to the estimate above.

The dose considerations discussed above demonstrate that compliance would result in substantial personnel exposure due to insulation modifications in containment, which is not commensurate with the expected safety benefit based on the risk evaluation results showing that the risk associated with post-accident debris effects is less than the threshold for Region III in RG 1.174 (Enclosure 4). Consequently, the special circumstances described in 10 CFR 50.12(a)(2)(iii) apply to the exemption requested by NextEra.

4.3 Environmental Consideration Pursuant to the requirements of 10 CFR 51.41, Requirement to submit environmental information, and 10 CFR 51.21, Criteria for and identification of licensing and regulatory actions requiring environmental assessments, the following information is provided. As demonstrated below, NextEra has determined that this exemption is eligible for categorical exclusion as set forth in 10 CFR 51.22, Criterion for categorical exclusion; identification of licensing and regulatory actions eligible for categorical exclusion or otherwise not requiring environmental review, paragraph (c)(9).

4.3.1 Environmental Impacts Consideration The requested exemption has been evaluated and determined to result in no significant radiological environmental impacts. This conclusion is based on the following.

The requested exemption is to allow the use of a risk-informed method to demonstrate that the design and licensing bases for the ECCS are not significantly affected by E1-11

Enclosure 1 Request for Exemption accident-generated and transported debris. The intent of the proposed change is to quantify the risk associated with debris effects. This quantification, provided in the form of risk metrics using the guidance in RG 1.174 (Reference 2), demonstrates that the risk is less than the threshold for Region Ill, Very Small Changes (Enclosure 4). Therefore, the requested exemption supports a change that represents a very small CDF and LERF, which is consistent with the Commissions Safety Goals for public health and safety.

Since the risk-informed analysis demonstrated that the increases in risk are very small, the requested exemption has a negligible effect on the consequences of an accident, and adequate assurance of public health and safety is maintained. The requested exemption does not involve any changes to the facility or facility operations that could create a new accident or release path, or significantly affect a previously analyzed accident or release path. Therefore, the requested exemption would not cause changes in the types or quantities of radiological effluents, or the permitted effluent release paths.

The requested exemption does not impact the release of radiological effluents during and following a postulated LOCA. The design-basis LOCA radiological consequence analysis in the current licensing basis is a deterministic evaluation based on the assumption of a major rupture of the reactor coolant system piping and a significant amount of core damage as specified in RG 1.183 (Reference 12). The current licensing basis analysis shows the resulting doses to the public and control room are acceptable.

The requested exemption does not change the radiological analysis for a LOCA.

Therefore, the requested exemption does not affect the amount of radiation exposure resulting from a postulated LOCA.

The requested exemption does not involve any changes to non-radiological plant effluents or any activities that would adversely affect the environment. The requested exemption only pertains to the licensing basis for components located within the restricted area of the facility, to which access is limited to authorized personnel.

Therefore, the requested exemption would not create any significant non-radiological impacts on the environment in the vicinity of the plant.

The requested exemption does not involve the use of any resources not previously considered by the NRC in its past environmental statements for issuance of the facility operating licenses or other licensing actions for the facility. Therefore, the requested exemption does not involve any unresolved conflicts concerning alternative uses of available resources.

4.3.2 Categorical Exclusion Consideration NextEra has evaluated the requested exemption against the criteria for identification of licensing and regulatory actions requiring environmental assessments in accordance with 10 CFR 51.21. It was determined that the requested exemption meets the criteria and is eligible for categorical exclusion as set forth in 10 CFR 51.22, Criterion for categorical exclusion; identification of licensing and regulatory actions eligible for categorical exclusion or otherwise not requiring environmental review, paragraph (c)(9).

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Enclosure 1 Request for Exemption This determination is based on the fact that this exemption request is from requirements under 10 CFR 50 with respect to the installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, specifically to authorize a change to the licensing basis for ECCS as it relates to acceptable containment sump performance in the recirculation mode following a postulated LOCA. The requested exemption has been evaluated to meet the following criteria under 10 CFR 51.22(c)(9).

(i) The exemption involves no significant hazards consideration.

An evaluation of the three criteria set forth in 10 CFR 50.92(c) as applied to the exemption is provided below. The evaluation is consistent with the no significant hazards consideration determination provided in the LAR (see Enclosure 2).

1. The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

The proposed change to allow an exemption from 10 CFR 50.46(a)(1) to implement a risk-informed evaluation methodology does not initiate an accident and therefore, the proposed change does not increase the probability of an accident occurring.

Approval of the requested exemption and accompanying LAR would allow the results of a risk-informed evaluation to be included in the PBN FSAR. The evaluation concludes that the ECCS and CSS will serve their safety functions with a high probability following a LOCA. The evaluation considers the impacts of accident-generated and transported debris on the containment emergency sump strainers in recirculation mode, as well as core blockage due to in-vessel effects.

The risk evaluation concludes that the risk associated with the proposed change is very small and within Region Ill as defined by RG 1.174, for both CDF and LERF (Enclosure 4). As a result, the required systems, structures, and components (SSCs) supported by the containment sumps will perform their safety functions with a high probability, and the proposed change does not alter or prevent the ability of SSCs to perform their intended function to mitigate the consequences of an accident previously evaluated within the acceptance limits. The safety analysis acceptance criteria in the FSAR continue to be met for the proposed change. Additionally, in accordance with the guidance of RG 1.174, there is substantial safety margin and defense-in-depth (Enclosure 5) that provide additional confidence that the design-basis functions are maintained.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of any of the accidents previously evaluated in the FSAR.

2. The proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

The proposed change is allowance of a risk-informed analysis of debris effects from accidents that are already evaluated in the PBN FSAR. No new or different kind of E1-13

Enclosure 1 Request for Exemption accident is created by the proposed change. No new failure mechanisms or malfunctions that can initiate an accident are created by the proposed change.

Therefore, the proposed change does not create the possibility for a new or different kind of accident from any accident previously evaluated.

3. The proposed change does not involve a significant reduction in a margin of safety.

The proposed change does not modify any functional requirements or method of performing functions of plant SSCs. The effects of debris are analyzed for a full spectrum of LOCAs, including DEGBs and partial breaks for all RCS piping sizes.

Appropriate redundancy, consideration of loss of offsite power, and worst-case single failure are retained such that defense-in-depth is maintained.

Application of the risk-informed methodology concludes that the increase in risk from the contribution of debris effects is very small as defined by RG 1.174 (Reference 2) and that there is adequate defense-in-depth and safety margin (Enclosure 5).

Consequently, PBN determined that the containment sumps would continue to support the safety-related components to perform their design functions when the effects of debris are considered.

The proposed change does not alter the manner in which safety limits are determined or the acceptance criteria associated with a safety limit. The proposed change does not implement any significant changes to plant operation that can challenge an SSCs capability to safely shut down the plant or maintain the plant in a safe shutdown condition. The proposed change does not affect the existing safety margins in the barriers for the release of radioactivity. There are no changes to any of the safety analyses in the PBN FSAR. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

(ii) The exemption involves no significant change in the types or significant increase in the amounts of any effluents that may be released offsite.

No physical modifications or changes to operating requirements are proposed for the facility as part of this exemption request, including any SSCs relied upon to mitigate the consequences of a LOCA. No changes are made to the safety analyses in the FSAR.

Approval of the exemption will require the calculated risk associated with post-accident debris effects to meet the Region III acceptance guidelines in RG 1.174 (Reference 2),

thereby maintaining public health and safety. Therefore, there is no significant change in the types or significant increase in the amounts of any effluents that may be released offsite.

(iii) The proposed exemption involves no significant increase in individual or cumulative occupational radiation exposure.

No new operator actions are implemented that could affect occupational radiation exposure. No physical modifications or changes to operating requirements are E1-14

Enclosure 1 Request for Exemption proposed for the facility as part of this exemption request, including any SSCs relied upon to mitigate the consequences of a LOCA. No changes are made to the safety analyses in the FSAR. Therefore, with respect to installation or use of a facility component located within the restricted area, approval of this exemption request will not result in a significant increase in individual or cumulative occupational radiation exposure.

Based on the above, NextEra concludes that the requested exemption meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).

5.0 TECHNICAL JUSTIFICATION FOR THE EXEMPTION Technical justification for the risk-informed method is provided in the accompanying LAR (Enclosure 2) and in Enclosures 3, 4, and 5.

The proposed risk-informed approach meets the key principles in RG 1.174 (Reference 2) in that it is consistent with the defense-in-depth philosophy, maintains sufficient safety margins, results in a very small increase in risk, and is monitored using performance measurement strategies. The requested exemption to allow use of the risk-informed method is consistent with the key principle in RG 1.174 that requires the proposed change to meet current regulations unless explicitly related to a requested exemption.

The PBN risk evaluation results (Enclosure 4) show that the risk associated with post-accident debris effects is within RG 1.174 Region Ill acceptance guidelines as a Very Small Change, and, therefore, is consistent with the Commissions Safety Goals for public health and safety.

6.0 CONCLUSION

Approval of the requested exemption to allow the use of a risk-informed approach will not present an undue risk to the public health and safety and is consistent with the common defense and security as required by 10 CFR 50.12(a)(1). Furthermore, special circumstances required by 10 CFR 50.12(a)(2) are present for 10 CFR 50.12(a)(2)(ii) and 10 CFR 50.12(a)(2)(iii). The requested exemption has been evaluated and determined to result in no significant radiological environmental impacts. Based on the determination that the risk of the exemption meets the acceptance guidelines of RG 1.174 (Reference 2), the results demonstrate reasonable assurance that the ECCS will function in the recirculation mode and that the public health and safety will be protected.

7.0 REFERENCES

1. NRC Generic Letter (GL) 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation during Design Basis Accidents at Pressurized-Water Reactors, September 13, 2004.

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Enclosure 1 Request for Exemption

2. Regulatory Guide 1.174, An Approach for Using Probabilistic Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, Revision 3, January 2018.

3. SECY-12-0093 (ML121310648), Closure Options for Generic Safety Issue - 191, Assessment of Debris Accumulation on Pressurized-Water Reactor Sump Performance, July 9, 2012.

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

5. ML103570354, 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.

6. ML12349A378, 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.

7. ML13140A013, Point Beach, Units 1 & 2, Resolution Option and Implementation Schedule for GSI-191 Closure, May 16, 2013.

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

9. NEI 04-07 Volume 2, Safety Evaluation by the Office of Nuclear Reactor Regulation Related to NRC Generic Letter 2004-02, Revision 0, December 2004.

10. SECY-16-0033 Enclosure 1 (ML15238B016), Federal Register Notice, Draft Final Rulemaking: Performance-Based Emergency Core Cooling Systems Cladding Acceptance Criteria, February 16, 2016.

11. ML17363A253, Point Beach Nuclear Plant, Units 1 and 2, Updated Final Response to NRC Generic Letter 2004-02, December 29, 2017.

12. Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors Revision 0, July 2000.

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Point Beach Nuclear Plant Licensing Submittal for a Risk-Informed Resolution of Generic Letter 2004-02 Enclosure 2 License Amendment Request Table of Contents 1.0 Summary Description ....................................................................................... 2 2.0 Detailed Description ......................................................................................... 2 2.1 System Design and Operation ................................................................... 2 2.2 Current Requirements................................................................................ 3 2.3 Reason for the Proposed Change ............................................................. 4 2.4 Description of the Proposed Change ......................................................... 4 3.0 Technical Evaluation ........................................................................................ 4 3.1 Engineering Analysis Overview ................................................................. 6 3.2 Conclusion for Technical Evaluation .......................................................... 6 4.0 Regulatory Evaluation ...................................................................................... 7 4.1 Applicable Regulatory Requirements/Criteria ............................................ 7 4.2 Precedent ................................................................................................ 10 4.3 No Significant Hazards Consideration ..................................................... 10 4.4 Conclusions for Regulatory Evaluation .................................................... 12 5.0 Environmental Consideration ......................................................................... 12 6.0 References ..................................................................................................... 13

Enclosure 2 Licensing Amendment Request Implementation of a Risk-Informed Approach for Addressing GL2004-02 1.0 Summary Description Pursuant to 10 CFR 50.90, NextEra Energy Point Beach LLC (NextEra) requests an amendment to Operating Licenses DPR-24 and DPR-27 for the Point Beach Nuclear Plant (PBN) Units 1 and 2. The proposed amendment will revise the licensing basis as described in the Point Beach Final Safety Analysis Report (FSAR) to allow the use of a risk-informed approach to address safety issues discussed in Generic Letter (GL) 2004-02 Potential Impact of Debris Blockage on Emergency Recirculation during Design Basis Accidents for Pressurized-Water Reactors (Reference 1) for PBN. The risk-informed approach is consistent with the guidance of Nuclear Regulatory Commission (NRC) Regulatory Guide (RG) 1.174 (Reference 2) and SECY-12-0093 (Reference 3).

The proposed changes will apply only for the effects of accident-generated debris as described in GL 2004-02 (Reference 1).

2.0 Detailed Description GL 2004-02 identified the possibility that debris generated during a loss of coolant accident (LOCA) could clog the containment recirculation sump strainers in pressurized water reactors (PWRs) and result in loss of net positive suction head (NPSH) for the emergency core cooling system (ECCS) and containment spray system (CSS) pumps, impeding the flow of water from the sump. Additionally, debris that passes through the strainer could affect safety functions of the components downstream of the strainer or challenge long-term core cooling due to debris accumulation in the reactor core.

GL 2004-02 requested the PWR licensees to address these issues and demonstrate compliance with the ECCS acceptance criteria in 10 CFR 50.46 by performing analyses using an NRC-approved methodology. PBNs latest response to GL 2004-02 is provided in Enclosure 3 of this submittal.

2.1 System Design and Operation A fundamental function of emergency core cooling is to recirculate water that has collected in the containment sump following a break in the reactor coolant system (RCS) piping to ensure long-term removal of decay heat from the reactor fuel. This section summarizes the design and operation of the PBN ECCS and CSS.

Emergency Core Cooling As stated in the PBN FSAR (Reference 6), adequate emergency core cooling is provided by the safety injection (SI) system (which constitutes the ECCS). The primary purpose of the SI system is to automatically deliver cooling water to the reactor core in the event of a LOCA. This protection is afforded for:

a. All pipe break sizes up to and including the hypothetical instantaneous circumferential rupture of a reactor coolant loop, assuming unobstructed discharge from both ends.

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Enclosure 2 Licensing Amendment Request

b. A loss of coolant associated with rod ejection accident.

c. A steam generator tube rupture.

The principal components of the SI system which provide emergency core cooling immediately following a LOCA are two accumulators, the two SI (high head) pumps and the two residual heat removal (RHR) (low head) pumps.

The accumulators are passive components and discharge into the cold legs of the reactor coolant piping when pressure decreases to about 750 psig. This assures rapid cooling of the core for large breaks.

The two high head SI pumps start on a SI signal and supply borated water to the RCS.

During the injection phase, SI pumps take suction from the refueling water storage tank (RWST) and deliver coolant through two separate headers into the RCS.

The two RHR pumps are used to inject borated water at low pressure to the RCS. The SI signal starts the RHR pumps and opens the low head injection line isolation valves.

During the injection phase RHR pumps take suction from the RWST and deliver coolant through two nozzles that penetrate the reactor vessel and core barrel.

After the injection phase, coolant spilled from the break and water collected from the containment spray is cooled and returned to the RCS by the RHR pumps which are aligned to take suction from the containment recirculation sump. Additionally, during the recirculation phase, RHR pumps are used to recirculate fluid to the suction of the containment spray pump or to the suction of the SI pump.

The RWST serves as a source of emergency borated cooling water for injection and containment spray.

Containment Spray System The CSS consists of two containment spray pumps, spray ring headers and nozzles, spray additive tank, valves and the necessary piping, instrumentation, and controls.

During the injection phase after an accident, the RWST supplies borated water to the CSS. During the recirculation phase of the accident, the CSS pump takes suction from the RHR pump. The primary purpose of the CSS is to depressurize the containment and remove elemental iodine and particulates from the atmosphere.

2.2 Current Requirements 10 CFR 50.46(a)(1) requires that the criteria set forth in 10 CFR 50.46 paragraph (b), be demonstrated with ECCS cooling performance calculated in accordance with an acceptable evaluation model and includes a requirement for other properties with regard to the methodology for showing those requirements are met. The methodology is governed by 10 CFR 50.46(a)(1) and is deterministic with no provision for a risk-informed approach. This license amendment request (LAR) supports an exemption to 10 CFR 50.46(a)(1) as described in Enclosure 1.

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Enclosure 2 Licensing Amendment Request 2.3 Reason for the Proposed Change In order to meet a deterministic threshold value for sump debris loads, the debris sources in containment would need to be significantly reduced. The amount of radiological exposure received during the removal and/or modification of insulation from the PBN containment is dependent on the scope of the changes. As discussed in of this submittal, the expected dose for replacing calcium silicate and asbestos calcium silicate insulation in the PBN containment is estimated to be approximately 900 rem for both units (total two-unit dose). An additional dose of 200 rem was estimated for replacing the mineral wool insulation on the resistance temperature detector (RTD) lines. This estimate does not include dose associated with replacement of fiberglass insulation on the Unit 2 steam generators.

The dose considerations discussed above demonstrate that compliance would result in substantial personnel exposure due to insulation modifications in containment, which is not commensurate with the expected safety benefit based on the risk evaluation results showing that the risk associated with post-accident debris effects is less than the threshold for Region III in Regulatory Guide (RG) 1.174 (Reference 2).

2.4 Description of the Proposed Change The proposed change in methodology in this LAR is to use a risk-informed approach to address the effects of accident-generated and transported debris on the containment emergency sumps instead of a deterministic approach. The details of the risk-informed approach are provided in Enclosure 4 of this submittal. The debris analysis covers a full spectrum of postulated LOCAs, including partial breaks and double-ended guillotine breaks (DEGBs), to provide assurance that the most severe postulated LOCAs are evaluated. of this enclosure provides markups to the PBN FSAR, which includes revision of applicable FSAR sections and design bases descriptions that take credit for the risk-informed evaluation, which will be described in FSAR Appendix A.8. The FSAR markups are provided for information only.

3.0 Technical Evaluation The methodology change affects the analysis of the effects of accident-generated debris on plant systems that are supported by the containment recirculation sumps and strainers during the recirculation phase of LOCA mitigation. These include the ECCS and CSS.

The primary goal of the risk-informed analysis is to quantify the risk increase due to strainer and reactor failures caused by accident-generated debris. A break-specific analysis was first performed to identify breaks that would fail the strainer due to exceeding any one of the following acceptance criteria:

x Debris limits based on PBN-specific strainer head loss testing x Pump NPSH margin E2-4

Enclosure 2 Licensing Amendment Request x Strainer structural margin (i.e., differential pressure limit) x Degasification limit due to void fraction and flashing x Partially-submerged strainer limit. Note that this criterion was implemented in the software used for the PBN risk analysis (i.e. NARWHAL) and automatically applied in the PBN model, but did not result in any failures as the strainers are fully submerged at the start of switchover to sump recirculation.

Bounding analyses were also performed to determine if any of the breaks would fail the following limits.

x Ex-vessel blockage and wear limits (ex-vessel downstream effects) x In-vessel downstream effects limits.

The analyses against the above limits considered possible equipment lineups during sump recirculation and resulted in conditional failure probabilities (CFPs) for strainer and reactor core failures caused by accident-generated debris. These CFP values, along with relevant LOCA frequencies from the PBN probabilistic risk assessment (PRA) model and equipment failure probabilities, were used to calculate the change in core damage frequency (CDF) associated with the effects of debris. In addition, the conditional large early release probability (CLERP) determined from the PBN PRA model was used to calculate the change in large early release frequency (LERF).

These values, along with the PBN baseline CDF and LERF, were used to determine the risk region based on the acceptance guidelines in RG 1.174 (Reference 2).

The results of the PBN evaluation show that the risk from the proposed change is "very small" in that it is in Region Ill of RG 1.174. These results meet the requirement for the risk from debris to be small in paragraph (e) of the proposed 10 CFR 50.46c rule change (Reference 4) and associated draft RG 1.229 (Reference 5). See Enclosure 4 for a more detailed description of the risk quantification.

The proposed PBN FSAR Appendix A.8 (see Attachment 1) describes the risk-informed approach used to confirm that there is a high probability the ECCS and CSS will perform their required function following a LOCA when considering the impacts of accident-generated debris. This new appendix identifies the key elements of the risk-informed analyses. Future changes to the key elements are to be evaluated as a potential departure from a method of evaluation described in the FSAR in accordance with 10 CFR 50.59(c)(2)(viii). The key elements include:

1. The methodology used to quantify the amount of debris generated at each break location, including the assumed zone of influence (ZOI) size based on the target destruction pressure and break size, and the assumed ZOI shape (spherical or hemispherical) based on whether the break is a DEGB or partial break (see the Responses to 3.a and 3.b in Enclosure 3).

2. The methodology used to evaluate debris transport to the containment sump recirculation strainers (see the Response to 3.e in Enclosure 3).

3. The methodology used to quantify chemical precipitates and determine the solubility of aluminum (see the Response to 3.o in Enclosure 3).

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Enclosure 2 Licensing Amendment Request

4. The strainer debris limits shown in Appendix A.8, which are based on tested and analyzed debris quantities (see Enclosure 3, Response to 3.f). Any changes to these debris limits are subject to 10 CFR 50.59 as well as 10 CFR 50.71(e) reporting requirements.

5. The methodology and acceptance criteria used to assess ex-vessel component blockage and wear (see the Response to 3.m in Enclosure 3).

6. The methodology used to assess in-vessel fiber accumulation and the associated limits (see the Response to 3.n in Enclosure 3).

7. The methodology used to quantify CDF and LERF (see Enclosure 4).

The performance evaluations for accidents requiring ECCS operation are described in FSAR Chapter 14 (Reference 6), based on the PBN Appendix K large-break LOCA analysis. System redundancy, independence, and diversity features are not changed for those safety systems credited in the accident analyses. No new programmatic compensatory activities or reliance on manual operator actions are required to implement this change.

3.1 Engineering Analysis Overview The design and licensing basis descriptions of accidents requiring ECCS and CSS operation, including analysis methods, assumptions, and results provided in FSAR Chapter 14 remain unchanged. The methodology for calculating the risk associated with failures caused by accident-generated debris evaluates a full spectrum of breaks up to and including DEGBs for all RCS pipe sizes. The results show that the risk is "very small" as defined by Region Ill in RG 1.174 (Reference 2). The detailed technical description of the risk quantification process is presented in Enclosure 4. Additionally, the functionality of the ECCS and CSS during design basis accidents is confirmed by demonstrating that safety margin and defense-in-depth (DID) are maintained with high probability (Enclosure 5).

This LAR is requesting a change to the licensing basis such that the effects of LOCA generated and transported debris can be evaluated using a risk-informed methodology.

Detailed evaluations of DID and safety margin are presented in Enclosure 5. The evaluations determined that there is substantial DID and safety margins that provide a high level of confidence that the calculated risk is conservative and that the actual risk is likely much lower.

3.2 Conclusion for Technical Evaluation The technical evaluation shows that the functionality of the ECCS and CSS during design basis accidents is confirmed by the very small risk increase due to strainer failures associated with debris effects, supported by the fact that the safety margin and DID are maintained with high probability.

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Enclosure 2 Licensing Amendment Request 4.0 Regulatory Evaluation 4.1 Applicable Regulatory Requirements/Criteria Approval of the proposed amendment is contingent upon approval of the request for exemption from certain aspects of 10 CFR 50.46(a)(1) as provided and justified in of this submittal.

Regulatory Guide 1.174 NRC RG 1.174 (Reference 2) provides the NRC staffs recommendations for using risk information in support of licensee-initiated licensing basis changes to a nuclear power plant that require NRC review and approval. This RG describes an acceptable approach for assessing the nature and impact of proposed licensing basis changes by considering engineering issues and applying risk insights.

In implementing risk-informed decision-making, licensing basis changes are expected to meet a set of key principles. These principles include the following:

(1) The proposed change meets the current regulations unless it is explicitly related to a requested exemption (i.e., a specific exemption under 10 CFR 50.12, "Specific Exemptions").

The exemption requested in Enclosure 1 of this submittal complies with this requirement.

(2) The proposed change is consistent with a DID philosophy.

Defense-in-depth is presented in detail in Enclosure 5 of this submittal. The proposed change is consistent with the DID philosophy in that the following aspects of the facility design and operation are unaffected:

x Functional requirements and the design configuration of systems x Existing plant barriers to the release of fission products x Design provisions for redundancy, diversity, and independence x Plant response to transients or other initiating events x Preventive and mitigative capabilities of plant design features The PBN risk-informed approach analyzes a full spectrum of postulated LOCAs, including DEGBs for all piping sizes up to and including the largest pipe in the RCS. By requiring that mitigative capability be maintained in a risk-informed evaluation of debris effects for a full spectrum of LOCAs, the approach ensures that DID is maintained.

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Enclosure 2 Licensing Amendment Request (3) The proposed change maintains sufficient safety margins.

As described in Enclosure 5 of this submittal, sufficient safety margins associated with the design will be maintained by the proposed change.

(4) When proposed changes result in an increase in CDF or risk, the increases should be small and consistent with the intent of the Commission's safety goal policy statement.

The proposed change involves evaluation of the risk associated with strainer and reactor failures caused by accident-generated and transported debris using a risk-informed methodology. Using engineering analysis and information from the PRA model, this risk is shown to be "very small" as defined by Region Ill in RG 1.174 (Reference 2) and is therefore consistent with the Commission's safety goal policy statement.

(5) The impact of the proposed change should be monitored using performance measurement strategies.

PBN has implemented procedures and programs for monitoring, controlling, and assessing changes to the plant that have a potential impact on plant performance related to effects of accident-generated debris. The following procedures and programs provide the capability to monitor the performance of the sump strainers and assess impacts to the inputs and assumptions used in the engineering analyses that support the proposed change.

x The PBN TS surveillance (SR 3.5.2.6) requires visual inspection to ensure each ECCS train containment sump suction inlet is not restricted by debris and the suction inlet strainers show no evidence of structural distress or abnormal corrosion.

x The PBN design change process requires the use of Design Attribute Review (DAR) forms to identify potential impact on long term core cooling by the proposed modification. Questions in the DAR forms that are related to GL 2004-02 include:

o Does the modification affect insulation?

o Does the modification add or remove components in containment?

o Does the modification change the amount of exposed aluminum and/or zinc in containment?

o Does the modification introduce materials that could affect sump performance or lead to equipment degradation?

o Does the modification repair, replace, or install coatings inside containment, including installing coated equipment?

o Does the modification affect installation, replacement, or storage of any structure, system, component or other items in containment that has vendor applied or site applied protective coatings?

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Enclosure 2 Licensing Amendment Request o Does the modification affect high/moderate energy line break analysis?

o Does the modification affect the design, performance or operation of pumps?

o Does the modification affect foreign material that would require cleaning to prevent degradation of downstream components?

x A 10 CFR 50.59 screening or evaluation is required to be completed for all design changes.

x The PBN containment coatings program monitors and assesses the containment building coatings, and establishes administrative controls for conducting coating examinations, deficiency reporting, and documentation.

x As part of the PBN condition reporting process, condition reports are written when adverse conditions are identified during containment inspections or during surveillances of the containment emergency sumps and strainers. Documentation and evaluation of nonconformances are discussed in Enclosure 4.

x The PBN Maintenance Rule program includes performance monitoring of any high safety significant functions associated with the ECCS and CSS.

The Maintenance Rule program provides continued assurance of the availability and reliability for performance of the required functions.

x The on-line configuration risk management procedure establishes the administrative controls for performing on-line maintenance of structures, systems, and components (SSCs) to enhance overall plant safety and reliability.

x The PBN quality assurance (QA) program is implemented and controlled in accordance with the Quality Assurance Topical Report (QATR) and is applicable to SSCs to an extent consistent with their importance to safety.

The QA program complies with the requirements of 10 CFR 50, Appendix B and other program commitments as appropriate.

The proposed change does not alter the ASME Section Xl inspection programs or mitigation strategies that have been shown to be effective in early detection and mitigation of weld and material degradation in Class I piping applications.

Periodic updates to the risk-informed analysis will be performed to capture the effects of any plant changes, procedure changes, or new information on the risk-informed analysis and to confirm that the results are still within the established acceptance criteria (see Enclosure 4).

Regulatory Guide 1.200 NRC RG 1.200 (Reference 7) describes one acceptable approach for determining whether the quality of the PRA, in total or the parts that are used to support an application, is sufficient to provide confidence in the results, such that the PRA can be used in regulatory decision-making for light water reactors. As described in Enclosure 4 E2-9

Enclosure 2 Licensing Amendment Request of this submittal, the PBN PRA model used for the risk-informed GL 2004-02 evaluation complies with RG 1.200 (Reference 7).

4.2 Precedent The proposed licensing change for PBN is very similar to the license amendment and 10 CFR 50.46(a)(1) exemption granted to South Texas Project (STP) Nuclear Operating Company, STP Units 1 and 2 (References 8 and 9) and to Southern Nuclear Operating Company, Vogtle Electric Generating Plant (VEGP) Units 1 and 2 (Reference 10) for implementation of the risk-informed approach to address GL 2004-02 concerns.

PBN requests implementation of a risk-informed methodology for resolution of GL 2004-02 that is similar to the approach used by VEGP. Key similarities include, but are not limited to:

1. Use of RG 1.174 acceptance guidelines and key principles.

2. Identification of key methods and approaches in the risk-informed methodology that, if changed after implementation, are to be evaluated as a potential departure from a method of evaluation described in the FSAR under 10 CFR 50.59.

3. Associated request for exemption from 10 CFR 50.46(a)(1) other properties.

4. Use of NARWHAL software for risk analysis

5. Risk quantification uses conditional failure probability methodology One key difference between the PBN methodology and VEGP is that the risk contribution from secondary side breaks at PBN was considered using a bounding evaluation rather than a break specific analysis.

4.3 No Significant Hazards Consideration The proposed amendments implement a risk-informed approach to address the effects of accident-generated and transported debris on the containment emergency sumps.

NextEra has evaluated whether a significant hazards consideration is involved with the proposed change by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

(1) Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No The proposed change is a methodology change for assessment of debris effects that adds the results of a risk-informed evaluation to the PBN licensing basis.

This is a viable approach for the resolution of GL 2004-02 per SECY-12-0093 (Reference 3). The analysis that supports the methodology change concludes that the functionality of the ECCS and CSS during design basis accidents is E2-10

Enclosure 2 Licensing Amendment Request confirmed by the very small risk increase due to strainer failures associated with the debris effects, supported by the fact that the safety margin and DID are maintained with high probability.

There is no significant increase in the probability of an accident previously evaluated. The proposed change addresses mitigation of LOCAs and has no effect on the probability of the occurrence of a LOCA. The proposed methodology change does not implement any changes in the facility or plant operation that could lead to a different kind of accident. The containment sump is not an initiator of any accident previously evaluated. The containment sump is a passive component and the proposed change does not increase the likelihood of the malfunction. The design and the capability of the containment sump assumed in the accident analysis is not changed. As a result, the probability of an accident is unaffected by the proposed change.

The proposed change does not involve a significant increase in the consequences of an accident previously evaluated. The methodology change confirms that required SSCs supported by the containment sumps will perform their safety functions with a high probability, as required, and does not alter or prevent the ability of SSCs to perform their intended function to mitigate the consequences of an accident previously evaluated within the acceptance limits.

The proposed change has no impact on existing barriers that prevent the release of radioactivity. The safety analysis acceptance criteria in the FSAR continue to be met for the proposed methodology change.

(2) Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No The proposed change is a methodology change for assessment of debris effects that adds the results of a risk-informed evaluation to the PBN licensing basis.

The proposed change does not install or remove any plant equipment, or alter the design, physical configuration, or mode of operation of any plant SSCs. The proposed change does not introduce any new failure mechanisms or malfunctions that can initiate an accident. No new credible accident is created that is not encompassed by the existing accident analyses that assumes the function of the containment sump.

Therefore, the proposed methodology does not create the possibility of a new or different kind of accident from any previously evaluated.

(3) Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No E2-11

Enclosure 2 Licensing Amendment Request The proposed change is a methodology change for assessment of debris effects that adds the results of a risk-informed evaluation to the PBN licensing basis.

The effects from a full spectrum of LOCAs and secondary side breaks inside containment, including DEGBs, are analyzed. Appropriate redundancy and consideration of loss of offsite power and worst-case single failure are retained, such that DID is maintained.

Application of the risk-informed methodology showed that the increase in risk from the contribution of debris effects is very small as defined by RG 1.174 (Reference 2) and that there is adequate DID and safety margin, which are extensively evaluated in Enclosure 5. This evaluation showed that there is substantial DID and safety margin that provide a high level of confidence that the calculated risk for the effects of debris is conservative and that the actual risk is likely much lower. Consequently, PBN determined that the risk-informed method demonstrates the containment sumps will continue to support the ability of safety-related components to perform their design functions when the effects of debris are considered. Note that the risk-informed approach was identified as viable for the response to GL 2004-02 per SECY-12-0093 (Reference 3).

The proposed change does not alter the manner in which safety limits are determined or the acceptance criteria associated with a safety limit. The proposed change does not implement any changes to plant operation and does not affect SSCs that respond to safely shutdown the plant and to maintain the plant in a safe shutdown condition. The proposed change does not significantly affect the existing safety margins in the barriers for the release of radioactivity.

There are no changes to any of the safety analyses in the FSAR.

Therefore, the proposed methodology does not involve a significant reduction in a margin of safety.

Based on the above, PBN concludes that the proposed amendment does not involve a significant hazards consideration under the standards in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

4.4 Conclusions for Regulatory Evaluation In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the license amendment will not be inimical to the health and safety of the public.

5.0 Environmental Consideration A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted E2-12

Enclosure 2 Licensing Amendment Request area, as defined in 10 CFR 20. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or a significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

6.0 References

1. NRC Generic Letter (GL) 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation during Design Basis Accidents for Pressurized-Water Reactors, September 13, 2004.

2. Regulatory Guide 1.174, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, Revision 3, January 2018.

3. SECY-12-0093 (ML121310648), Closure Options for Generic Safety Issue - 191, Assessment of Debris Accumulation on Pressurized-Water Reactor Sump Performance, July 9, 2012.

4. ML15238A947, SECY-16-0033: Final Draft Rulemaking - 10 CFR 50.46c:

Emergency Core Cooling Systems Performance During Loss-of-Coolant Accidents, April 4, 2016.

5. Draft Regulatory Guide 1.229 (ML16062A016), Risk-Informed Approach for Addressing the Effects of Debris on Post-Accident Long-Term Core Cooling, March 2016.

6. Point Beach Nuclear Plant Units 1 & 2, Final Safety Analysis Report 2020.

7. Regulatory Guide 1.200, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities, Revision 2, March 2009.

8. ML17019A001, South Texas Project, Units 1 and 2 - Issuance of Amendment Nos.

212 and 198 - Risk-Informed Approach to Resolve Generic Safety Issue 191 (CAC Nos. MF2400 and MF2401), July 12, 2017.

9. ML17037C871, South Texas Project, Units 1 and 2, Exemptions from the Requirements of 10 CFR Part 50, Section 50.46 and 10 CFR Part 50, Appendix A, General Design Criteria 35, 38, and 41 (CAC Nos. MF2402-MF2409), July 11, 2017.

10. ML20230A346, Southern Nuclear Operating Co, Inc - Exemption Request and License Amendment Request for a Risk-Informed Resolution to GSI-191, August 17, 2020.

E2-13

Point Beach Nuclear Plant Licensing Submittal for a Risk-Informed Resolution of Generic Letter 2004-02 Enclosure 2 License Amendment Request Attachment 1 Proposed FSAR Changes (Mark-Up) for Information Only

Safety Injection System (SI)

FSAR Section 6.2 6.2 SAFETY INJECTION SYSTEM (SI) 6.2.1 DESIGN BASIS Redundancy of Reactivity Control Criterion: Two independent reactivity control systems, preferably of different principles, shall be provided. (GDC-27)

In addition to the reactivity control achieved by the rod cluster control (RCC) described in Section 3.0, and the chemical and volume control system described in Chapter 9, the safety injection system provides an alternative boration path for shutdown reactivity control.

The refueling water storage tank may be aligned to the suction of the safety injection pumps as an alternative to the CVCS system. Use of this lineup requires reactor coolant system pressure to be less than the shutoff head of the safety injection pumps.

Emergency Core Cooling System Capability Criterion: An emergency core cooling system with the capability for accomplishing adequate emergency core cooling shall be provided. This core cooling system and the core shall be designed to prevent fuel and clad damage that would interface with the emergency core cooling function and to limit the clad metal-water reaction to acceptable amounts for all sizes of breaks in the reactor coolant piping up to the equivalent of a double-ended rupture of the largest pipe. The performance of such emergency core cooling system shall be evaluated conservatively in each area of uncertainty. (GDC 44)

Adequate emergency core cooling is provided by the safety injection system (which constitutes the emergency core cooling system) which operates in three modes. These modes are delineated as passive accumulator injection, active safety injection and residual heat removal recirculation.

The primary purpose of the safety injection system is to automatically deliver cooling water to the reactor core in the event of a loss-of-coolant accident. This limits the fuel clad temperature and thereby ensures that the core will remain intact and in place with its heat transfer geometry preserved. This protection is afforded for:

1. All pipe break sizes up to and including the hypothetical instantaneous circumferential rupture of a reactor coolant loop, assuming unobstructed discharge from both ends.

Debris effects on the safety injection system are evaluated using a risk informed approach for the full range of break sizes, as discussed in Appendix A.8.

2. A loss of coolant associated with the rod ejection accident.

3. A steam generator tube rupture.

The basic design criteria for loss-of-coolant accident evaluations are: (Reference 2)

1. The calculated peak cladding temperature shall not exceed 2200qF.

2. The calculated total oxidation of the cladding shall nowhere exceed 0.17 times the total cladding thickness before oxidation.

UFSAR 2020 Page 6.2-1 of 49

Safety Injection System (SI)

FSAR Section 6.2

3. The calculated total amount of hydrogen generated from the chemical reaction of the cladding with water or steam shall not exceed 0.01 times the hypothetical amount generated if all the cladding directly surrounding the fuel were to react.

4. Calculated changes in the core geometry shall be such that the core remains amenable to cooling.

5. After the initial successful operation of the ECCS, the calculated core temperature shall be maintained at an acceptable low value and decay heat shall be removed for the extended period of time required by the long-lived radioactivity remaining in the core.

For any rupture of a steam pipe and the associated uncontrolled heat removal from the core, the safety injection system adds shutdown reactivity so that with a stuck rod, no off-site power and minimum engineered safety features, there is no consequential damage to the reactor coolant system and the core remains in place and intact.

Redundancy and segregation of instrumentation and components are incorporated to assure that postulated malfunctions will not impair the ability of the system to meet the design criteria. The system is effective in the event of loss of normal plant auxiliary power coincident with the loss of coolant, and can accommodate the failure of any single component or instrument channel to respond actively in the system. During the recirculation phase of a loss-of-coolant accident, the system can accommodate a loss of any part of the flow path since backup alternative flow path capability is provided.

The ability of the safety injection system to meet its design criteria is presented in Section 6.2.3.

The analysis of the accidents is presented in Section 14.0. The design basis for the safety injection system with regard to the effects of accident-generated debris on the containment sump recirculation strainers, to the extent that the strainers support the post-accident core cooling function, is a risk-informed analysis that shows there is a high probability that the safety injection system can perform its design basis functions. This conclusion is based on plant-specific testing and analyses using assumptions that provide safety margin and defense-in-depth. The risk of strainer failures caused by accident-generated debris is very small and is acceptable in accordance with the guidelines of RG 1.174.

Details of the design basis for the effects of debris on the function of the emergency sump strainers are provided in FSAR Appendix A.

Inspection of Emergency Core Cooling System Criterion: Design provisions shall, where practical, be made to facilitate inspection of physical parts of the emergency core cooling system, including reactor vessel internals and water injection nozzles. (GDC 45)

Design provisions are made to facilitate access to the critical parts of the reactor vessel internals, injection nozzles, pipes, valves and safety injection pumps for visual or boroscopic inspection for erosion, corrosion and vibration wear evidence, and for nondestructive inspection where such techniques are desirable and appropriate.

UFSAR 2020 Page 6.2-2 of 49

Safety Injection System (SI)

FSAR Section 6.2 Protection against containment over-pressure following a loss-of-coolant accident or a steam line break accident is provided by the containment air recirculation cooling system (Section 6.3) and the containment spray system (Section 6.4).

Recirculation Phase After the injection phase, coolant spilled from the break and water collected from the containment spray is cooled and returned to the reactor coolant system by the residual heat removal pumps which are aligned to take suction on the containment recirculation sump. This water is pumped back to the core and/or to the suction of the containment spray pumps through the residual heat removal heat exchangers. The system is arranged to allow either or both of the residual heat removal pumps to take over the recirculation function.

The recirculation sump lines consist of two independent and redundant 10 in. lines which penetrate the containment. Each line has one remote hydraulically-operated valve located inside containment, and one remote motor-operated valve located outside containment. Each line is run independently to the suction of a residual heat removal pump. The 10 in. drain pipes pass through sleeves in the containment structure concrete. The sleeves are welded to the liner plate and to the drain pipe with all welds inspectable. The drains pass through a second set of sleeves between the tendon gallery and the auxiliary building. The system permits long-term recirculation in the event of a passive or active component failure.

Alternative flow paths are also provided from the discharge of the residual heat removal heat exchangers for both low and high head recirculation. This is evaluated in Section 6.2.3.

The design of the containment drains are shown in Figure 6.2-2 and Figure 6.2-3. As illustrated, the containment building serves as a sump that collects the spilled coolant, injected water and containment spray system drainage. This collected water is used during the recirculation phase.

During recirculation operation the collected water is filtered through a strainer assembly over each drain before leaving the containment sump. The individual cross sectional filter flow areas in each strainer assembly are no greater than a nominal 0.066 inch diameter opening. The size of the strainer openings restricts any sizable foreign matter from entering the recirculation system.

See Appendix A for further discussion on how long-term cooling requirements and 10 CFR 50.46 ECCS acceptance criteria are addressed for post-accident debris effects on the strainers.

The high head recirculation flow path via the high head safety injection pumps is required for the range of small break sizes for which the reactor coolant system pressure remains in excess of the shutoff head of the residual heat removal pumps at the end of the injection phase. The high head recirculation flow path is also required following a large break LOCA to control boric acid precipitation in the reactor vessel.

Those portions of the safety injection system located outside of the containment which are designed to circulate, under post accident conditions, radioactively contaminated water collected in the containment, meet the following requirements:

1. Shielding to maintain radiation levels within the limits set forth in 10 CFR 50.67. See Section 11.6.

2. Collection of discharges from pressure relieving devices into closed systems.

UFSAR 2020 Page 6.2-6 of 49

Safety Injection System (SI)

FSAR Section 6.2 Each RHR pump is located in an individual compartment which is equipped with a floor drain and separated equipment drains. The floor drain from each compartment flows through an individual pipe to the sump. Two 75 gpm sump pumps transfer the leakage to the waste disposal system.

Valving is provided to permit the operator to individually isolate the residual heat removal pumps.

The supply and discharge piping and valves for the RHR pumps are located in a pipeway adjacent to the pump compartments. A seven foot high shield wall divides the pipeway into two sections, each of which drains into a pump compartment through a 4-inch by 4-inch opening at floor level.

Openings in the wall have no effect on RHR pump protection from flooding events. The RHR pump seal failure rate is 50 gpm.

The RHR cubicle drain valves are maintained in the closed position. If a RHR pump seal failure occurred with the drain valves in the closed position, a RHR pump room high level alarm would eventually be indicated in the control room. The cubicle could then be drained to the sump by opening the drain valve. If flooding in EL.-19' occurred due to a source other than a failed RHR pump seal, the fluid would collect in the center cubicle (cubicle between the Unit 1 and Unit 2 RHR pumps) and flow to the sump via the floor drains. The flow path to the RHR pump cubicle would remain isolated.

Pump NPSH Requirements - Residual Heat Removal Pumps The NPSH of the residual heat removal pumps is evaluated for normal plant shutdown operation, and both the injection and recirculation phase operation of the design basis accident.

Recirculation operation gives the limiting NPSH requirement. The available NPSH is determined from the containment water level, and the pressure drop in the suction piping from the sump to the pumps, as described in Appendix A. During recirculation phase of a large break LOCA where RHR pump flow is sent to both the reactor vessel and the suction of the containment spray pump, maximum RHR pump flow requirements are set by system alignment to ensure RHR pump NPSH. Status lights are available on the main control boards to allow the operator to confirm the proper alignment of the containment spray pump discharge valves and to confirm that the preset throttle position has been reached for the SI-852A & B RHR pump core deluge valves. Flow instrumentation is available on the main control boards to allow the operators to monitor the operation of the containment spray and RHR systems during the ECCS recirculation phase of a LOCA. (Reference 4)

Coating debris can also play a role in affecting the available NPSH during post-LOCA ECCS recirculation operation. A program has been instituted at PBNP that provides adequate assurance that the applicable requirements for the procurement, application, inspection, and maintenance of Service Level I coatings in containment are implemented, and that maintains a detailed inventory of degraded and non-conforming coatings to ensure the coatings are maintained within the evaluated limits of design basis analyses for the ECCS. Refueling frequency coatings inspections ensure the total inventory of coatings remain bounded by the analyses. The impact of failed coatings on the performance of containment recirculation sump strainers during post-accident operation is analyzed in Appendix A. The debris limits for coatings are established based on tested and analyzed quantities, as shown in Appendix A.

Safety Injection Pumps The NPSH for the safety injection pumps is evaluated for both the injection and recirculation phase of operation of the design basis accident. The end of the injection phase operation gives the UFSAR 2020 Page 6.2-26 of 49

Containment Spray System FSAR Section 6.4 The spray system is designed to operate over an extended time period, following a primary coolant system failure. It has the capability of reducing the containment post accident pressure and consequent containment leakage taking into account any reduction due to single failures of active components.

Portions of other systems which share functions and become part of the containment cooling system, when required, are designed to meet the criteria of this section. Any single failure of active components in such systems does not degrade the heat removal capability of containment cooling.

Those portions of the spray systems located outside of the containment which are designed to circulate radioactively contaminated water collected in the containment, under post accident conditions, meet the following requirements:

1. Adequate shielding to maintain radiation levels within the limits of 10 CFR 50.67 (Section 11.6).

2. Collection of discharges from pressure relieving devices into closed systems.

3. Means to limit radioactivity leakage to the environs, to maintain radiation dose within the limits set forth in 10 CFR 50.67.

Recirculation loop leakage is discussed in Section 6.2.3.

System active components are redundant. System piping located within the containment is redundant and separable in arrangement unless fully protected from damage which may follow any primary coolant system failure. System isolation valves relied upon to operate for containment cooling are redundant, with automatic actuation or manual actuation.

Effects of Debris in Response to GL 2004-02 The design basis for the containment spray system with regard to the effects of accident-generated debris on the performance of containment sump strainers during post-accident recirculation is a risk-informed analysis, which shows the risk for failures caused by accident-generated debris is very small as defined by Regulatory Guide 1.174. The conclusion is based on plant-specific testing and analyses using inputs and assumptions that provide safety margin and defense-in-depth. See a summary of the risk-informed analysis in FSAR Appendix A.

Service Life All portions of the system located within containment are designed to withstand, without loss of functional performance, the post accident containment environment and operate without benefit of maintenance for the duration of time to restore and maintain containment conditions at near atmospheric pressure.

Codes and Classifications Table 6.4-1 tabulates the codes and standards to which the containment spray system components are designed.

6.4.2 SYSTEM DESIGN AND OPERATION

System Description

UFSAR 2015 Page 6.4-4 of 25

Containment Spray System FSAR Section 6.4 Normal and emergency power supply requirements are discussed in Section 8.0.

Shared Function Evaluation Table 6.4-8 is an evaluation of the main components which have been discussed previously and a brief description of how each component functions during normal operation and during the accident.

Containment Spray Pump NPSH Requirements The Net Positive Suction Head (NPSH) for the containment spray pumps was evaluated for both the injection and recirculation phases of operation (Reference 17).

During the injection phase the spray pump takes suction from the RWST. Available NPSH is dependant on the RWST level, RWST temperature, and the number of systems taking water from the RWST. Plant operating procedures ensure adequate water levels are maintained in the RWST such that spray pump NPSH requirements are satisfied.

During the recirculation phase the spray pump takes suction from the RHR system (the discharge of the RHR heat exchangers) which takes suction from the containment sump. The adequacy of RHR pump NPSH available was evaluated as described in Appendix A. There is adequate NPSH for the Containment Spray pumps during recirculation spray operation, with both the RHR and the Containment Spray pumps injecting, as long as containment spray is aligned through the reduced flow path and the SI-852 valves are throttled to the intermediate position. The available NPSH is adequate without crediting containment sump suction pressure in excess of normal atmospheric pressure.

6.4.4 REQUIRED PROCEDURES AND TESTS Inspection Capability All components of the containment spray system can be inspected periodically to demonstrate system readiness. The pressure containing systems are inspected for leaks from pump seals, valve packing, flanged joints and safety valves during system testing. During the operational testing of the containment spray pumps, the portions of the system subjected to pump pressure are inspected for leaks.

Component Testing All active components in the containment spray system are tested both in pre-operational performance tests in the manufacturer's shop and in-place testing after installation. The containment spray pumps can be tested singly using the full flow recirculation line. Each pump in turn can be started by operator action and checked for flow establishment. The spray injection valves can be tested with the pumps shut down.

The spray additive tank valves can be opened periodically for testing. The contents of the tank are periodically sampled to determine that the proper solution is present.

The containment spray nozzle availability is tested by blowing smoke or a gas mixture through the nozzles and observing the flow through the various nozzles in the containment visually or by telltales.

UFSAR 2015 Page 6.4-12 of 25

Safety Analysis FSAR Section 14.0 14.0 SAFETY ANALYSIS This section evaluates the safety aspects of either Unit 1 or Unit 2 of the plant, demonstrates that either or both units can be operated safely and that exposures from credible accidents do not exceed the guidelines of 10 CFR 50.67 or other applicable acceptance criteria.

This section is divided into three subsections, each dealing with a different behavior category:

Core and Coolant Boundary Protection Analysis, FSAR 14.1 With the exception of the Locked Rotor Accident, the abnormalities presented in FSAR 14.1 have no off-site radiation consequences. Radiological consequences, resulting from fuel cladding damage and a radioactivity release to the outside atmosphere, are assumed to occur as a result of the Locked Rotor Accident, presented in FSAR 14.1.8.

Standby Safety Features Analysis, FSAR 14.2 With the exception of the Locked Rotor Accident, the accidents presented in FSAR 14.2 are more severe than those discussed in FSAR 14.1 and may cause release of radioactive material to the environment.

Rupture of a Reactor Coolant Pipe, FSAR 14.3 The accident presented in FSAR 14.3, the rupture of a reactor coolant pipe, is the worst case accident and is the primary basis for the design of engineered safety features. It is shown that even the consequences of this accident are within the guidelines of 10 CFR 50.67.

Long term cooling following a rupture of a reactor coolant pipe may require operating the safety injection system and CSS in the recirculation mode. The impact of the debris generated by the rupture on the recirculation function was analyzed in a risk-informed evaluation in response to GL 2004-02. The evaluation provides confidence that the containment recirculation sump strainer design supports long-term core cooling following an accident. The evaluation meets the acceptance guidelines for a very small risk impact as defined in Regulatory Guide (RG) 1.174.

Additional details are provided in Appendix A.

Parameters and assumptions that are common to various accident analyses are described below to avoid repetition in subsequent sections.

Steady State Errors For most accidents which are DNB limited, nominal values of initial conditions are assumed.

The allowances on power, temperature, and pressure are determined on a statistical basis and are included in the limit DNBR, as described in WCAP-11397 (Reference 1). This procedure is known as the Revised Thermal Design Procedure, and is discussed more fully in FSAR 3.2.

For accidents in which the Revised Thermal Design Procedure is not employed, the initial conditions are obtained by adding the maximum steady state errors to rated values. The following conservative steady state errors were assumed in the analyses:

1. Core Power +/- 0.6% allowance for calorimetric error

2. Average Reactor Coolant Temp +/- 6.4qF allowance for controller deadband and measurement error

3. Pressurizer Pressure +/- 50 psi allowance for steady state fluctuations and measurement error UFSAR 2020 Page 14.0-1 of 16

PBNP-FSAR-A.8 APPENDIX A.8 Resolution of NRC Generic Letter 2004-02 A.8.1 Introduction and Risk-Informed Approach Summary NRC Generic Letter (GL) 2004-02 (Reference 1) required licensees to perform an evaluation of the emergency core cooling system (ECCS) and containment spray system (CSS) recirculation functions, and the flow paths necessary to support those functions, based on the potential susceptibility of sump strainers to debris blockage during design basis accidents requiring recirculation operation of ECCS or CSS. This generic letter resulted from Generic Safety Issue (GSI) 191.

The plant licensing basis considers long-term core cooling (LTCC) following a loss of coolant accident (LOCA) as identified in 10 CFR 50.46(b)(5). For PBNP, long-term cooling is supported by the safety injection (SI) system, which constitutes the ECCS. This system and the CSS are subject to the effects of accident-generated debris because they rely on the containment sump in the recirculation mode. Debris generated from non-LOCA initiating events are also considered. The risk-informed evaluation analyzes the following events:

1. Large, medium, and small LOCAs due to:

i. Pipe breaks ii. Failure of non-piping components iii. Water hammer

2. Secondary side breaks inside containment that result in a consequential LOCA upon failure to terminate safety injection or a stuck open PORV requiring sump recirculation

3. Seismically-induced LOCAs A risk-informed evaluation was performed to respond to GL 2004-02. The evaluation provides confidence that the sump design supports LTCC following a LOCA. The evaluation meets the acceptance guidelines for a very small risk impact as defined in Regulatory Guide (RG) 1.174 (Reference 2).

The licensing basis with regard to effects of debris is determination of a high probability that the ECCS and CSS can perform their design basis functions based on PBNP-specific testing and analyses using an NRC-approved methodology. The risk from breaks that could generate debris and result in strainer or reactor failures is very small and is, therefore, acceptable in accordance with the RG 1.174 guidelines (Reference 2).

A.8-1

PBNP-FSAR-A.8 The use of a risk-informed method, rather than the deterministic methods prescribed in the regulation, required an exemption to 10 CFR 50.46(a)(1), which has been granted pursuant to 10 CFR 50.12 (Reference 22).

The risk-informed approach identifies scenarios that fail any one of the following acceptance criteria, as determined by break-specific analysis:

1. Strainer debris limits (based on PBNP-specific head loss testing and analysis),

2. Strainer structural differential pressure limit,

3. Partially-submerged strainer head loss limit,

4. Pump void fraction limit,

5. Flashing limit

6. Pump net positive suction head (NPSH) margin The following failure scenarios were evaluated using bounding analyses, rather than break-specific analyses:

1. In-vessel downstream effects (core blockage limits)

2. Ex-vessel downstream effects (wear and blockage of recirculation flow path)

The results of both the break-specific analysis and the bounding evaluations are used to calculate conditional failure probabilities (CFPs). The resulting CFPs are used, along with inputs from the PBNP probabilistic risk assessment (PRA) model, to calculate the change in core damage frequency (CDF) and change in large early release frequency (LERF) due to failures caused by accident-generated debris. The CDF and LERF values, along with the PBNP baseline CDF and LERF from the PBNP PRA models, are compared to the guidelines in RG 1.174 (Reference 2). The results of the evaluation show that the risk due to strainer and reactor failures caused by accident-generated debris is "very small" (i.e., in Region Ill of RG 1.174). The methodology includes conservatisms in the plant-specific testing and in the assumption that all unbounded breaks result in loss of core cooling (Reference 12).

Key aspects of the risk-informed evaluation include (Reference 12):

1. The methodology used to quantify the amount of debris generated at each break location, including the assumed zone of influence (ZOI) sizes based on the target destruction pressure and break size, and the assumed ZOI shape (spherical or hemispherical) based on whether the break is a double-ended guillotine break (DEGB) or partial break.

2. The methodology used to evaluate debris transport to the strainers.

3. The methodology used to quantify chemical precipitates, including the refinements to WCAP-16530-P-A (Reference 5), application of the solubility correlation, and application of the WCAP-17788-P autoclave testing (Reference 6).

4. The strainer debris limits which are based on tested and analyzed debris quantities.

5. The methodology and acceptance criteria used to assess ex-vessel component blockage and wear.

6. The methodology used to assess in-vessel downstream effects and the associated limits.

A.8-2

PBNP-FSAR-A.8

7. The methodology used to quantify conditional failure probabilities, and CDF and LERF.

A.8.2 Debris Generation Post-accident debris includes failed insulation and coatings within the ZOI of the pipe break, as well as latent debris, unqualified coatings, and miscellaneous debris in containment. To support the debris generation evaluation, containment walkdowns were performed.

The pipe break characterization followed the methodology of NEI 04-07 (Reference 3) and associated NRC safety evaluation (SE) (Reference 4), with the exception that it characterized a full range of breaks rather than just the worst-case breaks as suggested by NEI 04-07. DEGBs and partial breaks on every Class 1 in-service inspection (ISI) weld at locations inside the first isolation valve were evaluated. The breaks were divided into three high-level categories: small-break LOCAs (SBLOCAs) - breaks smaller than 2 inches, medium-break LOCAs (MBLOCAs)

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

In the debris generation calculations (one for each unit, References 13 and 14), a three-dimensional CAD model of the Unit 1 and Unit 2 containment buildings was used to model the ZOI for each postulated break. ZOIs representing possible breaks on the reactor coolant system (RCS) piping were modeled at each ISI weld.

DEGBs are modeled using a spherical ZOI with a radius proportional to the pipe inner diameter.

Partial breaks are any breaks smaller than a DEGB and are modeled using a hemispherical ZOI with a radius proportional to the equivalent break size. Break sizes ranging from 1/2 inch up to a DEGB were modeled at each weld. In addition, because the orientation of partial breaks can have an effect on the results, partial breaks were modeled every 45 degrees around the circumference of the pipe at each weld. While DEGBs on main loop piping are typically bounding with regard to the volume of debris generated, smaller breaks are more likely to occur.

Since different material types have different destruction pressures, a size of ZOI was determined for each type of material. The quantity of generated debris for each break case was calculated using these material-specific ZOI sizes.

Unqualified coatings considered in the analysis may include coatings within containment that do not have a specified preparation, application, or inspection compliant with plant specifications, 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: epoxy, inorganic zinc, and alkyd coatings. Unqualified coatings were conservatively assumed to fail at the start of sump recirculation for all postulated breaks.

Additionally, actively delaminating qualified (ADQ) epoxy coatings are also considered as a debris source, which fail as a combination of particulate, flat fine chips, flat small chips, flat large chips, and curled large chips.

The total amount of latent debris was determined from containment walkdown. A larger quantity of 150 lbm is assumed in the strainer evaluation to bound the latent debris quantities for A.8-3

PBNP-FSAR-A.8 both units with ample operating margin. Per the guidance in NEI 04- 07 Volume 2, latent debris is assumed to consist of 15 percent fiber and 85 percent particulate by mass (Reference 4).

Similarly, the quantity of miscellaneous material, such as labels, tags, stickers, and placards was determined from containment walkdown for both units. A larger quantity of 200 ft2 of miscellaneous debris is assumed in the strainer evaluation to bound both units. Per the guidance in NEI 04-07 (Reference 3) and the SE (Reference 4), the total surface area of miscellaneous debris was assumed to block an equivalent surface area of the sump strainers after allowance for 25% overlap.

Although the probability of occurrence is low, a secondary side break inside containment could require ECCS recirculation. No detailed evaluation is performed to quantify the debris that could be generated and transported for the secondary side breaks. Instead, a bounding evaluation was performed to determine the risk contribution of the secondary side breaks by assuming that all secondary side breaks resulting in a consequential LOCA would fail the strainers due to effects of debris.

A.8.3 Debris Transport to the Sump Strainers The debris transport analysis (one for each unit, References 15 and 16) determines the fraction of each type and size of debris that could be transported to the sump strainers. The evaluation considers debris transport during the blowdown, washdown, pool fill, and recirculation phases based on plant-specific layout and flow conditions. For the recirculation phase, computational fluid dynamic (CFD) modeling was used to determine the sump pool flow conditions and transport of debris inside the pool for different break locations and pump lineups. For scenarios with both strainers in operation, debris accumulation on the two strainers is assumed to be proportional to the flow split across the strainers.

Potential upstream blockage points in containment were reviewed for both units.

Specifically, various lower containment compartments and the refueling canal drains were qualitatively evaluated and it was concluded that blockage would not occur at these locations.

A.8.4 Chemical Effects The post-LOCA sump strainer chemical effects analysis (Reference 17) methodology includes:

x Quantification of chemical precipitates using the WCAP-16530-NP-A (Reference 5) base methodology.

x Introduction of pre-prepared precipitates in strainer head loss testing.

x Application of an aluminum solubility correlation to determine the maximum precipitation temperature for strainer head loss related evaluations.

For each postulated break, the amount of chemical precipitate was determined using the WCAP-16530-NP-A (Reference 5) methodology using the break specific quantities of LOCA generated debris. The amount of chemical precipitate was maximized by applying conservative plant-specific inputs, such as pH, temperature, and aluminum quantity.

A.8-4

PBNP-FSAR-A.8 The generated amounts of chemical precipitates were used as inputs for strainer head loss testing to determine its impact on the head loss across the strainer. Sodium aluminum silicate used during testing was prepared according to the WCAP-16530-NP-A (Reference

5) recipe and settling test criteria.

An aluminum solubility correlation, developed by Argonne National Laboratory, was used to determine the conditions at which aluminum will precipitate, which effectively delays the onset of aluminum precipitation when analyzing strainer head loss. Precipitation is assumed to occur for all postulated breaks, regardless of the calculated maximum precipitate formation temperature. Functionally, precipitation is forced by assuming a maximum time to precipitation if the sump temperature remains above the calculated precipitation temperature.

A.8.5 Sump Strainer Evaluations There are several failure criteria considered in the sump strainer evaluation: tested debris limits, strainer partial submergence, vortexing, void fraction, flashing, pump NPSH, and strainer structural margin. A postulated break that fails one or more of these criteria for the strainer or RHR pumps is treated as a failure of the corresponding ECCS train(s). Each of the failure criteria was evaluated during the LOCA transient to determine if a strainer or RHR pump failure would occur (Reference 12).

The containment sump pool water volume following a LOCA was determined by considering all water sources (e.g., the refueling water storage tank, reactor coolant system, spray additive tank and accumulators) and subtracting the various holdup volumes. The holdup volumes include containment surface hold-up, refueling cavity, pressurizer cubicle, filling of empty pipe, water in transit, and steam holdup. The sump pool volume was used to determine the pool water level using a correlation between pool water depth and volume (Reference 12).

Strainer Head Loss Test Limits Strainer head loss tests were performed to measure the head losses of the 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 PBNP. Different test cases were performed with the thin bed and full debris load protocols, following the 2008 NRC staff review guidance (Reference 7).

The results of the four head loss tests provided a matrix of head loss data for various combinations of conventional and chemical debris loads. This matrix was used to derive debris head loss lookup tables for conventional and chemical debris. A rule-based approach was applied to the lookup table to determine the debris head loss for each break scenario based on its debris loads.

A sump strainer is conservatively assumed to fail for any break where the transported quantity of fiber, particulate, and/or chemical precipitate debris exceeds the tested debris quantities after being scaled from the test strainer surface area to the plant strainer surface area with adjustments for blockage by miscellaneous debris.

A.8-5

PBNP-FSAR-A.8 Partially Submerged Strainer Criteria This acceptance criterion was determined not to be applicable for PBNP Unit 1 or Unit 2 because the sump strainers are fully submerged at the start of sump recirculation for all breaks.

Strainer Vortexing Criteria In lieu of vortexing calculations, vortex testing was conducted during strainer head loss testing to identify under which conditions vortexing and air ingestion is expected to occur in the plant for both clean strainer and debris laden conditions. The vortex test used a prototype strainer module with a conservatively high approach velocity and low strainer submergence. Comparison of test results with plant conditions showed no vortex formation for the plant strainers up to the tested debris limits.

Void Fraction Criteria A pump failure due to degasification was assumed if the calculated steady-state gas void fraction at the pump suction is greater than 2 percent by volume. The quantity of air released from a given volume of water across the strainer was determined by calculating the difference between the concentration of air dissolved in the sump water and the concentration of air dissolved in water downstream of the strainer. The degasification evaluation is performed at mid-height of the strainer and the calculated void fraction is conservatively assumed to be the same as that at the pump suction. No containment accident pressure is credited for degasification evaluation.

Flashing Criteria A flashing failure is recorded if, at any time during sump recirculation, the pressure downstream of the strainer is lower than the vapor pressure at the sump temperature. The pressure downstream of the strainer is calculated based on the strainer submergence measured at the top of the strainer, containment pressure, and total strainer head loss. By crediting a small amount of containment accident pressure, flashing failures are reduced such that they had no impact to the overall risk quantification results.

Pump NPSH Criteria The RHR NPSH margin was calculated based on the NPSH available minus the NPSH required.

NPSH available was defined without considering strainer head loss. This was calculated as a time-dependent parameter for each postulated break based on containment pressure, sump temperature, water level, and vapor pressure. The NPSH required obtained from the pump curve was adjusted based on the time-dependent void fraction from the degasification evaluation, using the methodology in RG 1.82 (Reference 8).

Containment accident pressure is not credited in the analysis for pump NPSH. For containment pressure, the saturation pressure at the sump temperature was assumed for sump temperatures greater than 212qF. For sump temperatures below 212qF, the containment pressure of 14.7 psia was used to calculate the pump NPSH available.

A.8-6

PBNP-FSAR-A.8 Because NPSH available is calculated without considering strainer head loss, the acceptance criterion is that the total strainer head loss (i.e., clean screen head loss plus conventional debris head loss plus chemical debris head loss) must be less than the pump NPSH margin. This was evaluated on a time-dependent basis.

Because the SI pumps and CS pumps take suction from the RHR pumps during recirculation, only the NPSH margins of the RHR pumps are calculated.

Strainer Structural Criteria The head loss across each of the operating strainers was compared to the strainer structural margin to ensure that the structural margin is not exceeded. A failure was recorded if the total strainer head loss across the debris bed was greater than the strainer structural margin.

A.8.6 Downstream Effects - Components and Systems An analysis was performed to evaluate the impact of debris on the wear or blockage of the ECCS and CSS piping and components downstream of the strainer (excluding reactor vessel) following a LOCA (Reference 18). This ex-vessel downstream effects evaluation used the methodology presented in WCAP-16406-P-A (Reference 9). The analyzed effects of debris ingested through the containment sump strainers during the recirculation mode include erosive wear, abrasion, and potential blockage of downstream flow paths (Reference 18).

The smallest clearance for the PBNP heat exchangers, orifices, spray nozzles, and system piping in the recirculation flow paths is larger than the sump strainer hole size. Therefore, no blockage of the ECCS or CSS flow paths is expected.

ECCS and CSS instrumentation tubing was evaluated for potential debris accumulation.

Since the sensing lines are water solid and stagnant, settling is the only process by which the debris can be introduced into the instrument tubing. Plant walkdowns showed that all instruments tap into upper half of the process piping. This excludes the possibility of debris settling in the subjected instrument tubing. Therefore, debris effects will not cause an ECCS and CSS instrument failure.

The heat exchangers, orifices, and spray nozzles, and system piping were evaluated for the effects of erosive wear for a bounding debris concentration over the 30-day mission time. The erosive wear on these components was determined to be insufficient to affect the system performance.

The effects of debris ingestion were evaluated for three aspects of pump operability including hydraulic performance, mechanical shaft seal assembly performance, and the mechanical performance (vibration) of the pump. These performances were determined to not be negatively affected by the recirculating sump debris.

Valves on the ECCS and CSS flow paths were shown to pass the acceptance criteria for the blockage evaluation. The minimum recirculation flow rates are adequate to preclude debris sedimentation within the valves in all cases. Additionally, the valves that are subject to erosion pass the acceptable criteria for the 30-day mission time.

A.8-7

PBNP-FSAR-A.8 A.8.7 Downstream Effects - Fuel and Vessel During the post-LOCA sump recirculation phase, debris that passes through the sump strainers could accumulate at the reactor core inlet or inside the reactor vessel, potentially challenging LTCC. In-vessel downstream effects were analyzed following the NRC review guidance (Reference 11) and WCAP-17788-P (Reference 10). PBNP-specific penetration test data was used to quantify fiber bypass through the strainers (Reference 19).

Testing was conducted to collect time-dependent fiber penetration data using a test strainer and conditions that are prototypical to PBNP. A model was derived from the test results and was used to quantify fiber penetration for the sump strainer at plant conditions (Reference 20). This model defines the time dependent downstream debris source term used to calculate the fiber accumulation in the reactor vessel.

Methods and acceptance criteria contained in WCAP-17788-P, Revision 1 (Reference 10) were used to evaluate the accumulation of fiber inside the reactor vessel for a hot leg break (HLB)

(Reference 12), in accordance with the NRC staff review guidance for in-vessel effects (Reference 11). The quantity of fiber accumulation inside the reactor vessel was calculated for a HLB using a fiber debris quantity that bounds all breaks of both units. Additionally, various combinations of inputs (e.g., pump lineup and flow rate, CSS operation, sump pool volume) were analyzed to ensure the maximum in-vessel fiber load is calculated. The worst case in-vessel fiber load was shown to be less than the in core debris limit defined in WCAP-17788-P (Reference 10).

Therefore, no breaks fail the in-vessel debris limit.

A.8.8 Analyzed Debris Limits Containment accident generated and transported debris is defined as the quantity of debris calculated to arrive at the containment sump strainer. As described in the previous sections, the evaluation of the effects of debris includes strainer head loss, downstream ex-vessel effects, and downstream in-vessel effects. Of these three aspects of the evaluation, strainer head loss has the bounding debris limits (Reference 12).

Based on the tested and analyzed debris quantities, strainer debris limits were defined for all debris types (see Table A.8-1). These debris limits cannot be exceeded for breaks smaller than or equal to 12 inches for two RHR train operation, or breaks smaller than or equal to 8 inches for single RHR train operation (Reference 12). Larger breaks may exceed these debris limits without exceeding the RG 1.174 Region III acceptance guidelines (Reference 2).

If debris quantities greater than the analyzed debris limits are identified for the break sizes and operating configurations discussed above, the condition would be evaluated for operability and the applicable Technical Specifications (TS) action(s) would be entered if appropriate.

A.8-8

PBNP-FSAR-A.8 Table A.8-1: PBNP Sump Strainer Debris Limits Debris Type Debris Limit Unit Fiber Insulation 42.70 ft3 Mineral Wool 41.94 ft3 Cal-Sil and Asbestos Cal-Sil 384.44 lbm Coatings Particulate 12.29 ft3 ADQ Epoxy Fine and Small Chips 1.52 ft3 Latent Debris 150 lbm Miscellaneous Debris 200 ft3 A.8.9 References

1. NRC Generic Letter (GL) 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation during Design Basis Accidents for Pressurized-Water Reactors, September 13, 2004.

2. Regulatory Guide 1.174, An Approach for Using Probabilistic Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, Revision 3, January 2018.

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

4. NEI 04-07 Volume 2, Safety Evaluation by the Office of Nuclear Reactor Regulation Related to NRC Generic Letter 2004-02, Revision 0, December 2004.

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

6. WCAP-17788-P Volume 5, Comprehensive Analysis and Test Program for GSI-191 Closure (PA-SEE-1090) - Autoclave Chemical Effects Testing for GSI-191 Long- Term Cooling, Revision 1, December 2019.

7. NRC Staff Review Guidance Regarding Generic Letter 2004-02 Closure in the Area of Strainer Head Loss and Vortexing, March 2008 (ADAMS Accession No. ML080230038).

8. Regulatory Guide 1.82, Water Sources for Long Term Recirculation Cooling Following a Loss of Coolant Accident, Revision 3, November 2003.

9. WCAP-16406-P-A, Evaluation of Downstream Sump Debris Effects in Support of GSI-191, Revision 1, March 2008.

10. WCAP-17788-P Volume 1, Comprehensive Analysis and Test Program for GSI-191 Closure (PA-SEE-1090), Revision 1, December 2019.

11. U.S. Nuclear Regulatory Commission Staff Review Guidance for In- Vessel Downstream Effects Supporting Review of Generic Letter 2004-02 Responses, September 4, 2019 (ADAMS Accession No. ML19228A011).

12. NEE-591-CALC-002, Point Beach Units 1 and 2 Risk Quantification Calculation, Revision 0.

13. NEE-021-CALC-001, Point Beach Unit 1 Debris Generation Calculation, Revision 4.

14. NEE-021-CALC-002, Point Beach Unit 2 Debris Generation Calculation, Revision 4.

15. NEE-132-CALC-001, Point Beach Unit 1 Debris Transport Calculation, Revision 4.

16. NEE-132-CALC-003, Point Beach Unit 2 Debris Transport Calculation, Revision 4.

17. NEE-132-CALC-002, Point Beach Unit 1 and 2 GSI-191 Containment Sump Chemical Product Generation, Revision 2.

A.8-9

PBNP-FSAR-A.8

18. Calculation 51-5070148, Point Beach Unit 1 and Point Beach Unit 2 Downstream Effects Evaluation, Revision 4.

19. 1142PBNBYP-R2-01, Point Beach Large Scale Fibrous Debris Penetration Test Report, Revision 0.

20. 1142PBNBYP-600-00, Fibrous Debris Penetration Model for Point Beach Calculation, Revision 0.

21. NextEra Updated Final Response to GL 2004-02 for Point Beach (ADAMS Accession No.

TBD)

22. NRC SER on NextEra Updated Final Response to GL 2004-02 for Point Beach (ADAMS Accession No. TBD)

A.8-10

Point Beach Nuclear Plant Licensing Submittal for a Risk-Informed Resolution of Generic Letter 2004-02 Enclosure 3 Updated Final Responses to Generic Letter 2004-02 Table of Contents

1. Overall Compliance .............................................................................................. 4

2. General Description of and Schedule for Correction Actions ................................ 5

3. Specific Information Regarding Methodology for Demonstrating Compliance ...... 9

a. Break Selection .......................................................................................... 9

b. Debris Generation/Zone of Influence (excluding coatings) ...................... 14

c. Debris Characteristics .............................................................................. 23

d. Latent Debris ........................................................................................... 27

e. Debris Transport ...................................................................................... 30

f. Head Loss and Vortexing......................................................................... 69

g. Net Positive Suction Head ..................................................................... 105

h. Coatings Evaluation ............................................................................... 119

i. Debris Source Term ............................................................................... 125

j. Screen Modification Package ................................................................ 129

k. Sump Structural Analysis ....................................................................... 134

l. Upstream Effects ................................................................................... 147

m. Downstream Effects - Components and Systems ................................. 151

n. Downstream Effects - Fuel and Vessel ................................................. 156

o. Chemical Effects .................................................................................... 171

p. Licensing Basis ...................................................................................... 185

4. NRC Audit Report Responses .......................................................................... 186

5. References ....................................................................................................... 192 E3-1

Enclosure 3 Updated Final Responses to Generic Letter 2004-02 NextEra Energy Point Beach LLC (NextEra) has chosen a risk-informed approach for Point Beach Nuclear Plants (PBNs) responses to Generic Letter (GL) 2004-02 (Reference 1) to address the effects of loss-of-coolant accident (LOCA)-generated debris on the emergency core cooling system (ECCS) and containment spray system (CSS) recirculation functions. This enclosure summarizes various testing programs and analyses that supported the responses. It provides an update to PBNs final response to GL 2004-02. This enclosure follows the content guide provided by the NRC (Reference 2; 3; 4; 5) and addresses all topical areas in those documents. As a result, it supersedes all previous GL 2004-02 submittals for PBN. The text from the NRC guidance is presented in italic script. Questions raised by the Nuclear Regulatory Commission (NRC) staff during their audit of the previous submittal (Reference 6; 7) are addressed and dispositioned in this enclosure (see Section 4).

NRC Request, Summary-Level Description The GL supplemental response should begin with a summary-level description of the approach chosen. This summary should identify key aspects of design modifications, process changes, and supporting analyses that the licensee believes are relevant or important to the NRC staffs verification that corrective actions to address the GL are adequate. The summary should address significant conservatisms and margins that are used to provide high confidence the issue has been addressed even with uncertainties remaining. Licensees should address commitments and/or descriptions of plant programs that support conclusions.

Summary-Level Description for PBN The key aspects of the approach chosen by NextEra to resolve the concerns identified in GL 2004-02 are stated below for clarity:

x Extensive design modifications to significantly reduce the potential effects of post-accident debris and latent material on the functions of the ECCS and CSS during the recirculation phase of accident mitigation.

x Extensive testing and analysis to determine break locations, identify and quantify debris sources, quantify debris transport, determine upstream and downstream effects, and confirm the recirculation function.

x Extensive changes to plant programs, processes, and procedures to limit the introduction of materials into containment that could adversely impact the recirculation function, and development of monitoring programs to ensure containment conditions will continue to support the recirculation function.

x Application of conservative measures to assure adequate margins throughout the actions taken to address the GL 2004-02 concerns.

More details are provided below for the plant-specific analyses, changes to the licensing basis, improvements in processes and programs, and conservatisms and margins.

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02 Analyses An extensive debris generation analysis has been performed for PBN, which determined the debris generated for all break sizes from 0.5 inches up to 31 inches at all Class 1 in-service inspection (ISI) welds at locations inside the first isolation valve where reactor coolant system (RCS) pressure is expected to be present. The locations were analyzed as double ended guillotine breaks (DEGBs), single ended guillotine breaks (SEGBs) (where a closed valve is within 10 pipe diameters), and partial breaks at 45 degree intervals around the circumference of the pipe. This debris generation analysis was an automated evaluation based on a detailed computer-aided design (CAD) model of containment. Additional discussion of the debris generation analysis is provided in the Response to Sections 3.a through 3.c.

There were no reductions in the zone of influence (ZOI) sizes from the accepted values in Nuclear Energy Institute (NEI) Report 04-07 (Reference 8 p. 30) for any materials except qualified coatings and mineral wool. Note that the ZOI size (4.0D) for qualified coatings was based on testing that has been reviewed and accepted by the NRC.

Since the previous submittal, the ZOI size for mineral wool is increased to 5.4D.

Additional discussion is provided in the Responses to Sections 3.b and 3.h.

NextEra has performed extensive testing for strainer head loss and fiber debris penetration. The PBN strainer head loss and fiber penetration testing was performed at Alden Research Laboratory, Inc (Alden) using protocols that have been reviewed and observed by the NRC staff. Additional discussion is provided in the Responses to Sections 3.f, 3.n, and 3.o.

Since the previous submittal in 2017 (Reference 7), NextEra has performed a new in-vessel downstream effects evaluation for PBN to demonstrate that accumulation of debris within the reactor core will not challenge long term core cooling following the latest NRC review guidance on in-vessel effects (Reference 9). The evaluation used the methodology and acceptance limit from Revision 1 of WCAP-17788 (Reference 10), as directed by the NRC guide. Refer to the Response to Section 3.n for details.

NextEra has elected to use the risk-informed methodology to address the effects of LOCA-generated debris on ECCS and CSS recirculation functions for PBN. The risk-informed analysis covers a full spectrum of postulated LOCAs, including partial breaks of various sizes and DEGBs. PBN conservatively relegates to failure the individual breaks that can generate and transport debris that are not bounded by PBN analyzed limits. The results of the risk quantification in Enclosure 4 show that the risk from the failures related to LOCA-generated debris is very small as the risk falls in Region III of RG 1.174 for PBN Units 1 and 2 (PBN1 and PBN2). The methodology includes conservatisms in the plant-specific testing and analyses, as summarized in Enclosure 5.

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02 Changes to the Licensing Basis The PBN FSAR was updated in 2007 to reflect the containment sump recirculation strainer perforation size for the replacement strainers. In this submittal, PBN uses a risk-informed approach to respond to GL 2004-02, which replaces the current deterministic methodology in the current PBN licensing basis. Adopting this change requires an exemption from certain requirements of 10 CFR 50.46(a)(1) and an amendment to the PBN Unit 1 and Unit 2 operating licenses. Refer to the Response to Section 3.p.

Improvements in Processes and Programs NextEra has completed a review of PBN procedures, processes, and programs and has updated those procedures and design specifications or standards that will ensure the analysis inputs and assumptions can be maintained. This is discussed in the response to Section 3.i.

Conservatisms and Safety Margins NextEra applied conservative measures to assure adequate safety margins throughout the actions taken to address the GL 2004-02 concerns. The key areas in which these conservative measures were applied are discussed in Enclosure 5.

1. 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:

NextEra has completed all necessary testing and analyses for PBN to implement the risk-informed approach to respond to GL 2004-02. The testing and analyses informed the risk-informed analysis to quantify the risk increase due to strainer and reactor core failures caused by LOCA-generated debris. The results show that the risk increase is very small.

More detailed summary of the testing programs and analyses is presented in Section 3 of this enclosure. The risk quantification is summarized in Enclosure 4 of this submittal.

Replacing the existing deterministic approach described in the PBN licensing basis with a risk-informed method requires an amendment to the PBN Units 1 and 2 operating licenses to incorporate the revised methodology per the requirements of Title 10 of the E3-4

Enclosure 3 Updated Final Responses to Generic Letter 2004-02 Code of Federal Regulations (CFR) Section 50.59 (10 CFR 50.59). The proposed amendment to the operating license is described in Enclosure 2 of this submittal.

Additionally, an exemption from certain requirement of 10 CFR 50.46(a)(1) is required due to this change in methodology. The request for exemption is provided in Enclosure 1 of this submittal.

NextEra has previously completed several plant modifications for PBN1 and PBN2, and no additional modifications are required for either unit. This submittal, as well as all PBN testing and analyses, addresses the current plant configuration.

2. 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:

The corrective actions to address the concerns identified in GL 2004-02 at PBN consisted of plant modifications, testing and analyses, and changes to plant programs and processes. These actions have been completed in accordance with NextEra regulatory commitments and NRC-approved extensions.

Plant Modifications The list below summarizes the modifications that have been implemented at PBN for addressing the GL 2004-02 concerns. No additional modifications are proposed x At PBN1 and PBN2 the original sump screens have been removed and replaced with new strainer systems. The new strainers have a much larger surface area than the original sump screens.

x The mineral wool insulation on the sides of the pressurizers for both units has been replaced with reflective metallic insulation (RMI).

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02 x The fibrous insulation on both reactor coolant pumps (RCPs) in PBN1and PBN2 has been replaced with RMI.

x Portions of the fibrous insulation on the Unit 2 main RCS loop piping have been replaced with RMI. Note that the insulation on Unit 1 main RCS loop piping is RMI.

x A 16-inch diameter opening has been bored to connect the normal containment operating sump with the accident sump on each unit. This ensures that in the event of a break at a reactor vessel nozzle there will be an adequate flow path (that will not be blocked by debris) for break flow to return to the strainers without holding up volume in the instrumentation keyway tower.

x The refueling cavity drain lines in each unit have been relocated to prevent direct impingement on, and ingestion of air into, one of the strainer trains. The drain lines are extended to move the discharge point away from the sump strainers to minimize turbulence caused by its discharge near the sump strainers and to prevent debris from depositing near the strainers.

x A hard cover under Stairway #22 in Unit 1 has been installed to protect Strainer A, which is located below the stairway, from debris washed down by containment spray.

The cover is extended to the entire length of the staircase to the bottom of the floor at 8 Elevation.

x Debris interceptors on the 8 Elevation, 10 Elevation, and 66 Elevation were installed in Unit 1 in 2008. The original design function of these interceptors was to capture and retain transportable debris. However, this intent was subsequently found to not be viable / defendable, and they are no longer credited for debris holdup. Refer to the Response to Section 3.e.4 for the modeling of the interceptors in the debris transport analysis. Since 2010, several debris interceptors have been removed. Debris transport analysis has shown that making these modifications has minimal impact on the resolution of GSI-191.

Testing and Analyses Most of the testing and analyses needed to implement the risk-informed responses to GL 2004-02 were completed prior to 2017. Since the previous submittal in 2017, NextEra has revised the PBN debris generation calculations to incorporate an increased ZOI size for mineral wool. The debris transport calculations were revised to correct the pool fill-up transport fractions to the reactor cavity for the reactor nozzle breaks. The chemical effects calculation was revised to incorporate the updated mineral wool debris loads, and containment and sump temperature profiles. Section 3 of this enclosure provides a summary of the relevant testing programs and analyses. A new risk quantification is performed to determine the risk increase due to strainer and reactor core failures caused by LOCA-generated debris. The risk quantification and associated uncertainty analysis are summarized in Enclosure 4 of this submittal.

Plant Programs and Processes Significant program and process changes necessary to address the GL 2004-02 concerns were completed by December 31, 2007.

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02 x Procedural controls are in place to reduce and control the amount of loose debris and fibrous material in containment. Procedures require inspection of all accessible areas to verify that no loose debris, fibrous material that could degrade into loose debris, or bubbling/chipping paint is present prior to setting containment integrity. Any entry performed while containment integrity is set requires subsequent walkdowns of areas affected by the entry to confirm no loose debris or foreign material was left in containment.

x The maintenance director is in charge of maintaining the general housekeeping of containment, which includes tracking the overall cleanliness of containment and promptly correcting identified deficiencies.

x Foreign material exclusion programmatic controls are in place, which ensure that proper work control is specified for debris-generating activities within the containment building. This assists in preventing introduction of foreign material into containment, which could potentially challenge the containment recirculation function. Additionally, the foreign material exclusion program requires that engineering be consulted any time foreign material covers are placed on, or modifications are performed on, the containment sump strainers. Lastly, the containment entry procedure provides additional controls to evaluate foreign materials to be brought into containment and ensure they are removed during at power entries.

x PBN engineering change processes and procedures ensure modifications that may affect the ECCS, including sump performance, are evaluated for GL 2004-02 compliance. During engineering change preparation, the process requires specific critical attributes be listed, evaluated, and documented when affected.

This includes the introduction of materials into containment that could affect sump performance or lead to equipment degradation. It also includes repair, replacement, or installation of coatings inside containment, including installing coated equipment.

x PBN has adopted the industrys standard design change process. The standard process and tools are intended to facilitate sharing of information, solutions and design changes throughout the industry. This process requires activities that affect UFSAR described structure, system, or component (SSC) design functions to be evaluated as a design change in accordance with NextEras 10 CFR 50 Appendix B program. This includes modifications that would impact the containment sump. Design changes require a final impact review meeting (i.e.,

final design workshop) and assessment in accordance with 10 CFR 50.59.

Additional meetings may be required based on the complexity and risk of the change. A failure modes and effects analysis is required if the design change introduces any new failure modes or changes failure modes for the affected SSCs.

x This guidance has been enhanced by an engineering specification that brings together, in one document, the insulation design documents that determine the design basis for the insulation debris component of the containment recirculation strainer design. This specification provides guidance for evaluating and maintaining piping and component insulation configuration within the containment buildings at PBN1 and PBN2.

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02 x Temporary configuration changes are controlled by plant procedure, which maintain configuration control for non-permanent changes to plant structures, systems, and components while ensuring the applicable technical and administrative reviews and approvals are obtained.

x In accordance with 10 CFR 50.65 (Maintenance Rule), an assessment of risk resulting from the performance of maintenance activities is required. Prior to performing maintenance, PBN assesses and manages the increase in risk that may result from the proposed maintenance activities. In general, the risk assessment ensures that the maintenance activity will not adversely impact a dedicated/protected train, which ensures a system is capable of performing its intended safety function.

Licensing Basis In this submittal, PBN replaced the deterministic methodology for analyzing debris effects on post-accident long term cooling with a risk-informed approach. The licensing requirements to address this change is summarized in the Response to Section 3.p.

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02

3. 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 PBN1 and PBN2 debris generation calculations followed the methodology of NEI 04-07 (Reference 11) and associated NRC SE (Reference 8) with the exception that they analyzed a full range of breaks, not just the worst-case breaks as suggested by NEI 04-07. The purpose of the debris generation calculations was to obtain debris quantities for the full range of possible break scenarios. This method ensures that the most challenging break will be identified. The calculations evaluated debris generation quantities for breaks on every ISI weld identified within the Class 1 pressure boundary inside the first isolation valve, including breaks at the reactor nozzles. The following types of loss of coolant accident (LOCA) breaks were considered in the debris generation analysis:

x Double-ended guillotine breaks (DEGBs) with the largest break being a DEGB of the 31 cross-over leg, x Partial breaks, orientated 45° apart, at size increments of 0.5, 2, 4, 6, 8, 10, 12, 14, 17, 20, 23, and 26 inches x Single-ended guillotine breaks (SEGBs) within 10 pipe diameters of a normally closed isolation valve or termination point.

In the debris generation calculations, three-dimensional computer-aided design (CAD) models of the PBN1 and PBN2 containment buildings were updated to work with ENERCONs BADGER software. BADGER was used to place ZOIs representing possible breaks on every 1/2 or larger ISI weld identified in containment inside the first isolation valve. Figure 3.a.1-1 and Figure 3.a.1-2 show sketches with the approximate welding locations for PBN1 and PBN2, respectively.

Per Section 3.3.5.2 of the NRC SE of NEI 04-07, evaluating breaks at equal increments is only a reminder to be systematic and thorough. The use of Class 1 ISI welds as break locations is both systematic and thorough because they are closer to the components that contain the greatest quantity of debris sources as opposed to a span of straight pipe further way from these sources (see Figures 3.a.1-1 and 3.a.1-2). Also, welds are almost exclusively recognized as likely failure locations because they can have relatively high residual stress, are preferentially-attacked by many degradation mechanisms, and are most likely to have preexisting fabrication defects.

Since each of the weld locations were evaluated for determination of the quantity of debris that would be generated, the results provided comprehensive results for E3-9

Enclosure 3 Updated Final Responses to Generic Letter 2004-02 quantifying the risk increase due to strainer and reactor core failures caused by the debris generated by the breaks.

Figure 3.a.1-1: PBN1 Weld Locations Where Postulated LOCAs Occur E3-10

Enclosure 3 Updated Final Responses to Generic Letter 2004-02 Figure 3.a.1-2: PBN2 Weld Locations Where Postulated LOCAs Occur As stated above, at each weld, partial breaks at 8 different angles along the circumference of the weld were analyzed with an angular increment of 45°. This is consistent with the guidance in the NRC SE on NEI 04-07, which states that licensees will need to simulate various directions around the RCS main-loop piping to determine the limiting break location (Reference 8 p. 117). Figure 3.a.1-3 shows a graphical representation of the 8 ZOIs at different angles for a given weld. As shown in the figure, there is significant overlap between the ZOIs. Additionally, partial breaks were analyzed for various sizes: 0.5 in, 2 in, 4 in, 6 in, 8 in, 10 in, 12 in, 14 in, 17 in, 20 in, 23 in and 26 in. The potential for missing significant debris quantity spikes or trends by using these orientation and size increments was evaluated by running a BADGER E3-11

Enclosure 3 Updated Final Responses to Generic Letter 2004-02 sensitivity evaluation using the model for a 4-loop Westinghouse plant with low density fiberglass insulation (Reference 12).

The sensitivity evaluation considered 6-in and 8-in breaks at 100 different weld locations and refined the orientation increment from 45° to 15° and the size increment to 0.25 in. The comparison between the refined and original results showed that the maximum quantity of debris generated at the worst case orientation was approximately 2% higher on average with the 15° increment data compared to the 45° increment data. Additionally, reducing the break size increment did not result in any spikes or trends in the maximum debris loads. Therefore, refining the angle and/or break size increments will not impact the debris loads or the risk quantification results significantly.

Figure 3.a.1-3: Visualization of Partial Break ZOIs at One Weld with 45° Increment

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:

Feedwater and main steam piping were not considered for debris generation analysis.

However, the impact on the risk quantification by the secondary line breaks was accounted for in a bounding evaluation. See Section 3.2.4 in Enclosure 4 of this submittal.

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02

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:

The quantities of debris generated by a full range of breaks have been determined for PBN1 and PBN2 (see the Response to 3.a.1 and the Response to 3.b). These debris generation quantities were used to quantify the risk increase due to strainer and reactor core failures caused by LOCA-generated debris. The strainer failures related to strainer head loss were evaluated in a holistic, time-dependent manner against various failure criteria in the NARWHAL software. The evaluation showed that, generally, breaks that generate the most Cal-Sil and fine fiber debris present the greatest challenge to post-accident sump performance. Therefore, Table 3.a.3-1 shows the bounding breaks (i.e., highest fiber and Cal-Sil) for each unit. Note that the reactor core failure (i.e., in-vessel downstream effects) was analyzed in a bounding evaluation outside of NARWHAL and resulted in no failures (see the Response to 3.n in this enclosure).

Table 3.a.3-1: PBN1 and PBN2 Bounding Breaks Unit Loop Limiting Debris Type Weld Location PBN1 B DEGB with High Fiber Debris Load RC-36-MRCL-BII-01 PBN1 B DEGB with High Fiber Debris Load RC-34-MRCL-BI-03 PBN1 B DEGB with High Cal-Sil Debris Load RC-34-MRCL-BI-03 PBN1 B DEGB with High Cal-Sil Debris Load RC-34-MRCL-BI-02 PBN2 B DEGB with High Fiber Debris Load RC-36-MRCL-BII-01A PBN2 B DEGB with High Fiber Debris Load RC-36-MRCL-BII-01R1 PBN2 A DEGB with High Cal-Sil Debris Load RC-34-MRCL-AI-03 PBN2 A DEGB with High Cal-Sil Debris Load RC-34-MRCL-AI-04R1 E3-13

Enclosure 3 Updated Final Responses to Generic Letter 2004-02

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 ZOIs for generating debris.

Identify which debris analyses used approved methodology default values. For debris with ZOIs not defined in the guidance report/SE, or if using other than default values, discuss method(s) used to determine ZOI and the basis for each.

Response to 3.b.1:

In a pressurized water reactor (PWR) containment building, the worst-case pipe break is typically 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 8, p. vii) (Reference 11, p. 1-3) 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 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, coating, or other material impacted by the break jet.

For any break smaller than a DEGB (i.e., a partial break) NEI 04-07, Volume 2 accepts the use of a hemispherical ZOI centered at the edge of the pipe (Reference 8, p. 94).

Because these types of breaks can occur anywhere along the circumference of the pipe, the partial breaks were analyzed using hemispheres at eight different angles that are 45° apart from each other around the pipe.

Since different materials may have different destruction pressures, the ZOI sizes may vary with the types of materials. Table 3.b.1-1 shows the primary side break equivalent ZOI radii divided by the break diameter (L/D) for each representative material in the PBN1 and PBN2 containment buildings.

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02 Table 3.b.1-1: Primary Side Break ZOI Radii for PBN1 and PBN2 Insulation Types Destruction Pressure ZOI Radius/Break Diameter Insulation Type (psi) (L/D)

Nukon 6 17.0*

LDFG 6 17.0*

Temp-Mat 10.2 11.7*

Cal-Sil 20* 6.4*

Transco RMI 114 2.0*

Mirror RMI 2.4 28.6*

Asbestos Cal-Sil 20*** 6.4***

Qualified Coatings - 4.0**

Mineral Wool - 5.4****

• NRC SE for NEI 04-07 (Reference 8, p. 30 and II-20)

• • Revised Guidance Regarding Coatings Zone of Influence for Review of Final Licensee Responses to Generic Letter 2004-02 (Reference 14 p. 2)

• • • The destruction pressure of Asbestos Calcium-Silicate (PBN1 and PBN2) was assumed to be the same as Calcium-Silicate.

• • • • See the Response to 3.b.3 for the justification of the ZOI size for mineral wool.

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 ZOI 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 11, pp. 3-14 through 3-15). Due to the compartmentalization of containment in PBN1 and PBN2, the insulation on the opposite side of the compartment walls can be assumed to remain intact. All ZOIs were truncated to account for robust barriers per NEI 04-07 Volume 2 (Reference 8,

p. vii).

Volumetric debris quantities were determined by measuring the interference between a ZOI and its corresponding debris source. This was done within the CAD model environment.

No insulation debris would be generated outside of the ZOIs (Reference 11, pp. 3-19 through 3-20). This practice is considered acceptable by the NRC as stated in the SE for NEI 04-07 (Reference 8, Section 3.4.3.2).

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

Response to 3.b.2:

See the Response to 3.b.1.

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02

3. Identify if destruction testing was conducted to determine ZOIs. 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:

No destruction testing was conducted to determine ZOI sizes. PBN1 and PBN2 have applied the ZOI refinement discussed in NEI 04-07 Volume 2 (Reference 8, Section 4.2.2.1.1), which allows the use of debris-specific spherical ZOIs. The only ZOIs being used that are different from those listed in NEI 04-07 are those for mineral wool and qualified coatings.

The mineral wool at PBN was provided by Transco and is encapsulated in stainless steel cassettes. The insulation panels include a 20 gage stainless steel outer layer, a 26 gage stainless steel inner layer (in contact with the pipe), as well as 24 gage stainless steel end closures such that the mineral wool insulation for each panel is completely encapsulated. The insulation panels are fastened to the piping with a buckle fastener which includes a locking latch.

It is assumed that a ZOI size of 5.4D is applicable for the mineral wool insulation based on a comparison with the K-Wool insulation. K-Wool is classified as unjacketed mineral wool with wire mesh reinforcement of the blanket (Reference 11 Table 4-1).

The K-Wool insulation used during the air jet impact testing (AJIT) had a wire mesh lining and a silver coated heavy cloth fabric outer cover (Reference 15 p. 142). The blanket had stapled seams and was secured to the pipe using wire hooks and stainless-steel wire lacing (Reference 15 p. 142). The testing concluded a destruction pressure of 40 psi for K-Wool, which was reduced to 24 psi in the NRC SE of NEI 04-07, resulting in a ZOI size of 5.4D (Reference 8 Table 3-2). The mineral wool insulation at PBN has a more robust stainless steel cassette encapsulation. It is also assumed that 100% of the mineral wool within the 5.4D ZOI would be destroyed into fines that are more transportable. Thus, based on mineral wools similarity to K-Wool in material properties, more robust encapsulation of the PBN mineral wool insulation, and the conservative assumption that 100% of the mineral wool within the ZOI would be destroyed into fines, the 5.4D ZOI for mineral wool is conservative and acceptable.

A sensitivity BADGER run was performed to compare the mineral wool debris load obtained using the above assumption with that based on a 17D ZOI. The sensitivity run was performed on the DEGBs using the Unit 2 CAD model. When using the 17D ZOI, the Nukon size distribution as detailed in the Response to 3.c.1 was applied. The 17D ZOI results for fines also included 10% erosion of small and large piece debris (see the Response to 3.e.1). The comparison showed that utilizing a 5.4D ZOI for mineral wool while assuming 100% fines resulted in over three times more fine debris than use of a 17D ZOI with the corresponding size distribution. The sensitivity run was not performed for Unit 1 because of the similarities in the mineral wool insulation configurations between the two units, with the majority of the mineral wool insulation on the resistance temperature detector (RTD) lines and the bottom of the pressurizer.

E3-16

Enclosure 3 Updated Final Responses to Generic Letter 2004-02 This treatment of mineral wool is similar to that used in the North Anna GL 2004-02 submittal. The mineral wool insulation installed at North Anna is stainless steel jacketed with stainless steel bands on 18-inch centers. North Anna noted that the tested K-Wool insulation was only wire mesh lined (Reference 16 p. 14). In the NRCs audit of the North Anna submittal, the staff accepted the use of 5.4D ZOI for mineral wool based on a general qualitative similarity to K-Wool and noted that the assumption that 100% of the mineral wool within the 5.4D ZOI would be destroyed into fines would provide a conservative estimate of the transportable debris (Reference 16 p. 15). PBN has also provided a similar qualitative comparison of the PBN mineral wool cassettes and the tested K-Wool and utilizes the same assumption for destruction as 100% fine debris. The PBN mineral wool is also qualitatively similar to the North Anna mineral wool insulation.

Additionally, Surry assumed a 5.4D ZOI for mineral wool insulation in their GL 2004-02 response based on its similarity to K-Wool insulation (Reference 17). The NRC did not issue any requests for additional information (RAIs) on Surrys use of this ZOI size for mineral wool.

The ZOI size for qualified coatings is discussed in the Response to 3.h.

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 ZOIs listed in this section and the debris size distribution in the Response to 3.c of this enclosure, each of the breaks identified in the Response to 3.a was analyzed for the quantities of generated debris. Table 3.b.4-1 and Table 3.b.4-3 show the PBN1 and PBN2 most limiting DEGBs with respect to fiber and Cal-Sil debris, as determined in the PBN1 and PBN2 debris generation calculations. It should be noted that the 31 DEGB at RC-34-MRCL-BI-03 is the second highest fiber break and the highest Cal-Sil break for PBN1. Table 3.b.4-2 and Table 3.b.4-4 show the quantities of debris generated for the four most limiting breaks that do not result in any strainer failures.

Note that the volumes of qualified epoxy and IOZ coatings shown in the tables are converted from masses. For qualified IOZ, its density of 300 lbm/ft3 was used. PBN has several types of qualified epoxy coatings with slightly different densities. The minimum density of 97.1 lb/ft3 was conservatively used to convert the total mass of qualified epoxy coatings debris to volume. See the Response to 3.h.1 for the quantities of unqualified coatings and actively delaminating qualified (ADQ) epoxy coatings. See the Response to 3.d.3 for the quantity of latent debris.

All of the volumes shown in the tables for fiber debris are LDFG-equivalent volumes, which are calculated by dividing the actual fiber debris mass by the density of LDFG.

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Table 3.b.4-1: PBN1 Generated Debris Quantities for Worst-Case DEGBs Break Location RC-36-MRCL-BII-01 RC-34-MRCL-BI-03* RC-34-MRCL-BI-03* RC-34-MRCL-BI-02 Description Worst Fiber Break Worst Cal-Sil Break Break Size 31" 31" 31" 29" Break Type DEGB DEGB DEGB DEGB Fine 11.46 11.56 11.56 10.80 Small 44.19 44.70 44.70 40.90 LDFG (ft3)

Large 4.69 4.40 4.40 6.59 Intact 5.06 4.74 4.74 7.11 Mineral Wool Fine 126.11 120.07 120.07 101.71 (ft3)

Fine 0.00 0.00 0.00 0.00 Small 0.00 0.00 0.00 0.00 Temp-Mat (ft3)

Large 0.00 0.00 0.00 0.00 Intact 0.00 0.00 0.00 0.00 Cal-Sil and Fine 382.70 618.24 618.24 525.21 Asbestos Cal- Small 230.89 437.02 437.02 340.24 Sil (lbm) Intact 690.44 829.72 829.72 843.27 Small Mirror and 26876.15 27143.14 27143.14 26783.18

(<4)

Transco RMI Large (

(ft2) 8958.72 9047.71 9047.71 8927.73 4)

Qualified Epoxy Fine 1.05 0.95 0.95 0.97 (ft3)

Qualified IOZ Fine 0.29 0.34 0.34 0.23 (ft3)

• . Break RC-34-MRCL-BI-03 is the first worst Cal-Sil break and second worst fiber break.

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02 Table 3.b.4-2: The Four Worst-Case Breaks that Do Not Fail Any Acceptance Criteria for the Single Train Failure Equipment Configuration for PBN1 Break Location RC-36-MRCL-AII-02 RC-34-MRCL-AI-03 RC-36-MRCL-AIII-01 RC-36-MRCL-AIII-01A Description Worst Fiber Breaks Worst Cal-Sil Breaks Break Size 23" 23" 20" 17" Break Type Partial (225q) Partial (270q) Partial (0q) Partial (0q)

Fine 6.86 6.21 5.75 5.21 Small 26.16 23.84 21.82 19.21 LDFG (ft3)

Large 3.65 2.78 3.38 4.68 Intact 3.94 3.00 3.65 5.05 3

Mineral Wool (ft ) Fine 44.03 43.54 22.76 14.42 Fine 0.00 0.00 0.00 0.00 Small 0.00 0.00 0.00 0.00 Temp-Mat (ft3)

Large 0.00 0.00 0.00 0.00 Intact 0.00 0.00 0.00 0.00 Particulate 27.89 46.09 313.34 254.00 Cal-Sil and Asbestos Small 13.14 25.47 238.02 188.20 Cal-Sil (lbm)

Intact 111.26 93.61 346.75 302.28 Mirror and Transco Small (<4) 5022.03 9181.12 13973.99 9755.53 RMI (ft2) Large ( 4) 1674.01 3060.37 4658 3251.84 3

Qualified Epoxy (ft ) Particulate 0.25 0.12 0.13 0.12 3

Qualified IOZ (ft ) Particulate 0.07 0.08 0.00 0.00 E3-19

Enclosure 3 Updated Final Responses to Generic Letter 2004-02 Table 3.b.4-3: PBN2 Generated Debris Quantities for Worst-Case DEGBs RC-36-MRCL-BII- RC-34-MRCL-AI-Break Location RC-36-MRCL-BII-01A RC-34-MRCL-AI-03 01R1 04R1 Description Worst Fiber Breaks Worst Cal-Sil Breaks Break Size 31" 31" 29" 29" Break Type DEGB DEGB DEGB DEGB Fine 0.03 0.02 1.81 1.75 Small 0.03 0.02 7.20 6.88 LDFG (ft3)

Large 0.22 0.18 0.10 0.28 Intact 0.92 0.97 0.11 0.30 Mineral Wool (ft3) Fine 128.34 128.67 103.47 107.84 Fine 61.40 60.09 34.98 38.89 Small 207.56 202.29 105.39 121.05 Nukon (ft3)

Large 108.90 108.98 99.04 98.96 Intact 117.66 117.75 107.03 106.94 Fine 0.00 0.00 0.00 0.00 Small 0.00 0.00 0.00 0.00 Temp-Mat (ft3)

Large 0.00 0.00 0.00 0.00 Intact 0.00 0.00 0.00 0.00 Cal-Sil and Particulate 199.56 196.54 767.31 756.99 Asbestos Cal-Sil Small 139.23 137.16 514.90 495.56 (lbm) Intact 310.74 312.46 1152.49 1192.39 Mirror and Transco Small (<4) 1246.11 1242.18 1115.84 1408.30 RMI (ft2) Large ( 4) 415.37 414.06 371.95 469.43 3

Qualified Epoxy (ft ) Particulate 1.70 1.71 1.40 1.29 3

Qualified IOZ (ft ) Particulate 0.49 0.49 0.41 0.49 E3-20

Enclosure 3 Updated Final Responses to Generic Letter 2004-02 Table 3.b.4-4: The Four Worst-Case Breaks that Do Not Fail Any Acceptance Criteria for the Single Train Failure Equipment Configuration for PBN2 Break Location RC-34-MRCL-BI-03 RC-34-MRCL-BI-03 RC-10-AC-2001-07 RC-36-MRCL-AII-05 Description Worst Fiber Breaks Worst Cal-Sil Breaks Break Size 20 20 8.75 26 Break Type Partial (315q) Partial (0q) Full (0q) Partial (90q)

Fine 0.00 0.00 0.34 1.21 Small 0.00 0.00 0.42 4.18 LDFG (ft3)

Large 0.00 0.00 2.69 1.84 Intact 0.00 0.00 2.91 1.99 3

Mineral Wool (ft ) Fine 43.43 38.41 0.00 1.56 Fine 30.19 32.02 1.17 7.40 Small 98.57 105.19 2.09 14.39 Nukon (ft3)

Large 63.54 65.59 7.42 43.61 Intact 68.66 70.87 8.03 47.13 Fine 0.00 0.00 0.00 0.00 Small 0.00 0.00 0.00 0.00 Temp-Mat (ft3)

Large 0.00 0.00 0.00 0.00 Intact 0.00 0.00 0.00 0.00 Cal-Sil and Particulate 47.74 50.32 175.41 171.06 Asbestos Cal-Sil Small 30.13 28.66 130.64 111.31 (lb) Intact 94.23 120.53 205.75 272.47 Mirror and Transco Small (<4) 184.16 252.93 0.00 124.68 RMI (ft2) Large ( 4) 61.39 84.31 0.00 41.56 3

Qualified Epoxy (ft ) Particulate 0.20 0.35 0.07 0.21 3

Qualified IOZ (ft ) Particulate 0.09 0.14 0.00 0.13 E3-21

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. The amount of foreign materials recorded for PBN1 and PBN2 was 120 ft² and 152 ft², respectively. However, for conservatism, a total surface area of 200 ft² was assumed in the PBN1 and PBN2 debris generation analyses.

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02

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. See Response to 3.d.3 for the material properties of latent debris.

Table 3.c.1-1: Debris Material Properties Density Characteristic Debris Distribution (lbm/ft³) Size (m) 2.4 (bulk)

Nukon/LDFG See section below 7 159 (fiber) 11.8 (bulk)

Temp-Mat See section below 9 162 (fiber) 8 (bulk)

Mineral Wool 100% Fines 5-7 90 (fiber) 75% small pieces <4 Mirror/ Transco RMI -

25% large pieces 4 14.5 (bulk)

Cal-Sil 5 See section below 144 (particulate)

Asbestos Cal-Sil 16 (bulk) 10 300 (Dimetcote 6 -

IOZ) 97.1 (Amercoat 66 -

Epoxy)

Qualified Coatings 100% Particulate 10 101.3 (Phenoline 305 - Epoxy) 109 (Carboline 195 -

Epoxy) 208 (IOZ)

Unqualified and 100% Particulate 94 (Epoxy) 10 Degraded Coatings 98 (Alkyd)

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02 Nukon Low-Density Fiberglass Insulation The debris characteristics for Nukon, and generic LDFG are listed in Table 3.c.1-1.

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 11, Section 3.4.3.3.1). The PBN debris generation calculation used a four-category size distribution based on the guidance in NEI 04-07 Volume 2 (Reference 8, Appendix II and Appendix VI, p. VI-14). This guidance provides an approach for determining a size distribution for LDFG using the AJIT data, with conservatism added due to the potentially higher level of destruction from a two-phase jet. Within the 17.0D ZOI, the size distribution varies based on the distance of the insulation from the break (i.e.,

insulation debris generated near the break location consists of more small pieces than insulation debris generated near the edge of the ZOI).

Based on NEI 04-07 guidance, 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).

if

  if

 if

if

  if

 if

if

  if

if

if

  if

if

E3-24

Enclosure 3 Updated Final Responses to Generic Letter 2004-02 Temp-Mat High-Density Fiberglass Insulation The debris characteristics for Temp-Mat are listed in Table 3.c.1-1.

Similar to Nukon and other types of LDFG, a refinement to the standard methodology was used and takes into account a size distribution for Temp-Mat using AJIT data.

Based on NEI 04-07 guidance, 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 within an 11.7D ZOI between the break point and the centroid of the affected debris measured in units of pipe diameters (C).

if

  if

 if

if

  if

 if

if

  if

if

if

  if

if

Cal-Sil Insulation The debris characteristics for Cal-Sil and Asbestos Cal-Sil are listed in Table 3.c.1-1.

Similar to Nukon and other types of LDFG, a refinement to the standard methodology was used and takes into account a size distribution for Cal-Sil using jet test data. The following equations are developed to determine the fraction of fines (particulate), small pieces (less than 1 inch up to 3 inches), and intact pieces (remains on the target) as a function of the average distance within a 6.4D ZOI between the break point and the centroid of the affected debris measured in units of pipe diameters (C).

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02 if

  if

 if

if

  if

 if

if

  if

if

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, testing was used to determine strainer head loss and not an analytical method, so specific surface areas were not calculated or used for the PBN head loss evaluation (see the Response to 3.f).

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

Response to 3.c.4:

The debris characterizations for all debris types follow NRC-approved guidance.

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02

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 following discussion summarizes the methods used in the latent debris calculation.

The method used to estimate the quantity of latent debris was a representative sampling of containment surfaces as described in the guidance of NEI 04-07 Volume 2 (Reference 8, pp. 45-50). The samples were taken by Masslinn swipes and the amount of accumulated dust and lint quantified by weight. The fiber content of the latent debris was assumed to be 15% by weight, consistent with NEI 04-07 Volume 2 (Reference 8, p. 50). The balance of the latent debris is assumed to be particulate, also consistent with NEI 04-07 (Reference 8, p. 50).

Samples were taken to determine the latent debris mass distribution per unit area of representative surfaces throughout containment including vertical surfaces such as the liner and walls. These debris densities were then applied to all of the surface areas inside containment to calculate the total amount of latent debris inside containment.

The latent debris density was estimated by weighing Masslinn swipes before and after sampling, and dividing the net weight increase by the sampled surface area.

There were 21 samples taken at each unit, and included a mix of both horizontal and vertical surfaces, as well as surfaces that are routinely decontaminated and those surfaces that are not, such as the top surfaces of overhead duct work, cable trays, etc.

Because of the several different types of insulation used in the two containments, the statistical sample mass collections (e.g., three samples from each category of surface) was not used. PBN used an alternative approach to minimize personnel risk and exposure.

Representative samples were taken from accessible surfaces. Visual observations of these sample locations were compared to visual observations of other surfaces and estimates of bounding debris loadings were made. Although similar in magnitude, the data from PBN1 and the data from PBN2 were used to substantiate unit-specific latent debris source terms for both units.

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02

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

Response to 3.d.2:

There were three assumptions used in the evaluation of latent debris in containment.

These assumptions and their technical bases follow.

Assumption 1: The top surfaces area of the major structural heat sinks are periodically decontaminated.

Basis: Accessible floor areas are routinely wiped down to control contamination spread and to reduce the quantity of latent debris in containment. While there are top surfaces of major structural heat sinks that are not routinely cleaned due to ALARA concerns or inaccessibility (such as the regenerative heat exchanger room, the bottom of the pressurizer cubicle, etc.), most of these areas are also above the El. 8' sump and sheltered from direct spray impingement and washdown. Additional areas were added to account for those areas that are not routinely cleaned. Therefore, assuming that 100% of the floor areas are routinely cleaned over-estimates the total area that is routinely cleaned while not diminishing those areas that are not cleaned. The result is a conservatively high estimate of the routinely cleaned horizontal surface areas.

Assumption 2: The horizontal surface area of containment that is not routinely cleaned, and is subject to direct spray impingement and/or washdown during a LOCA, is equal to the horizontal surface area that is routinely cleaned per Assumption 1.

Basis: The horizontal surface areas not routinely cleaned yet still subject to wash down are primarily limited to those above the refueling floor El. 66'. Horizontal areas above this elevation are very limited, primarily due to the necessity of moving large loads above the floor such as the reactor vessel head, RCP motors, etc. Areas below El. 66' are largely sheltered from direct spray impingement, and only those in the RCS loop compartments may be subjected to scouring during the blowdown phase of a LOCA.

Assumption 3: The vertical surface area of miscellaneous equipment such as cable trays, ladders, tanks, etc. is equal to the vertical surface area of all the major structural heat sinks inside of containment.

Basis: The major structural heat sinks include the containment building wall and all compartment walls. Other major vertical surface areas are equipment such as the steam generators, the pressurizer, the RCP motors, and the reactor vessel. In addition, there are various cable trays, piping, ladders, etc. The vertical surface of any tank or vessel is less than the vertical surface of the compartment surrounding it.

Considering that much of the vertical surface areas are sheltered from spray impingement by floors above, and that there is a substantial amount of vertical surface area represented by the containment liner itself, the assumption was considered a reasonable and bounding approximation.

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02

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:

The quantity of latent debris was assumed to be 150 lbm, but the actual amount of latent debris documented for the plant is 62 lbm for PBN1 and 55 lbm for PBN2. These quantities are well below the quantity used to determine the strainer head loss.

Table 3.d.3-1 lists the assumed latent fiber and particulate constituents and their material characteristics.

Latent debris was assumed to consist of 15 percent fiber and 85 percent particulate by mass per the NRC NEI 04-07 SE (Reference 8, p. 50).

Based on NEI 04-07 Volume 2 (Reference 8, p. 50-52, V-11), the size and density of latent particulate were assumed to be 17.3 m (specific surface area of 106,000 ft-1) and 168.6 lbm/ft³ (2.7 g/cm3), respectively. Additionally, the bulk density and microscopic density of latent fiber were assumed to be 2.4 lbm/ft³ and 93.6 lbm/ft³ (1.5 g/cm3), respectively.

Latent fiber was assumed to have a characteristic size of 5.5 m. This is reasonably conservative, as it is the smallest fiber diameter listed in Table 3-2 of the general reference for LDFG found in NEI 04-07 (Reference 11, p. 3-28).

Table 3.d.3-1: Latent Fiber and Particulate Constituents Latent Bulk Microscopic Characteristic Volume Debris Density Density Size (ft3)

(lbm) (lbm/ft³) (lbm/ft³) (m)

Particulate (85%) 127.5 - 168.6 17.3 0.756 Fiber (15%) 22.5 2.4 93.6 5.5 9.375 Total 150

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

Response to 3.d.4:

As discussed in the Response to 3.b.5, a total surface area of 200 ft² of miscellaneous debris was conservatively assumed in the PBN1 and PBN2 debris generation calculations. This surface area would result in a 150 ft² reduction in strainer area (75%

of 200 ft²) (Reference 8, p. 49).

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02

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 8) for refined analyses, as well as the refined methodologies suggested by the SE in Appendices III, IV, and VI (Reference 8). The specific effect of each of the four modes of transport was analyzed in the debris transport calculations for each type of debris generated. These modes of transport are:

x Blowdown Transport - the vertical and horizontal transport of debris to all areas of containment by the break jet x Washdown Transport - the vertical (downward) transport of debris by the containment sprays, break flow, and condensation x Pool Fill-Up Transport - the transport of debris by break and containment spray flows from the refueling water storage tank (RWST) to regions that may be active or inactive during recirculation x 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 listed in 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 8) 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.

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02 Figure 3.e.1-1: Generic Debris Transport Logic Tree E3-31

Enclosure 3 Updated Final Responses to Generic Letter 2004-02 The basic methodology for the PBN1 and PBN2 transport analysis is summarized below.

1. The CAD model was used to determine break locations and sizes.

2. The debris generation calculation was used to identify debris types and sizes.

3. Potential upstream blockage points were qualitatively addressed.

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 CFD model was developed 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 CFD model output, along with the determined initial distribution of debris.

10. The initial recirculation transport fractions from the CFD 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 no upstream blockage points in the PBN1 and PBN2 containment buildings that adversely impact sump level.

Upstream effects are discussed in the Response to 3.l.

CFD Model of Containment Recirculation Pool A diagram showing the significant parts of the CFD model is shown in Figure 3.e.1-2 for PBN1, and Figure 3.e.1-3 for PBN2. The strainer module mass sinks and the various direct and runoff spray regions are highlighted.

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02 Figure 3.e.1-2: Significant Features in CFD Model (PBN1)

Figure 3.e.1-3: Significant Features in CFD Model (PBN2)

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02 The key CFD modeling attributes/considerations included the following:

Computational Mesh A rectangular mesh was defined in the CFD model that was fine enough to resolve important features, but not so fine that the simulation would take excessively long to run. A 6-inch cell length was chosen as the largest cell size that could reasonably resolve the concrete structures that compose the containment floor. For the cells right above the containment floor (8 Elevation and 10 Elevation), the mesh was set to 3 inches tall in order to closely resolve the vicinity (area right above the floor where tumbling velocities are analyzed) of settled debris. The total cell count in the model was 3,904,112 at both PBN1 and PBN2.

Modeling of Containment Spray Flows Various plan and section drawings, as well as the containment building CAD model, were considered when determining the spray flow path to the pool. Spray water would drain to the pool through many pathways. Some of these pathways include stairways

1. 22 and #23 (PBN1), stairways #38 and #39 (PBN2), and for both units, the steam generator compartments through the open area above the steam generators and RCPs, the keyway (reactor cavity), the 3-inch gap around the periphery, and the 4-inch diameter drain line from the refueling canal. The sprays were defined as regions and populated with discrete mass source particles. The appropriate flow rate and velocity was set for the sprays in each region.

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 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 Strainers The A strainer and the B strainer at both PBN1 and PBN2 each consist of strainer modules that sit 3 inches above the floor. Each strainer array in the CFD model is modeled as having flow across its surfaces proportional to the area of each strainer.

Each unit of stacked disks was modeled to draw flow from all surfaces (including the bottom surface). A negative flow rate was set for the strainer modules, which tells the CFD model to draw the specified amount of water from the pool over the entire exposed surface area of the module obstacle.

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02 Turbulence Modeling Several different turbulence-modeling approaches can be selected for a Flow-3D calculation. The approaches (ranging from least to most sophisticated) are:

x Prandtl mixing length x Turbulent energy model x Two-equation k- model x Renormalized group theory (RNG) model x Large eddy simulation model The RNG turbulence model was determined to be the most appropriate for this CFD analysis. The RNG model has a large spectrum of length scales that would likely exist in a containment pool during 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.

Steady-State Metrics The CFD model was 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 PBN1 and PBN2 transport analyses originate from the following sources.

x NUREG/CR-6772 Tables 3.1, 3.5, and C.19(a) (Reference 18, pp. 16, 22, and C-16) x NUREG/CR-6808 Figure 5.2, Table 5-1 and Table 5-3 (Reference 19, pp. 5-14, 5-22, and 5-33)

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.

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02 x Colored contour velocity and TKE maps were generated from the Flow-3D results in the form of bitmap files indicating regions of the pool through which a particular type of debris could be expected to transport.

x 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.

x Closed polylines were drawn around the contiguous areas where velocity and TKE were high enough that debris could be carried in suspension or tumbled along the floor to the sump strainers for uniformly distributed debris.

x The areas within the closed polylines were determined using an AutoCAD querying feature.

x The combined area within the polylines was compared to the initial debris distribution area.

x 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 (refer to the figures that follow). The overlying yellow areas represent 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.1-6 and Figure 3.e.1-7). 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 PBN1 and PBN2.

As shown in Figure 3.e.1-4 (PBN1) and Figure 3.e.1-5 (PBN2), the small debris (depicted by green shading) was initially assumed to be distributed in the vicinity of the break location for a postulated Loop B break in each unit at the beginning of recirculation.

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02 Figure 3.e.1-4: Distribution of Small Debris in Lower Containment (PBN1)

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02 Figure 3.e.1-5: Distribution of Small Debris in Lower Containment (PBN2)

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02 For PBN1, Figure 3.e.1-6 shows that the turbulence (yellow regions) and the velocity (red regions) in the pool (blue regions) generated by the break flow are not high enough to transport the generic small debris present in the pool to the sump strainers during recirculation. Therefore, the transport for small debris blown to lower containment is 0% for PBN1.

Figure 3.e.1-6: TKE and Velocity with Limits Set at Suspension/

Tumbling of Small Generic Debris (PBN1)

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02 For PBN2, Figure 3.e.1-7 shows that the turbulence (yellow regions) and the velocity (red regions) in the pool (blue regions) generated by the break flow are high enough to transport the generic small debris present in the pool to the sump strainers during recirculation. The initial distribution area (Figure 3.e.1-5) was overlaid on top of the plot showing tumbling velocity, TKE, and flow vectors (Figure 3.e.1-7) to determine the recirculation transport fraction (Figure 3.e.1-8).

Figure 3.e.1-7: TKE and Velocity with Limits Set at Suspension/

Tumbling of Small Generic Debris (PBN2)

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02 Figure 3.e.1-8: Floor Area where Small Generic Debris Would Transport to the Sump Strainers (hatched area - PBN2)

This same analysis was applied for each type of debris at PBN1 and PBN2.

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 through the annulus.

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02 Erosion Discussion Due to the turbulence in the recirculation pool and the force of break and spray flow, Nukon (PBN2 only), LDFG, Temp-Mat, and Cal-Sil debris may erode into smaller pieces, making transport of this debris to the strainer more likely. To estimate erosion that would occur in the recirculation pools at PBN1 and PBN2, generic 30-day testing was performed. Based on a validation that the results apply to PBN1 and PBN2 (ensuring that the flow rates and turbulence values are similar to what is expected in the PBN1 and PBN2 recirculation pools), an erosion fraction of 10% was assumed for the small and large pieces of fiberglass debris in the pool. An erosion fraction of 17%

was assumed for the small chunks of Cal-Sil debris in the pool. This fraction was applied to both transportable debris and settled debris present in the pool to maximize the amount of erosion. For pieces of debris held up on grating above the pool, an erosion fraction of 1% was used for fiberglass debris, and 17% for Cal-Sil debris.

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 was based on and does not deviate from the NRC approved NEI 04-07 guidance and the associated NRC SE for refined analyses, as well as the refined methodologies suggested by the SE in Appendices III, IV, and VI (Reference 8).

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 computational fluid dynamics (CFD) simulations were run using Flow-3D, a commercially available software package.

For PBN1, four break cases form the basis for the debris transport analysis to determine the recirculation transport fractions. Two cases were analyzed for each operational strainer - a break in Loop A and a break in Loop B (a total of four cases).

All cases were run with the maximum ECCS flow rate through the strainer (2,200 gpm) and with the minimum water level at the start of full recirculation (4.23 ft). Using the maximum flow rates and minimum water level maximize the turbulence and velocity in the pool.

For PBN2, four break cases also form the basis for the debris transport analysis to determine the recirculation transport fractions. Two cases were analyzed for each operational strainer - a break in Loop A and a break in Loop B (a total of four cases).

All cases were run with the maximum ECCS flow rate through the strainer (2,200 gpm)

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02 and with the minimum water level at the start of full recirculation (4.23 ft) . Using the maximum flow rates and the minimum water level maximize the turbulence and velocity in the pool.

In general, a break close to the strainer tends to transport a larger fraction of small and large debris than a break farther from the strainer. 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 GSI-191 literature to determine the recirculation transport fractions for all debris types present in the PBN1 and PBN2 containment buildings. See the Response to 3.e.1 for additional discussion of the CFD results.

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

Response to 3.e.4:

At PBN1, debris interceptors within the secondary shield wall were not credited for preventing any debris from reaching the strainer. In the CFD model, the interceptors were assumed to be completely blocked during the simulation (debris interceptors were modeled as flow diverters). This conservatively causes all of the flow to be diverted through the open passageways between the steam generator compartments and the annulus to the strainers which increases the velocities in the pool.

No credit was taken for debris interceptors at PBN2.

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.

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02

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 are shown for blowdown, washdown, pool fill, and recirculation. Note that these fractions result in the bounding quantity of debris transported to the strainer. Cells with a - in the tables of this subsection represent values that are not applicable (i.e., debris type not generated for a specific location, debris type not available for washdown/pool-fill, etc.).

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02 Blowdown Transport Table 3.e.6-1 and Table 3.e.6-2 show the bounding (minimum amount of debris remaining in the compartment) blowdown transport fractions as a function of break location and debris type. Note that only the limiting break locations with respect to the maximum overall debris transport fractions are listed in these tables.

Table 3.e.6-1: Blowdown Transport Fractions (PBN1)

Transport Fraction Break To Upper To Lower Debris Type Remaining in Location Containment Containment Compartment (UC) (LC)

Fines (all) 66% 34% 0%

Small Fiberglass 57% 30% 13%

Large Fiberglass 27% 12% 61%

Steam Intact Fiberglass Blankets 0% 0% 100%

Generator Small RMI 64% 33% 3%

(SG) Large RMI 40% 15% 45%

Compartments Small Cal-Sil 64% 33% 3%

Qualified Coatings 66% 34% 0%

Unqualified Coatings - - -

Latent Debris - - -

Fines (all) 66% 34% 0%

Small Fiberglass 57% 30% 13%

Large Fiberglass 27% 12% 61%

Intact Fiberglass Blankets 0% 0% 100%

Small RMI not in Cavity 64% 33% 3%

Small RMI in Cavity 20% 30% 50%

Reactor Cavity Large RMI not in Cavity 40% 15% 45%

Large RMI in Cavity 0% 0% 100%

Small Cal-Sil not in Cavity 64% 33% 3%

Small Cal-Sil in Cavity 20% 30% 50%

Qualified Coatings 66% 34% 0%

Unqualified Coatings - - -

Latent Debris - - -

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02 Transport Fraction Break To Upper To Lower Debris Type Remaining in Location Containment Containment Compartment (UC) (LC)

Fines (all) 66% 34% 0%

Small Fiberglass 64% 29% 7%

Large Fiberglass 40% 12% 48%

Intact Fiberglass Blankets 0% 0% 100%

Pressurizer Small RMI - - -

Compartment Large RMI - - -

Small Cal-Sil 65% 33% 2%

Qualified Coatings 66% 34% 0%

Unqualified Coatings - - -

Latent Debris - - -

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02 Table 3.e.6-2: Blowdown Transport Fractions (PBN2)

Transport Fraction Break To Upper To Lower Debris Type Remaining in Location Containment Containment Compartment (UC) (LC)

Fines (all) 65% 35% 0%

Small Fiberglass 57% 31% 12%

Large Fiberglass 26% 13% 61%

Intact Fiberglass Blankets 0% 0% 100%

SG Small RMI 63% 34% 3%

Compartments Large RMI 40% 20% 40%

Small Cal-Sil 63% 34% 3%

Qualified Coatings 65% 35% 0%

Unqualified Coatings - - -

Latent Debris - - -

Fines (all) 65% 35% 0%

Small Fiberglass 57% 31% 12%

Large Fiberglass 26% 13% 61%

Intact Fiberglass Blankets 0% 0% 100%

Small RMI not in Cavity 63% 34% 3%

Small RMI in Cavity 20% 30% 50%

Reactor Cavity Large RMI not in Cavity 40% 20% 40%

Large RMI in Cavity 0% 0% 100%

Small Cal-Sil not in Cavity 63% 34% 3%

Small Cal-Sil in Cavity 20% 30% 50%

Qualified Coatings 65% 35% 0%

Unqualified Coatings - - -

Latent Debris - - -

Fines (all) 65% 35% 0%

Small Fiberglass 63% 30% 7%

Large Fiberglass 40% 12% 48%

Intact Fiberglass Blankets 0% 0% 100%

Pressurizer Small RMI 65% 34% 1%

Compartment Large RMI 45% 25% 30%

Small Cal-Sil 65% 34% 1%

Qualified Coatings 65% 35% 0%

Unqualified Coatings - - -

Latent Debris - - -

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02 Washdown Transport Table 3.e.6-3 and Table 3.e.6-4 show the bounding washdown transport fractions (maximum amount of debris washed to lower containment) for each debris type. Note that these transport fractions do not depend on the location of the break. The difference in the fractions between the two units is due to the presence of a curb around the perimeter of the operating deck in PBN1.

Table 3.e.6-3: Washdown Transport Fractions (PBN1)

Transport Fraction Washed Down Washed Down Washed Debris Type Inside Steam Refueling Down in Generator Canal (RFC)

Annulus Compartments Drain Fines/Particulate (all) 59% 24% 17%

Small Fiberglass 59% 19% 17%

Large Fiberglass 59% 0% 17%

Intact Fiberglass Blankets Small RMI 59% 24% 17%

Large RMI 59% 0% 17%

Small Cal-Sil 59% 24% 17%

Qualified Coatings 59% 24% 17%

Unqualified Coatings - - -

Latent Debris - - -

Table 3.e.6-4: Washdown Transport Fractions (PBN2)

Transport Fraction Washed Down Washed Debris Type Inside Steam Washed Down Down in Generator RFC Drain Annulus Compartments Fines/Particulate (all) 63% 21% 16%

Small Fiberglass 63% 16% 16%

Large Fiberglass 63% 0% 16%

Intact Fiberglass Blankets Small RMI 63% 21% 16%

Large RMI 63% 0% 16%

Small Cal-Sil 63% 21% 16%

Qualified Coatings 63% 21% 16%

Unqualified Coatings - - -

Latent Debris - - -

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02 Pool-Fill Transport The calculation used to determine the portion of debris washed to inactive cavities during pool fill is based on the following equation:

Where:

Xpool-fill = Amount of debris transported to cavity during pool fill Vcavity = Cavity volume Vpool = Pool volume (sum of active and inactive volumes)

The primary cavity below the floor elevation at PBN1 and PBN2 is the reactor cavity.

As the pool fills in the PBN containment, the reactor cavity would fill first. Water would flow through the 16 core bore at the bottom of the keyway in the reactor cavity from the containment recirculation pool. The volume of the reactor cavity at a recirculation pool water level of 6 inches was calculated to be 3,267 ft3 for PBN1 and 3,189 ft3 for PBN2. The volume of the pool in the recirculation sump at six inches was calculated to be 2,654 ft3 for PBN1 and 2,179 ft3 for PBN2.

Inserting these values into the equation above yields a pool fill-up debris transport fraction of 71% for PBN1 and 77% for PBN2. However, debris transport to the inactive cavity was limited to 15% by Section 3.6.3 of the SER (Reference 8).

Table 3.e.6-5 shows the minimum pool fill transport fractions for non-reactor cavity breaks as a function of debris type, applicable for both PBN units. The pool fill transport is not applicable for breaks inside the reactor cavity.

Table 3.e.6-5: Pool Fill Transport Fractions For Non-Reactor Cavity Breaks Pool Fill Transport Fraction Debris Type Directly to Strainer Inactive Cavity Fines/Particulate (all) 0% 15%

Small Fiberglass 0% 15%

Large Fiberglass 0% 15%

Intact Fiberglass Blankets - -

Small Temp-Mat 0% 15%

Large Temp-Mat 0% 15%

Intact Temp-Mat Blankets - -

Small RMI 0% 15%

Large RMI 0% 15%

Small Cal-Sil 0% 15%

Qualified Coatings 0% 15%

Unqualified Coatings - -

Latent Debris 0% 15%

E3-49

Enclosure 3 Updated Final Responses to Generic Letter 2004-02 Recirculation Transport For the recirculation transport fractions, four different break cases form the basis for each debris transport analysis, and were evaluated in each of the debris transport calculations. Note that recirculation transport fractions are presented separately for each unit. This is because the location of the strainers and the presence of debris interceptors are different between the two units.

The cases for PBN1 are:

x Case 1: LBLOCA in Loop A, Train A Operational x Case 2: LBLOCA in Loop A, Train B Operational x Case 3: LBLOCA in Loop B, Train A Operational x Case 4: LBLOCA in Loop B, Train B Operational The cases for PBN2 are:

x Case 1: LBLOCA in Loop A, Train A Operational x Case 2: LBLOCA in Loop B, Train A Operational x Case 3: LBLOCA in Loop A, Train B Operational x Case 4: LBLOCA in Loop B, Train B Operational It was assumed that for any breaks that could occur in the reactor cavity or in the pressurizer compartment, the recirculation transport fractions for a break inside the secondary shield wall (Loop A or Loop B for a reactor cavity break, and Loop B for a pressurizer break) could be applied.

The bounding (maximum) recirculation transport fractions for fibrous debris as a function of evaluation case are shown in Table 3.e.6-6 and Table 3.e.6-7.

See Response to 3.e.1 for the methodology used for recirculation transport.

E3-50

Enclosure 3 Updated Final Responses to Generic Letter 2004-02 Table 3.e.6-6: Recirculation Transport Fractions for Fibrous Debris (PBN1)

Debris Debris Debris in Washed inside Debris Washed Case Debris Size Lower Steam Washed in down RFC Containment Generator Annulus Drain Compartments Fines 100% 100% 100% 100%

Small 0% 8% 21% 0%

Case 1 Large 0% - 0% 0%

Intact Blankets - - - -

Fines 100% 100% 100% 100%

Small 0% 0% 7% 0%

Case 2 Large 0% - 4% 0%

Intact Blankets - - - -

Fines 100% 100% 100% 100%

Small 0% 0% 3% 0%

Case 3 Large 0% - 0% 0%

Intact Blankets - - - -

Fines 100% 100% 100% 100%

Small 0% 0% 7% 0%

Case 4 Large 0% - 4% 0%

Intact Blankets - - - -

Table 3.e.6-7: Recirculation Transport Fractions for Fibrous Debris (PBN2)

Debris Washed Debris Debris in Debris inside Steam Washed Case Debris Size Lower Washed in Generator down RFC Containment Annulus Compartments Drain Fines 100% 100% 100% 100%

Small 32% 4% 42% 0%

Case 1 Large 0% - 11% 0%

Intact Blankets - - - -

Fines 100% 100% 100% 100%

Small 72% 41% 50% 0%

Case 2 Large 11% - 9% 0%

Intact Blankets - - - -

Fines 100% 100% 100% 100%

Small 68% 44% 45% 0%

Case 3 Large 0% - 0% 0%

Intact Blankets - - - -

Fines 100% 100% 100% 100%

Small 34% 33% 16% 0%

Case 4 Large 0% - 0% 0%

Intact Blankets - - - -

E3-51

Enclosure 3 Updated Final Responses to Generic Letter 2004-02 The bounding recirculation transport fractions for Temp-Mat debris as a function of evaluation case are shown in Table 3.e.6-8 and Table 3.e.6-9. It was conservatively assumed that Temp-Mat debris would float in the recirculation pool until it is transported to the vicinity of the strainers, which results in a recirculation transport fraction of 100%.

Table 3.e.6-8: Recirculation Transport Fractions for Temp-Mat (PBN1)

Debris Debris Debris in Washed inside Debris Washed Case Debris Size Lower Steam Washed in down RFC Containment Generator Annulus Drain Compartments Fines 100% 100% 100% 100%

Small 100% 100% 100% 100%

Case 1 Large 100% 100% 100% 100%

Intact Blankets - - - -

Fines 100% 100% 100% 100%

Small 100% 100% 100% 100%

Case 2 Large 100% 100% 100% 100%

Intact Blankets - - - -

Fines 100% 100% 100% 100%

Small 100% 100% 100% 100%

Case 3 Large 100% 100% 100% 100%

Intact Blankets - - - -

Fines 100% 100% 100% 100%

Small 100% 100% 100% 100%

Case 4 Large 100% 100% 100% 100%

Intact Blankets - - - -

E3-52

Enclosure 3 Updated Final Responses to Generic Letter 2004-02 Table 3.e.6-9: Recirculation Transport Fractions for Temp-Mat (PBN2)

Debris Debris Debris in Washed inside Debris Washed Case Debris Size Lower Steam Washed in down RFC Containment Generator Annulus Drain Compartments Fines 100% 100% 100% 100%

Small 100% 100% 100% 100%

Case 1 Large 100% - 100% 100%

Intact Blankets - - - -

Fines 100% 100% 100% 100%

Small 100% 100% 100% 100%

Case 2 Large 100% - 100% 100%

Intact Blankets - - - -

Fines 100% 100% 100% 100%

Small 100% 100% 100% 100%

Case 3 Large 100% - 100% 100%

Intact Blankets - - - -

Fines 100% 100% 100% 100%

Small 100% 100% 100% 100%

Case 4 Large 100% - 100% 100%

Intact Blankets - - - -

The bounding recirculation transport fractions for RMI debris as a function of evaluation case are shown in Table 3.e.6-10 and Table 3.e.6-11.

Table 3.e.6-10: Recirculation Transport Fractions for RMI Debris (PBN1)

Debris Debris Debris in Washed inside Debris Washed Case Debris Size Lower Steam Washed in down RFC Containment Generator Annulus Drain Compartments Small 0% 0% 0% 0%

Case 1 Large 0% - 0% 0%

Small 0% 0% 5% 0%

Case 2 Large 0% - 5% 0%

Small 0% 0% 1% 0%

Case 3 Large 0% - 1% 0%

Small 0% 0% 5% 0%

Case 4 Large 0% - 5% 0%

E3-53

Enclosure 3 Updated Final Responses to Generic Letter 2004-02 Table 3.e.6-11: Recirculation Transport Fractions for RMI Debris (PBN2)

Debris Debris Debris in Washed inside Debris Washed Case Debris Size Lower Steam Washed in down RFC Containment Generator Annulus Drain Compartments Small 0% 0% 16% 0%

Case 1 Large 0% - 16% 0%

Small 16% 8% 13% 0%

Case 2 Large 16% - 13% 0%

Small 0% 0% 0% 0%

Case 3 Large 0% - 0% 0%

Small 0% 0% 0% 0%

Case 4 Large 0% - 0% 0%

The bounding recirculation transport fractions for Cal-Sil debris as a function of evaluation case are shown in Table 3.e.6-12 and Table 3.e.6-13.

Table 3.e.6-12: Recirculation Transport Fractions for Cal-Sil Debris (PBN1)

Debris Debris Debris in Washed inside Debris Washed Case Debris Size Lower Steam Washed in down RFC Containment Generator Annulus Drain Compartments Particulate 100% 100% 100% 100%

Case 1 Small 0% 0% 0% 0%

Particulate 100% 100% 100% 100%

Case 2 Small 0% 0% 5% 0%

Particulate 100% 100% 100% 100%

Case 3 Small 0% 0% 2% 0%

Particulate 100% 100% 100% 100%

Case 4 Small 0% 0% 5% 0%

Table 3.e.6-13: Recirculation Transport Fractions for Cal-Sil Debris (PBN2)

Debris Debris Debris in Washed inside Debris Washed Case Debris Size Lower Steam Washed in down RFC Containment Generator Annulus Drain Compartments Particulate 100% 100% 100% 100%

Case 1 Small 0% 0% 16% 0%

Particulate 100% 100% 100% 100%

Case 2 Small 15% 5% 14% 0%

Particulate 100% 100% 100% 100%

Case 3 Small 0% 0% 0% 0%

Particulate 100% 100% 100% 100%

Case 4 Small 0% 0% 0% 0%

E3-54

Enclosure 3 Updated Final Responses to Generic Letter 2004-02 The bounding recirculation transport fractions for qualified coatings, unqualified coatings, and latent debris as a function of evaluation case are shown in Table 3.e.6-14 and Table 3.e.6-15.

Table 3.e.6-14: Recirculation Transport Fractions for Qualified Coatings, Unqualified Coatings, Latent Debris (PBN1)

Debris Washed Debris Debris in Debris inside Steam Washed Case Debris Size Lower Washed in Generator down RFC Containment Annulus Compartments Drain Case 1 Fine/Particulate 100% 100% 100% 100%

Case 2 Fine/Particulate 100% 100% 100% 100%

Case 3 Fine/Particulate 100% 100% 100% 100%

Case 4 Fine/Particulate 100% 100% 100% 100%

Table 3.e.6-15: Recirculation Transport Fractions for Qualified Coatings, Unqualified Coatings, Latent Debris (PBN2)

Debris Washed Debris Debris in Debris inside Steam Washed Case Debris Size Lower Washed in Generator down RFC Containment Annulus Compartments Drain Case 1 Fine/Particulate 100% 100% 100% 100%

Case 2 Fine/Particulate 100% 100% 100% 100%

Case 3 Fine/Particulate 100% 100% 100% 100%

Case 4 Fine/Particulate 100% 100% 100% 100%

E3-55

Enclosure 3 Updated Final Responses to Generic Letter 2004-02 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 overall transport fractions are provided in Table 3.e.6-16 through Table 3.e.6-25.

Table 3.e.6-16: Overall Transport Fractions for a Break in the SG Compartment Break in Loop A (PBN1)

Train A Train B Debris Type Debris Size Operational Operational Fines 95% 95%

Small Transport as Erosion Fines 8% 8%

Pieces Transport as Small Pieces 7% 2%

LDFG Large Transport as Erosion Fines 4% 4%

Pieces Transport as Large Pieces 0% 1%

Intact Blankets 0% 0%

Fines - -

Small Transport as Erosion Fines - -

Pieces Transport as Small Pieces - -

Temp-Mat Large Transport as Erosion Fines - -

Pieces Transport as Large Pieces - -

Intact Blankets - -

Small Pieces 0% 2%

Mirror RMI Large Pieces 0% 1%

Small Pieces 0% 2%

Transco RMI Large Pieces 0% 1%

Mineral Wool Fines 95% 95%

Fines 95% 95%

Cal-Sil Small Transport as Erosion Fines 16% 16%

Pieces Transport as Small Pieces 0% 2%

Fines 95% 95%

Asbestos Cal-Sil Small Transport as Erosion Fines 16% 16%

Pieces Transport as Small Pieces 0% 2%

Qualified Coatings Particulate 95% 95%

Unqualified Coatings Particulate 100% 100%

Latent Debris Particulate/Fiber 85% 85%

E3-56

Enclosure 3 Updated Final Responses to Generic Letter 2004-02 Table 3.e.6-17: Overall Transport Fractions for a Break in the SG Compartment in Loop B (PBN1)

Train A Train B Debris Type Debris Size Operational Operational Fines 95% 95%

Small Transport as Erosion Fines 8% 8%

Pieces Transport as Small Pieces 1% 2%

LDFG Large Transport as Erosion Fines 4% 4%

Pieces Transport as Large Pieces 0% 1%

Intact Blankets 0% 0%

Fines 95% 95%

Small Transport as Erosion Fines 8% 8%

Pieces Transport as Small Pieces 74% 74%

Temp-Mat Large Transport as Erosion Fines 4% 4%

Pieces Transport as Large Pieces 28% 28%

Intact Blankets 0% 0%

Small Pieces 0% 2%

Mirror RMI Large Pieces 0% 1%

Small Pieces 0% 2%

Transco RMI Large Pieces 0% 1%

Mineral Wool Fines 95% 95%

Fines 95% 95%

Cal-Sil Small Transport as Erosion Fines 16% 16%

Pieces Transport as Small Pieces 1% 2%

Fines 95% 95%

Asbestos Cal-Sil Small Transport as Erosion Fines 16% 16%

Pieces Transport as Small Pieces 1% 2%

Qualified Coatings Particulate 95% 95%

Unqualified Coatings Particulate 100% 100%

Latent Debris Particulate/Fiber 85% 85%

E3-57

Enclosure 3 Updated Final Responses to Generic Letter 2004-02 Table 3.e.6-18: Overall Transport Fractions for a Reactor Cavity Break Loop A (PBN1)

Train A Train B Debris Type Debris Size Operational Operational Fines 100% 100%

Small Transport as Erosion Fines 9% 9%

Pieces Transport as Small Pieces 7% 2%

LDFG Large Transport as Erosion Fines 4% 4%

Pieces Transport as Large Pieces 0% 1%

Intact Blankets 0% 0%

Fines - -

Small Transport as Erosion Fines - -

Pieces Transport as Small Pieces - -

Temp-Mat Large Transport as Erosion Fines - -

Pieces Transport as Large Pieces - -

Intact Blankets - -

Small Pieces 0% 2%

Mirror RMI Large Pieces 0% 1%

Small Pieces 0% 2%

Transco RMI Large Pieces 0% 1%

Transco RMI In Fines 0% 1%

Cavity Large Pieces 0% 0%

Mineral Wool Fines 100% 100%

Fines 100% 100%

Cal-Sil Small Transport as Erosion Fines 17% 17%

Pieces Transport as Small Pieces 0% 2%

Fines 100% 100%

Cal-Sil In Cavity Small Transport as Erosion Fines 17% 17%

Pieces Transport as Small Pieces 0% 1%

Fines 100% 100%

Asbestos Cal-Sil Small Transport as Erosion Fines 17% 17%

Pieces Transport as Small Pieces 0% 2%

Qualified Coatings Particulate 100% 100%

Unqualified Coatings Particulate 100% 100%

Latent Debris Particulate/Fiber 100% 100%

E3-58

Enclosure 3 Updated Final Responses to Generic Letter 2004-02 Table 3.e.6-19: Overall Transport Fractions for a Reactor Cavity Break Loop B (PBN1)

Train A Train B Debris Type Debris Size Operational Operational Fines 100% 100%

Small Transport as Erosion Fines 9% 9%

Pieces Transport as Small Pieces 1% 2%

LDFG Large Transport as Erosion Fines 4% 4%

Pieces Transport as Large Pieces 0% 1%

Intact Blankets 0% 0%

Fines 100% 100%

Small Transport as Erosion Fines 9% 9%

Pieces Transport as Small Pieces 78% 78%

Temp-Mat Large Transport as Erosion Fines 4% 4%

Pieces Transport as Large Pieces 29% 29%

Intact Blankets 0% 0%

Small Pieces 0% 2%

Mirror RMI Large Pieces 0% 1%

Small Pieces 0% 2%

Transco RMI Large Pieces 0% 1%

Transco RMI In Fines 0% 1%

Cavity Large Pieces 0% 0%

Mineral Wool Fines 100% 100%

Fines 100% 100%

Cal-Sil Small Transport as Erosion Fines 17% 17%

Pieces Transport as Small Pieces 1% 2%

Fines 100% 100%

Cal-Sil In Cavity Small Transport as Erosion Fines 17% 17%

Pieces Transport as Small Pieces 0% 1%

Fines 100% 100%

Asbestos Cal-Sil Small Transport as Erosion Fines 17% 17%

Pieces Transport as Small Pieces 1% 2%

Qualified Coatings Particulate 100% 100%

Unqualified Coatings Particulate 100% 100%

Latent Debris Particulate/Fiber 100% 100%

E3-59

Enclosure 3 Updated Final Responses to Generic Letter 2004-02 Table 3.e.6-20: Overall Transport Fractions for a Pressurizer Compartment Break (PBN1)

Train A Train B Debris Type Debris Size Operational Operational Fines 95% 95%

Small Transport as Erosion Fines 9% 9%

Pieces Transport as Small Pieces 1% 2%

LDFG Large Transport as Erosion Fines 5% 5%

Pieces Transport as Large Pieces 0% 1%

Intact Blankets 0% 0%

Fines 95% 95%

Small Transport as Erosion Fines 9% 9%

Pieces Transport as Small Pieces 78% 78%

Temp-Mat Large Transport as Erosion Fines 5% 5%

Pieces Transport as Large Pieces 37% 37%

Intact Blankets 0% 0%

Small Pieces - -

Mirror RMI Large Pieces - -

Small Pieces - -

Transco RMI Large Pieces - -

Mineral Wool Fines 95% 95%

Fines 95% 95%

Cal-Sil Small Transport as Erosion Fines 16% 16%

Pieces Transport as Small Pieces 1% 2%

Fines 95% 95%

Asbestos Cal-Sil Small Transport as Erosion Fines 16% 16%

Pieces Transport as Small Pieces 1% 2%

Qualified Coatings Particulate 95% 95%

Unqualified Coatings Particulate 100% 100%

Latent Debris Particulate/Fiber 85% 85%

E3-60

Enclosure 3 Updated Final Responses to Generic Letter 2004-02 Table 3.e.6-21: Overall Transport Fractions for a Break in the SG Compartment in Loop A (PBN2)

Train A Train B Debris Type Debris Size Operational Operational Fines 95% 95%

Small Transport as Erosion Fines 8% 8%

Pieces Transport as Small Pieces 22% 35%

LDFG Large Transport as Erosion Fines 4% 4%

Pieces Transport as Large Pieces 1% 0%

Intact Blankets 0% 0%

Fines 95% 95%

Small Transport as Erosion Fines 8% 8%

Pieces Transport as Small Pieces 22% 35%

Nukon Large Transport as Erosion Fines 4% 4%

Pieces Transport as Large Pieces 1% 0%

Intact Blankets 0% 0%

Fines - -

Small Transport as Erosion Fines - -

Pieces Transport as Small Pieces - -

Temp-Mat Large Transport as Erosion Fines - -

Pieces Transport as Large Pieces - -

Intact Blankets - -

Small Pieces 6% 0%

Transco RMI Large Pieces 4% 0%

Mineral Wool Fines 95% 95%

Fines 95% 95%

Cal-Sil Small Transport as Erosion Fines 16% 17%

Pieces Transport as Small Pieces 5% 0%

Fines 95% 95%

Asbestos Cal-Sil Small Transport as Erosion Fines 16% 17%

Pieces Transport as Small Pieces 5% 0%

Qualified Coatings Particulate 95% 95%

Unqualified Coatings Particulate 100% 100%

Latent Debris Particulate/Fiber 85% 85%

E3-61

Enclosure 3 Updated Final Responses to Generic Letter 2004-02 Table 3.e.6-22: Overall Transport Fractions for a Break in the SG Compartment in Loop B (PBN2)

Train A Train B Debris Type Debris Size Operational Operational Fines 95% 95%

Small Transport as Erosion Fines 8% 8%

Pieces Transport as Small Pieces 37% 17%

LDFG Large Transport as Erosion Fines 4% 4%

Pieces Transport as Large Pieces 2% 0%

Intact Blankets 0% 0%

Fines 95% 95%

Small Transport as Erosion Fines 8% 8%

Pieces Transport as Small Pieces 37% 17%

Nukon Large Transport as Erosion Fines 4% 4%

Pieces Transport as Large Pieces 2% 0%

Intact Blankets 0% 0%

Fines 95% 95%

Small Transport as Erosion Fines 8% 8%

Pieces Transport as Small Pieces 74% 74%

Temp-Mat Large Transport as Erosion Fines 4% 4%

Pieces Transport as Large Pieces 28% 28%

Intact Blankets 0% 0%

Small Pieces 11% 0%

Transco RMI Large Pieces 6% 0%

Mineral Wool Fines 95% 95%

Fines 95% 95%

Cal-Sil Small Transport as Erosion Fines 17% 17%

Pieces Transport as Small Pieces 9% 0%

Fines 95% 95%

Asbestos Cal-Sil Small Transport as Erosion Fines 17% 17%

Pieces Transport as Small Pieces 9% 0%

Qualified Coatings Particulate 95% 95%

Unqualified Coatings Particulate 100% 100%

Latent Debris Particulate/Fiber 85% 85%

E3-62

Enclosure 3 Updated Final Responses to Generic Letter 2004-02 Table 3.e.6-23: Overall Transport Fractions for a Reactor Cavity Break Loop A (PBN2)

Train A Train B Debris Type Debris Size Operational Operational Fines - -

Small Transport as Erosion Fines - -

Pieces Transport as Small Pieces - -

LDFG Large Transport as Erosion Fines - -

Pieces Transport as Large Pieces - -

Intact Blankets - -

Fines - -

Small Transport as Erosion Fines - -

Pieces Transport as Small Pieces - -

Nukon Large Transport as Erosion Fines - -

Pieces Transport as Large Pieces - -

Intact Blankets - -

Fines - -

Small Transport as Erosion Fines - -

Pieces Transport as Small Pieces - -

Temp-Mat Large Transport as Erosion Fines - -

Pieces Transport as Large Pieces - -

Intact Blankets - -

Small Pieces 6% 0%

Transco RMI Large Pieces 4% 0%

Small Pieces 3% 0%

Transco RMI in Cavity Large Pieces 0% 0%

Mineral Wool Fines 100% 100%

Fines 100% 100%

Cal-Sil Small Transport as Erosion Fines 17% 17%

Pieces Transport as Small Pieces 5% 0%

Fines 100% 100%

Cal-Sil in Cavity Small Transport as Erosion Fines 17% 17%

Pieces Transport as Small Pieces 3% 0%

Fines 100% 100%

Asbestos Cal-Sil Small Transport as Erosion Fines 17% 17%

Pieces Transport as Small Pieces 5% 0%

Fines 100% 100%

Asbestos Cal-Sil in Transport as Erosion Fines Small 17% 17%

Cavity Pieces Transport as Small Pieces 3% 0%

Qualified Coatings Particulate 100% 100%

Unqualified Coatings Particulate 100% 100%

Latent Debris Particulate/Fiber 100% 100%

E3-63

Enclosure 3 Updated Final Responses to Generic Letter 2004-02 Table 3.e.6-24: Overall Transport Fractions for a Reactor Cavity Break Loop B (PBN2)

Train A Train B Debris Type Debris Size Operational Operational Fines - -

Small Transport as Erosion Fines - -

Pieces Transport as Small Pieces - -

LDFG Large Transport as Erosion Fines - -

Pieces Transport as Large Pieces - -

Intact Blankets - -

Fines - -

Small Transport as Erosion Fines - -

Pieces Transport as Small Pieces - -

Nukon Large Transport as Erosion Fines - -

Pieces Transport as Large Pieces - -

Intact Blankets - -

Fines - -

Small Transport as Erosion Fines - -

Pieces Transport as Small Pieces - -

Temp-Mat Large Transport as Erosion Fines - -

Pieces Transport as Large Pieces - -

Intact Blankets - -

Small Pieces 12% 0%

Transco RMI Large Pieces 7% 0%

Transco RMI in Small Pieces 7% 0%

Cavity Large Pieces 0% 0%

Mineral Wool Fines 100% 100%

Fines 100% 100%

Cal-Sil Small Transport as Erosion Fines 17% 17%

Pieces Transport as Small Pieces 9% 0%

Fines 100% 100%

Cal-Sil in Cavity Small Transport as Erosion Fines 17% 17%

Pieces Transport as Small Pieces 6% 0%

Fines 100% 100%

Asbestos Cal-Sil Small Transport as Erosion Fines 17% 17%

Pieces Transport as Small Pieces 9% 0%

Fines 100% 100%

Asbestos Cal-Sil in Transport as Erosion Fines Small 17% 17%

Cavity Pieces Transport as Small Pieces 6% 0%

Qualified Coatings Particulate 100% 100%

Unqualified Coatings Particulate 100% 100%

Latent Debris Particulate/Fiber 100% 100%

E3-64

Enclosure 3 Updated Final Responses to Generic Letter 2004-02 Table 3.e.6-25: Overall Transport Fractions for a Pressurizer Compartment Break (PBN2)

Train A Train B Debris Type Debris Size Operational Operational Fines 95% 95%

Small Transport as Erosion Fines 9% 9%

Pieces Transport as Small Pieces 38% 17%

LDFG Large Transport as Erosion Fines 5% 5%

Pieces Transport as Large Pieces 2% 0%

Intact Blankets 0% 0%

Fines - -

Small Transport as Erosion Fines - -

Pieces Transport as Small Pieces - -

Nukon Large Transport as Erosion Fines - -

Pieces Transport as Large Pieces - -

Intact Blankets - -

Fines 95% 95%

Small Transport as Erosion Fines 9% 9%

Pieces Transport as Small Pieces 78% 78%

Temp-Mat Large Transport as Erosion Fines 5% 5%

Pieces Transport as Large Pieces 38% 38%

Intact Blankets 0% 0%

Small Pieces 11% 0%

Transco RMI Large Pieces 7% 0%

Mineral Wool Fines 95% 95%

Fines 95% 95%

Cal-Sil Small Transport as Erosion Fines 16% 16%

Pieces Transport as Small Pieces 9% 0%

Fines 95% 95%

Asbestos Cal-Sil Small Transport as Erosion Fines 16% 16%

Pieces Transport as Small Pieces 9% 0%

Qualified Coatings Particulate 95% 95%

Unqualified Coatings Particulate 100% 100%

Latent Debris Particulate/Fiber 85% 85%

E3-65

Enclosure 3 Updated Final Responses to Generic Letter 2004-02 The transported debris quantities for selected breaks are presented below for each unit. These transported debris quantities were calculated in NARWHAL for the single train failure case using the debris transport fractions provided in this section. Note that the debris quantities are shown for various debris groups defined in NARWHAL. The table below summarizes the individual debris types included in each debris group.

Table 3.e.6-26: Definition of Debris Groups Debris Group Represented Plant Debris Types Transported fines for LDFG, Nukon (Unit 2 only), latent Fiber Fine fiber and Temp-Mat, as well as fiber fines from erosion (in LDFG-equivalent ft3) of small and large pieces Mineral Wool Transported fine Mineral Wool debris 3

(in LDFG-equivalent ft )

Cal-Sil (in mass) Transported Cal-Sil and Asbestos Cal-Sil particulate Transported qualified, unqualified and ADQ epoxy Coating Particulate (ft3) coatings particulate Transported qualified, unqualified and ADQ epoxy Coating Particulate and coatings particulate, as well as transported ADQ epoxy Chips (ft3) flat fine and small chips Coating Particulate and Transported qualified, unqualified and ADQ epoxy Dirt/Dust (ft3) coatings particulate, as well as latent particulate The debris transport quantities for the most limiting fiber and Cal-Sil breaks for PBN1 and PBN2 are listed in Table 3.e.6-27 and Table 3.e.6-29, respectively. The quantities of debris transported for the four most limiting breaks that do not fail any of the strainer or core acceptance criteria are listed in Table 3.e.6-28 and Table 3.e.6-30 for PBN1 and PBN2, respectively. Note that the volumes for fine fiber and mineral wool are LDFG-equivalent volumes.

E3-66

Enclosure 3 Updated Final Responses to Generic Letter 2004-02 Table 3.e.6-27: Transported Debris Quantities of Worst Fiber and Cal-Sil Breaks at PBN1 RC-36-MRCL- RC-34-MRCL- RC-34-MRCL- RC-34-MRCL-Break Location BII-01 BI-03* BI-03* BI-02 Description Worst Fiber Break Worst Cal-Sil Break Break Size 31" 31" 31" 29" Break Type DEGB DEGB DEGB DEGB Fine Fiber (ft3) 22.70 22.84 22.84 21.88 Mineral Wool (ft3) 119.68 113.95 113.95 96.53 Cal-Sil (lbm) 400.49 657.32 657.32 553.40 Coating Particulate 13.15 13.10 13.10 13.02 (ft3)

Coating Particulate 14.67 14.62 14.62 14.54 and Chips (ft3)

Coating Particulate and Latent 13.80 13.75 13.75 13.66 Particulate (ft3)

• . Break RC-34-MRCL-BI-03 is the first worst Cal-Sil break and second worst fiber break.

Table 3.e.6-28: Transported Debris Quantities of Four Worst Breaks that Do Not Fail any Acceptance Criteria at PBN1 RC-36-MRCL- RC-34-MRCL- RC-36-MRCL- RC-36-MRCL-Break Location AII-02 AI-03 AIII-01 AIII-01A Description Worst Fiber Breaks Worst Cal-Sil Breaks Break Size 23 23 20 17 Break Type Partial Partial Partial Partial Fine Fiber (ft3) 16.80 15.95 15.38 14.69 Mineral Wool (ft3) 41.79 41.32 21.60 13.69 Cal-Sil (lbm) 28.59 47.86 335.82 271.46 Coating Particulate 12.18 12.07 12.00 11.99 (ft3)

Coating Particulate 13.70 13.59 13.52 13.51 and Chips (ft3)

Coating Particulate 12.82 12.72 12.65 12.64 and Dirt/Dust (ft3)

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02 Table 3.e.6-29: Transported Debris Quantities of Worst Fiber and Cal-Sil Breaks at PBN2 RC-36-MRCL- RC-36-MRCL- RC RC-34-MRCL-Break Location BII-01A BII-01R1 MRCL-AI-03 AI-04R1 Description Worst Fiber Breaks Worst Cal-Sil Breaks Break Size 31" 31" 29" 29" Break Type DEGB DEGB DEGB DEGB Fine Fiber (ft3) 87.72 86.03 56.05 60.98 Mineral Wool (ft3) 121.60 121.91 98.04 102.18 Cal-Sil (lbm) 211.55 208.35 810.09 797.19 Coating Particulate 13.96 13.97 13.60 13.57 (ft3)

Coating Particulate 15.48 15.49 15.12 15.09 and Chips (ft3)

Coating Particulate 14.60 14.61 14.24 14.21 and Dirt/Dust (ft3)

Table 3.e.6-30: Transported Debris Quantities of Four Worst Breaks that Do Not Fail any Acceptance Criteria at PBN2 RC-34-MRCL- RC-34-MRCL- RC-10-AC- RC-36-MRCL-Break Location BI-03 BI-03 2001-07 AII-05 Description Worst Fiber Breaks Worst Cal-Sil Breaks Break Size 20 20 8.75 26 Break Type Partial Partial Full Partial Fine Fiber (ft3) 47.25 49.62 10.00 19.41 Mineral Wool (ft3) 41.15 36.39 0.00 1.48 Cal-Sil (lbm) 50.09 52.30 187.27 180.04 Coating Particulate 12.15 12.35 11.95 12.20 (ft3)

Coating Particulate 13.67 13.87 13.47 13.72 and Chips (ft3)

Coating Particulate 12.79 12.99 12.59 12.84 and Dirt/Dust (ft3)

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02

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 through Figure 3.f.1-3 for ECCS and CSS schematics of PBN1.

Although the figures depict the PBN1 installation, they are also representative of the PBN2 systems.

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02 Figure 3.f.1-1: ECCS and CSS Schematic Diagram (1 of 3)

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02 Figure 3.f.1-2: ECCS and CSS Schematic Diagram (2 of 3)

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02 Figure 3.f.1-3: ECCS and CSS Schematic Diagram (3 of 3)

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02

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

Response to 3.f.2:

Table 3.f.2-1 summarizes the minimum submergence of the strainers at both PBN1 and PBN2. These strainer submergence values are based on sump water levels determined in a bounding hand calculation. See more details in the Response to 3.g.1.

Table 3.f.2-1: Minimum Strainer Submergence PBN1 Strainer PBN2 Strainer Break Break Submergence Submergence Size Elevation (ft) (ft)

SBLOCA Top of the pressurizer 0.17 0.19 Below the elevation of the SBLOCA 0.29 0.31 top of the hot leg nozzles LBLOCA Top of the pressurizer 0.17 0.19 Below the elevation of the LBLOCA 0.72 0.74 top of the hot leg nozzles

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

Response to 3.f.3:

Vortex testing was performed on a PBN prototypical strainer module to observe the size, shape, and location of vortices that may develop at different debris loads and strainer submergence levels. The vortex tests were performed during the full-load head loss test described in the Response to 3.f.4. Both clean screen and debris laden vortex tests were performed.

Prior to any debris additions, a vortexing check was performed on the clean strainer.

No vortexing was observed with 2 inches of strainer submergence and an approach velocity of 0.00267 ft/s.

Throughout the duration of the test sequence, there were two instances of vortex formation. During the drain operation performed prior to the fourth conventional debris addition of the Full Debris Load (FDL) Test 1 (FDL1), a full-core vortex developed at a strainer submergence of approximately 3 inches. As shown in Figure 3.f.3-1, the water level was then increased. Vortex formation ceased at a submergence of approximately 4 inches. During the FDL Test 2 (FDL2), a small vortex formed above the test strainer after the final flow sweep was conducted prior to drain down. When the debris laden vortex tests were performed, the test strainer approach velocity was maintained at 0.00267 ft/s or slightly higher.

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02 Figure 3.f.3-1: Minimum Submergence Level for Vortexing during Full Debris Load Test 1 The response to 3.g.1 shows a minimum strainer submergence of 0.72 ft (or 8.64 inches) for LBLOCAs on the main loop for PBN1 (0.74 ft for PBN2). Vortexing was only seen at a strainer submergence of less than 4 inches during the drain operation in the FDL1 test and the flow sweep after the FDL2 test. Therefore, it is reasonable to conclude that vortexing will not cause air entrainment for the limiting breaks, which have a submergence of greater than 4 inches.

The Response to 3.g.1 shows a smaller minimum strainer submergence for LOCAs at the top of the pressurizer (0.17 ft or ~ 2 inches). These breaks produce much less debris than those at the primary loop elevation. Additionally, the minimum water levels were calculated at the start of sump recirculation when the strainer is mostly clear of debris. The breaks at the top of the pressurizer would be at the minimum strainer submergence only momentarily while the pool continues to rise due to CS injection from the RWST. Therefore, it is reasonable to conclude that vortexing would not occur based on the clean screen vortex evaluation at 2 inches of submergence which did not have any vortex formation.

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02

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 the plant. Different test cases were performed with the thin bed (TB) and full debris load (FDL) protocols, following the 2008 NRC Staff Review Guidance (Reference 3). Note that two separate test programs were conducted in 2015 and 2016, respectively. The discussion in this section is based on the 2016 test program, unless otherwise noted.

For PBN, target values of debris were established that ranged from the smallest breaks to the largest breaks including both DEGBs and partial breaks. The debris quantities to be tested were then derived from these target values. The testing sequence was performed following a test for success strategy with a total of 4 tests, as summarized below. These tests were applied in the risk quantification for both PBN units.

x The FDL1 test targeted a low Cal-Sil, high fiber trajectory.

x The FDL2 test targeted a higher Cal-Sil to fiber ratio than the FDL1 test.

x The confirmatory test (CT) targeted the highest Cal-Sil to fiber ratio.

x The TB test simulated a dense debris bed with high particulate to fiber ratio.

Test Setup The PBN sump strainer system consists of two independent module assemblies of passive strainer disks each attached to their own suction pipe, supplying flow to one ECCS and CS train. The ECCS and CSS are not independent of each other downstream of the strainer. During recirculation, the CS pump in the associated train will be supplied by the RHR pump in that train, as will the SI pump during simultaneous upper plenum and cold leg injection, after the CS pump is stopped. Each strainer assembly consists of a suction pipe and 14 strainer disk modules. The strainer assembly is flow-controlled such that the flow rate through each strainer assembly is uniform. Perforated spacers maintain the horizontal separation between adjacent disks within a strainer module such that a 1 gap exists between the perforated plates of adjacent disks.

Figures 3.f.4-1 and 3.f.4-2 show the test strainer in the test tank as well as the debris introduction section of the test tank, including the mixing lines and hopper inlet. Note that these figures are from the 2015 test program, for which the test set-up was very similar to the 2016 testing. The test strainer assembly consisted of one prototypical 10-disk strainer module with a flow-controlled suction pipe passing through the center (core tube). The total surface area of the test strainer was 136 ft2. To simulate the E3-75

Enclosure 3 Updated Final Responses to Generic Letter 2004-02 module-to-module clearance, the width of the tank was designed to model the active strainer module length and the 5-inch module-to-module clearance in the plant.

Figure 3.f.4-1: Test Tank and Strainer Figure 3.f.4-2: Mixing Lines and Hopper Inlet in the Test Tank E3-76

Enclosure 3 Updated Final Responses to Generic Letter 2004-02 A schematic piping diagram of the test loop is provided in Figure 3.f.4-3. Downstream of the main recirculation pump, a small portion of the flow can be directed through a heat exchanger to control the test loop temperature. A large majority of the flow then passes directly through the mixing nozzle configuration, placed at the upstream end of the test tank (see response to 3.f.12). The remainder of flow that does not travel through the mixing nozzles is divided and the two streams pass through the debris introduction hopper and transition tank respectively. The continuously mixed transition tank was brought online during conventional and chemical precipitate debris introduction to increase the test loop water capacity and decrease the amount of draining required during testing. Flow directed to the debris introduction hopper supplies turbulence through the bottom of the hopper to encourage mixing of the debris slurry. The discharge of the hopper gravity drains into the test tank. The filter bag housings were used only during pre-test cleaning, and were isolated and bypassed during head loss testing.

Figure 3.f.4-3: Piping Diagram of Head Loss Test Loop Test Parameters and Scaling The test strainer replicated all hydraulic dimensions of the plant strainer except for the number of strainer modules. The test debris quantities and test flow rate were scaled from plant values based on the ratio of test strainer surface area to the plant strainer surface area (1,754.6 ft2). This strainer surface area was determined by deducting 150 ft2 from the total surface area of each PBN strainer (1,904.6 ft2) to account for blockage by miscellaneous debris. For a total plant strainer flow rate of 2,100 gpm, the test flow rate was determined to be 162.8 gpm, which corresponds to a strainer approach velocity of 0.00267 ft/s.

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02 Debris Materials and Preparation Conventional debris consists of fiber and particulate debris from failed insulation and coatings, and latent materials that could be transported to the sump strainers following a LOCA. PBN has four types of fibrous debris: LDFG, Mineral Wool, Temp-Mat, and latent fiber. Nukon and Mineral Wool were the only two fibrous debris types used during testing. Temp-Mat fines were not used since Temp-Mat is not generated for approximately 95% of all postulated breaks.

Nukon fines were used as a surrogate to model latent fiber on a basis of similar macroscopic density and characteristic fiber size. Heat treated Nukon sheets were procured and processed right before each test. Some of the Mineral Wool used during testing was heat treated by the testing vendor prior to processing. The required burn out gradient reached approximately half way through the blanket.

Particulate debris sources for PBN include Cal-Sil, asbestos Cal-Sil, qualified and unqualified coatings, ADQ epoxy coatings, and latent particulate. Pulverized Cal-Sil was purchased and used during the test for Cal-Sil and asbestos Cal-Sil debris. Due to their similar characteristic sizes and microscopic densities, silica flour, with a material density of 165.4 lbm/ft3 and a median size of approximately 13.5 microns was used as a surrogate for qualified coatings, unqualified coatings, and the fine particle portion of the ADQ epoxy coatings. Pant chips were used as a surrogate for the flat small chip portion of ADQ epoxy coatings. Pressure washed paint chips, with a nominal size of approximately 0.125 in, were used as a surrogate for the flat fine chip portion of the ADQ epoxy coatings. The material density of the paint chips was 89.3 lbm/ft3. Latent particulate was modeled with PCI Dirt and Dust Mix which was procured from PCI and used without additional processing.

Preparation of Nukon and Mineral Wool fiber started by cutting the insulation sheets into approximately 2 by 2 cubes. The base material for both types of debris, ready for fiber fines preparation, is shown in Figure 3.f.4-4.

Figure 3.f.4-4: Base Debris Material before Pressure Wash E3-78

Enclosure 3 Updated Final Responses to Generic Letter 2004-02 The required quantity of debris was weighed out per the debris batching schedule.

The debris was then wetted in preheated test water and processed into fines following the method developed by NEI (Reference 20). The processing involves pressure washing the debris using a nominal 1,500 psi pressure washer with nozzles that produce a fan-type flow distribution. The nozzle position within the preparation vessel and the amount of time the spray was applied were controlled between debris batches.

Prepared fiber fines consisted of primarily Class 2 fibers as defined in NUREG/CR-6224 (Reference 21, Table B-3). Figure 3.f.4-5 shows pictures of the processed fiber samples inside an acrylic column on top of a light table.

Figure 3.f.4-5: Prepared Nukon and Mineral Wool Debris Silica flour was used as a surrogate for qualified and unqualified coatings debris, and ADQ epoxy coatings particulate debris on an equal volume basis. The required quantity of silica flour for a debris batch was first weighed out before being wetted in test water. For the FDL tests, the wetted silica flour was combined with the prepared fibrous debris slurry to form a homogeneous suspension. For the TB test, silica flour was mixed in barrels of heated test water and sufficiently diluted to allow for direct introduction through the debris hopper.

Cal-Sil was prepared in a similar manner as silica flour. For the FDL tests, the desired amount of Cal-Sil was weighed out, wetted with heated test water, and combined with the fibrous debris slurry. For the TB test, Cal-Sil was diluted with sufficient test water to allow for direct introduction through the debris hopper.

The PCI Dirt and Dust Mix, which was used as a surrogate for latent particulate debris, did not require processing. It was introduced in its dry form and sprinkled directly into the test tank upstream of the strainer.

The paint chips were wetted down with test water and repeatedly mixed to minimize the potential for flotation. For the FDL tests, the paint chips were combined with the homogenous debris slurry prior to introduction. For the TB test, the wetted paint chips were added directly to the debris introduction hopper prior to the fibrous debris. It should be mentioned that the large flat chips and curled chips from the ADQ epoxy coatings were not introduced during the 2016 test program as it was demonstrated E3-79

Enclosure 3 Updated Final Responses to Generic Letter 2004-02 during the 2015 test program that these chips would not transport to the strainer even with agitation.

Sodium aluminum silicate (SAS) was used as the chemical debris surrogate for the head loss testing. The chemical debris was prepared in accordance with and met the acceptance criteria specified in WCAP-16530-NP-A (Reference 22). For chemical precipitate generation, a chemical salt and a base were weighed out to provide a specific chemical concentration in a measured volume of tap water. The prepared chemical debris was continuously mixed until it was added to the test tank. 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 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> from the time the surrogate was to be used. The specifics of the chemical surrogates used during testing are described in the Response to 3.o.2.12.

Debris Introduction Full Debris Load and Confirmatory Tests For the FDL and Confirmatory tests, the prepared silica flour, Cal-Sil, paint chips, pressure washed paint chips, and fibrous debris for a given batch were combined in barrels prior to addition into the test loop. The debris was agitated into a homogeneous mixture prior to introduction, and was continuously mixed during introduction to prevent agglomeration and to maintain the concentration as constant as practical. The homogeneous mixture of debris was transferred to the hopper via 5 gallon buckets. Debris additions to the test tank were performed utilizing the debris hopper, which mixed the debris slurry with test loop water before transporting the debris to the upstream end of the test tank. The flow pattern in the hopper caused the debris to be held in suspension, which prevented agglomeration prior to adding the debris to the tank. The dirt and dust for the given batch was added directly to the test tank. To achieve the desired transport of debris in the test tank, five mixing nozzles were implemented. These mixing nozzles maintained turbulence in the test tank to prevent debris from settling but were demonstrated to be sufficiently far from the strainer to prevent any disruption of the debris bed on the strainer.

Thin Bed Test During the TB test, the particulate debris was added before any fibrous debris. The prepared silica flour was added first, followed by the introduction of Cal-Sil, pressure washed paint chips, paint chips, and dirt and dust. All of these particulate debris types were introduced through the hopper with the exception of dirt and dust, which was sprinkled directly into the test tank in its dry form. All particulate debris was added in quick succession, and no fiber was added to the test until all particulate was introduced. After all the particulate debris was added to the test tank, homogeneous mixed batches of Nukon and Mineral Wool fines were added through the debris hopper. Note that the fibrous debris was continuously mixed during introduction to E3-80

Enclosure 3 Updated Final Responses to Generic Letter 2004-02 prevent agglomeration and to maintain the concentration as constant as practical. The size of each fiber batch was equivalent to a 1/16 theoretical uniform debris bed thickness for the test strainer. The mixing nozzles were utilized for the TB test as well to prevent settling.

After conventional debris introduction was completed for the FDL, TB, and Confirmatory tests, the SAS chemical precipitate debris was added to the test tank in batches. Chemical debris was pumped from the preparation tank into the test tank through a chemical introduction line.

Head Loss Test Cases and Results Four head loss tests were performed for PBN: two FDL tests, one TB test, and the Confirmatory Test.

PBN FDL Test 1 The total conventional debris loads at the test scale for the FDL1 Test are provided in the table below. The peak conventional debris head loss observed for this test is shown in Table 3.f.4-9.

Table 3.f.4-1: Conventional Debris Loads for FDL1 Test Pressure Mineral Dirt/ Silica Paint Nukon Cal-Sil Washed Wool Dust Flour Chips Paint Chips (g) (g) (g) (g) (lbm) (g) (g) 4394.5 3538.9 6503.1 3815.8 159 973.3 3798.8 After all conventional debris was added, the head loss had stabilized, and a flow sweep had been performed, chemical precipitate debris was added to the test tank.

The volumes of the prepared SAS solution added to the test tank during the FDL1 Test are summarized in the table below. Note that the concentration of the SAS solution is 9.97 g/L. The total mass of SAS used in the FDL1 test is 7.1 kg.

Table 3.f.4-2: Chemical Debris Batches Added for the FDL1 Test Batch # Volume of SAS Solution (gal) 1 80 2 107 Total 187 Figure 3.f.4-6 and Figure 3.f.4-7 show plots of raw head loss test data for the PBN FDL1 Test with time to identify the key testing activities. Note that the flow rates shown in these figures are at the test scale and the head loss values have not been adjusted to subtract the test strainers clean screen head loss. The clean screen head loss for the FDL1 Test was 0.03 psi.

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02 Figure 3.f.4-6: FDL1 Test Conventional Debris Timeline E3-82

Enclosure 3 Updated Final Responses to Generic Letter 2004-02 Figure 3.f.4-7: FDL1 Test Chemical Debris Timeline PBN FDL Test 2 The conventional debris loads at the test scale for the FDL2 Test are provided in the table below. The peak conventional debris head loss observed for this test is shown in Table 3.f.4-9.

Table 3.f.4-3: Conventional Debris Loads for FDL2 Test Pressure Mineral Dirt/ Paint Nukon Cal-Sil Silica Flour Washed Wool Dust Chips Paint Chips (g) (g) (g) (g) (lbm) (g) (g) 2486.4 2435.6 10808.8 3052.8 123.6 778.7 3039.2 After all conventional debris was added, the head loss had stabilized, and a flow sweep had been performed, chemical precipitate debris was added to the test tank.

The volumes of the prepared SAS solution added to the test tank during the FDL2 Test are summarized in the table below. The total mass of SAS is 6.0 kg.

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02 Table 3.f.4-4: Chemical Debris Batches Added for the FDL2 Test Batch # Volume of SAS Solution (gal) 1 60 2 100 Total 160 Figure 3.f.4-8 and Figure 3.f.4-9 show plots of raw head loss test data for the FDL2 Test with time to identify the key testing activities. Note that the flow rates shown in these figures are at the test scale and the head loss values have not been adjusted to subtract the test strainers clean screen head loss. The clean screen head loss for the FDL2 Test was 0.03 psi.

Figure 3.f.4-8: FDL2 Test Conventional Debris Timeline E3-84

Enclosure 3 Updated Final Responses to Generic Letter 2004-02 Figure 3.f.4-9: FDL2 Test Chemical Debris Timeline PBN Thin-Bed Test The conventional debris loads at the test scale for the PBN TB Test are summarized in the table below. Six batches of fiber fines were introduced to the test tank during the TB test, which resulted in a cumulative theoretical uniform debris bed thickness of approximately 3/8. The peak conventional debris head loss observed for this test is shown in Table 3.f.4-9.

Table 3.f.4-5: Conventional Debris Loads for the TB Test Pressure Mineral Silica Paint Nukon Cal-Sil Dirt/ Dust Washed Wool Flour Chips Paint Chips (g) (g) (g) (g) (lbm) (g) (g) 2119.1 2075.9 13517.1 3815.9 154.2 973.3 3799 After all conventional debris was added, the head loss had stabilized, and a flow sweep had been performed, chemical precipitate debris was added to the test tank.

The volumes of the prepared SAS solution added to the test tank during the TB Test are summarized in the table below. The total mass of SAS used is 6.0 kg.

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02 Table 3.f.4-6: Chemical Debris Batches Added for the TB Test Batch # Volume of SAS Solution (gal) 1 100 2 60 Total 160 Figure 3.f.4-10 and Figure 3.f.4-11 show plots of raw head loss test data for the TB test with time to demonstrate the key testing activities. Note that the flow rates shown in these figures are at the test scale and the head values have not been adjusted to subtract the test strainers clean screen head loss. The clean screen head loss for the TB Test was 0.03 psi.

Figure 3.f.4-10: TB Test Conventional Debris Timeline E3-86

Enclosure 3 Updated Final Responses to Generic Letter 2004-02 Figure 3.f.4-11: TB Test Chemical Debris Timeline PBN Confirmatory Test The conventional debris loads at the test scale for the PBN CT Test are summarized in the table below. The conventional debris head loss observed for this test is shown in Table 3.f.4-9.

Table 3.f.4-7: Conventional Debris Loads for CT Test Pressure Mineral Dirt/ Silica Paint Nukon Cal-Sil Washed Wool Dust Flour Chips Paint Chips (g) (g) (g) (g) (lbm) (g) (g) 978.77 1779.16 10487.6 3816 153.93 973.28 3798.8 After all conventional debris was added, the head loss had stabilized, and a flow sweep had been performed, chemical precipitate debris was added to the test tank.

The volumes of the prepared SAS solution added to the test tank during the CT Test are summarized in the table below. The total mass of SAS used is 6.0 kg.

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02 Table 3.f.4-8: Chemical Debris Batches Added for the CT Test Batch # Volume of SAS Solution (gal) 1 100 2 60 Total 160 Figure 3.f.4-12 and Figure 3.f.4-13 show plots of raw head loss test data for the CT Test with time to demonstrate the key testing activities. Note that the flow rates shown in this figure are at the test scale and the head loss values have not been adjusted to subtract the test strainers clean screen head loss. The clean screen head loss for the CT Test was 0.03 psi.

Figure 3.f.4-12: CT Test Conventional Debris Timeline E3-88

Enclosure 3 Updated Final Responses to Generic Letter 2004-02 Figure 3.f.4-13: CT Test Chemical Debris Timeline Summary of PBN Head Loss Test Data A summary of the debris head loss results from the PBN tests are provided in the table below. As discussed in the response to 3.f.7, the maximum conventional and chemical debris head losses of the four tests are used to derive the head loss lookup tables as inputs to the risk quantification. Compared with the previous submittal, the stabilized conventional debris head losses are updated to the values measured right before start of chemical debris introduction. This is necessary for determining the head loss increase resulting from the chemical debris addition.

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02 Table 3.f.4-9: Summary of Debris Head Loss Results Test Flow Rate Debris Head Loss Temperature Test Point (at plant scale)

(psi) (°F)

(gpm)

FDL1 Test Conventional Debris 165.7 1.135 119.7 Max Head Loss (2,138)

Conventional Debris 164.3 1.113 120.5 Stable Head Loss (2,120)

Aluminum Precipitate 164 1.84 120.1 Max Head Loss (2,116)

FDL2 Test Conventional Debris 165.5 1.592 123 Max Head Loss (2,135)

Conventional Debris 164.4 1.208 120.8 Stable Head Loss (2,121)

Aluminum Precipitate 162.7 2.142 99.9 Max Head Loss (2,099)

TB Test Conventional Debris 164.9 0.525 120 Max Head Loss (2,128)

Conventional Debris 164.0 0.497 120.2 Stable Head Loss (2,116)

Aluminum Precipitate 162.6 1.402 121.7 Max Head Loss (2,098)

CT Test Conventional Debris 160.2 0.695 120.1 Max Head Loss (2,067)

Conventional Debris 164.5 0.668 120.3 Stable Head Loss (2,122)

Aluminum Precipitate 159.8 1.352 119.8 Max Head Loss (2,062)

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 impact on the plant strainer of the volume of debris arriving at the strainer is evaluated for each postulated break in NARWHAL using a risk-informed evaluation.

The head loss testing described in the Response to 3.f.4 provides the basis for the head loss effects on the strainer in the risk-informed evaluation.

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02 As discussed in the Response to 3.f.4, the head loss tests used a test strainer that is prototypical to the plant strainer design. Additionally, the test debris loads were scaled based on the ratio of the test strainer surface area and the plants net strainer surface area. The arrangement of the test strainer with respect to the test tank is representative of the module-to-module spacing in the plant strainer. As a result, the flow profile approaching the test strainer is comparable to that of the plant strainer.

The debris quantities used in the 2016 test program do not bound the debris loads of some of the postulated breaks. To account for this in the risk quantification, debris limits were implemented in the NARWHAL analysis and were defined as maximum allowable debris quantities for each strainer in operation. When analyzing a given break, NARWHAL compares the debris loads of one strainer with the debris limits during each time step. If any of the debris limits were exceeded, a failure was recorded for that break.

The debris limits were taken to be the maximum quantity of conventional and chemical debris that was tested in 2016 (see the Responses to 3.f.4). Instead of defining a limit for each type of debris, various debris groups were used in NARWHAL (see the Response to 3.e.6), for which the debris limits were given. Table 3.f.5-1 shows the debris limits for each of the debris groups at the test scale for one strainer. The Response to 3.f.10 explains how the raw tested debris loads were converted and grouped to build the head loss lookup table. The maximum tested quantity of each debris group in that table is taken to be the debris limit.

Table 3.f.5-1: Debris Limit Failure Criteria Debris Limit Debris Type Unit at Test Scale Fiberglass and Latent Fiber ft3 4.037 Mineral Wool ft3 3.251 Cal-Sil lbm 29.8 Coatings Particulate ft3 0.961 Coatings Particulate and Chips ft3 1.079 Coatings Particulate and Latent Particulate ft3 1.011 SAS lbm 15.65

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 PBN head loss testing E3-91

Enclosure 3 Updated Final Responses to Generic Letter 2004-02 included a test for thin-bed effects. During this test, the particulate debris was added into the test tank first, followed by six batches of fiber fines with a batch size equivalent to a 1/16-inch theoretical uniform Nukon-equivalent bed thickness (see Table 3.f.4-1).

The total theoretical fiber bed thickness after all of the fiber batches was 3/8. This batching schedule allowed the formation of a debris bed with high particulate to fiber ratios. As was demonstrated by testing, the thin-bed effect was not observed.

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

Response to 3.f.7:

As stated previously, strainer failures associated with the strainer head loss were analyzed in NARWHAL. Several failure criteria were evaluated by NARWHAL for each break: 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 strainers/pumps would result in a failure of the ECCS. Each of the failure criteria was evaluated at each time step within the NARWHAL model to determine if an ECCS failure would occur.

Strainer Structural Margin Limits The structural design differential pressure for the PBN sump strainers is 10.0 ft. The total head loss across each strainer 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.

Unsubmerged Strainer Limits If the strainers are partially submerged, NARWHAL would record a strainer failure if the total strainer head loss is greater than half of the submerged strainer height per RG 1.82 (Reference 23). For PBN, it was demonstrated that the strainers for both units are fully submerged during recirculation for all breaks (see Section 3.g.1).

Void Fraction Limits A pump failure due to degasification was recorded if the steady state gas void fraction at the RHR pump suction was greater than 2% by volume. Void fraction is evaluated at the middle height of the strainer disks and voids are assumed to transport intact to the RHR pump suction without crediting compression. Note that no containment accident pressure was credited for the degasification evaluation.

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02 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 calculated at the top of the strainer, containment pressure and total strainer head loss. A small amount of containment accident pressure was credited to mitigate flashing failures, see the Response to 3.f.14 for additional information.

Pump NPSH Limits A pump failure was recorded if the total strainer head loss exceeded the RHR pump NPSH margin. When calculating pump NPSH margin, no strainer head loss was accounted for. Note that because the SI and CS pumps take suction from the RHR pumps during recirculation, only the NPSH margins of the RHR pumps need to be considered. The RHR pump NPSH margins were calculated in NARWHAL. See the Response to 3.g.16 for details of pump NPSH margin evaluation.

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

Response to 3.f.8:

Vortexing Evaluation 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.

All vortex evaluations used a strainer approach velocity of 0.00267 ft/s, which is based on a conservatively smaller strainer surface area by accounting for a sacrificial area of 150 ft2 for miscellaneous debris. The actual reduction in strainer surface area due to blockage by the miscellaneous debris is less than 100 ft2 at PBN1 (120 ft2 of miscellaneous debris with 25% overlap) and less than 115 ft2 at PBN2 (152 ft2 of miscellaneous debris with 25% overlap). This is conservative as a vortex is more prone to form at higher velocities.

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 had 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. Also, the depth of the containment sump pool continues to increase following the start of sump recirculation (due to injection of the CS pumps from the RWST). Therefore, the submergence levels at which vortex formation was evaluated are conservatively low.

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02 Strainer Head Loss The quantity of latent debris used to determine the strainer head loss is 150 lbm, but the actual plant latent debris loads are much lower for both units. The actual debris loads are well below the quantity used to determine the strainer head loss. Similarly, a sacrificial strainer area of 150 ft2 was used when determining the testing parameters. In reality, the reduction in strainer surface area due to blockage of miscellaneous debris is much less for both units, as stated above.

The strainer flow rates used to analyze strainer head loss are conservatively higher than the plant flow rate. The clean strainer head loss was evaluated at a flow rate of 2,200 gpm (see the Response to 3.f.9). The debris head losses were evaluated at a flow rate of 2,100 gpm (see the Response to 3.g.1). Both flow rates are greater than the maximum expected flow rate of 2,088 gpm from hydraulic modeling.

When correcting the chemical debris head loss from the test conditions (e.g., water temperature and strainer approach velocity) to plant conditions, the scaling factor was calculated from flow sweep data of all four tests, and the highest value was conservatively applied. See the Response to 3.f.10 for additional information.

The rule-based approach for determining strainer debris head loss (see the Response to 3.f.10) used a lookup table based on the peak head loss measured during each test after the full load of debris is added to the test tank. As a result, all postulated breaks that result in strainer debris loads bounded by a test have the peak head loss applied. This is conservative because the actual debris loads for some of the breaks may be much lower than those tested.

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 for PBN was calculated by the strainer vendor. Clean strainer head loss test data, from a generic (non-plant specific) PCI prototype, was curve fit to a second-order polynomial function of the strainers core tube exit velocity.

The function was used to calculate the head loss for the PBN strainer disks using the PBN core tube exit velocity. It should be noted that this test performed by the strainer vendor was not part of the debris laden head loss test program described in the Response to 3.f. Since the tested PCI prototype strainer has differences from that installed in PBN, adjustments were made to account for the physical differences between the two designs.

The PCI prototype clean strainer testing used an approach velocity higher than that of the PBN strainer design. Since head loss increases with approach velocity, the head loss through the PCI prototype strainers perforated plate was expected to be greater than that through the PBN strainer perforated plates. Therefore, for conservatism, no E3-94

Enclosure 3 Updated Final Responses to Generic Letter 2004-02 adjustment was made to the perforated plate head loss calculated from the PCI prototype test data.

The PCI prototype strainer had a core tube length of 54 inches, but the PBN strainer has core tube length of 279.44 inches. The additional head loss due to longer length was calculated using the Darcy-Weisbach equation.

The Darcy-Weisbach equation, with head loss coefficients from standard industry handbooks, was also used to model the module-to-module transition head loss, and head losses inside the attached pipe and fittings. The head loss from the attached pipe to containment outlet was also calculated.

Finally, the head loss was calculated from internal flow restrictions inside the disks caused by the reinforcing wires in the disks. The internal flow restrictions were modeled as an orifice.

The clean strainer head loss was determined to be 0.56 ft at 212qF (0.24 psi) for a total strainer flow rate of 2,200 gpm.

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

Response to 3.f.10:

Strainer head loss analysis was performed for each postulated break in NARWHAL to determine which breaks result in strainer failures due to exceeding head loss based criteria. Such failures contribute to the overall risk quantification (see Enclosure 4).

For each break, the total strainer head loss is calculated as the sum of the clean strainer head loss, the conventional debris head loss, and the chemical head loss (if applicable). The conventional and chemical head losses were based on PBN-specific head loss test results that were corrected from the test conditions (i.e., strainer approach velocity and water temperature) to plant conditions. Strainer head loss results for select breaks are shown as examples in the Response to 3.g.16.

Clean Strainer Head Loss The bounding clean strainer head loss of 0.56 ft calculated at 2,200 gpm and 212qF was used for all cases in NARWHAL. No flow or temperature correction was applied to the clean strainer head loss. The clean strainer flow sweep data obtained during the PBN head loss tests showed that the flow condition through the clean strainer is mostly turbulent. As a result, the clean strainer head loss varies mostly with water density. For the sump temperature range of interest, water density varies little with temperature. As a result, the increase in clean strainer head loss due to a reduction in temperature is negligible.

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02 Conventional Debris Head Loss NARWHAL uses a rule-based approach to calculate debris head loss based on the results of PBN head loss tests. Table 3.f.10-11 shows a lookup table with the tested debris loads at the test scale and the peak head loss recorded for each test. The debris loads shown in the table were converted from the raw test data shown in the Response to 3.f.4 and were combined as necessary for the debris groups defined in the Response to 3.e.6. The rows are sorted in the order of increasing head loss.

Note that NARWHAL analyzed time dependent debris transport. As a result, the debris load on an operating strainer increases gradually over time. For a given time step, NARWHAL scales the plant strainer debris load to the test scale based on the active plant strainer surface area before using it to determine the conventional debris head loss. The comparison begins with the thin bed test. If any one or more strainer debris quantities are not bounded by the tested debris loads, it moves onto the next row.

When a test is found that bounds all debris types, the conventional head lost from that test is used. If no tests are found to bound the plant debris loads, a failure of the strainer is recorded for that break.

Table 3.f.10-1: Conventional Debris Head Loss Lookup Table Coating Coating Fine Mineral Coating Peak Conv Cal-Sil Particulate Particulate &

Test Fiber* Wool Particulate Head Loss

& Chips Dirt/ Dust ft3 ft3 lbm ft3 ft3 ft3 ft TB 1.947 1.907 29.800 0.932 1.050 0.982 1.221 CT 0.899 1.634 23.121 0.931 1.048 0.981 1.617 FDL1 4.037 3.251 14.337 0.961 1.079 1.011 2.640 FDL2 2.284 2.237 23.829 0.747 0.842 0.787 3.704

• This applies to fine LDFG, Temp-Mat, Nukon, and latent fiber debris, excluding mineral wool.

Chemical Head Loss A similar but simplified approach was used to determine the chemical head loss. The table below shows the maximum head loss increase due to addition of chemical debris for each of the four head loss tests. This was calculated by subtracting the stabilized conventional debris head loss from the measured peak head loss recorded after chemical debris addition (see Table 3.f.4-9). As shown in the table, the maximum quantity of tested SAS was used in the FDL1 test (7.1 kg or 15.65 lbm) but the maximum increase in head loss due to chemical debris was seen in the FDL2 test (2.173 ft). For conservatism, any breaks with a chemical debris load less than 15.65 lbm at test scale has a chemical debris head loss of 2.173 ft. For breaks that have a chemical debris load greater than 15.65 lbm at test scale, a strainer failure is recorded for that break (Section 3.f.5).

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02 Table 3.f.10-2: Chemical Debris Head Losses from Testing HL Increase due to Test SAS (lbm)

Chemical Debris (ft)

TB 13.23 2.105 CT 13.23 1.591 FDL1 15.65 1.691 FDL2 13.23 2.173 Debris Head Loss Correction A head loss correction factor was implemented into NARWHAL to scale the measured head losses from test conditions (i.e., the strainer approach velocity and water temperature) to plant conditions. During each time step for which conventional and chemical head losses are evaluated, the head losses obtained from the lookup tables above are adjusted based on the plant flow rate through the strainer and the pool temperature for the given time step. The correction was performed based on the debris bed characteristics obtained through flow sweeps conducted during head loss tests. For the 2016 test program, flow sweeps were performed during each of the four tests for both the conventional debris bed and after chemical debris was added.

The table below shows the flow sweep data taken after addition of all conventional debris during the FDL1 test. The first three columns are the raw data. The approach velocity is calculated from the test flow rate using the test strainer surface area (see the Response to 3.f.4). The head loss in the last column is converted from the raw values in psi using the water density (61.89 lbm/ft3) at the average flow sweep temperature of 120.53qF. The water viscosity at this temperature was determined to be 0.000375 lbm/(s-ft).

Table 3.f.10-3: Flow Sweep Data for Conventional Debris Head Loss (FDL1 Test)

Test Flow Head Approach Head T (qF)

Rate (gpm) Loss (psi) Velocity (ft/s) Loss (ft) 120.6 164.3 1.114 0.00269 2.59 120.8 145.6 0.929 0.00239 2.16 120.6 131.8 0.798 0.00216 1.86 120.4 115.6 0.653 0.00189 1.52 120.3 101.1 0.537 0.00166 1.25 120.5 168.4 1.099 0.00276 2.56 The flow sweep data of each test was used to derive a scaling factor. Multiplying the measured debris head loss from testing by this scaling factor adjusts the measured debris head loss to the plant conditions of interest. The equation for calculating the head loss scaling factor XHL in NARWHAL is shown below. In this equation, a, b and PHL are constants derived from flow sweep data and are inputs into NARWHAL.

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02 Parameters P and U are water viscosity and density at the plant sump temperature of a given time step, and vstrainer is the strainer approach velocity of the same time step.

Where:

XHL = Head loss scaling factor a, b = Parameters derived from flow sweep data

= Water viscosity at plant condition of interest (lbm/(ft-s))

= Water density at plant condition of interest (lbm/ft3)

PHL = Head Loss from flow sweep at the test approach velocity and temperature (ft)

An example for deriving the parameters a, b and PHL are shown below using the data given in Table 3.f.10-3 above.

The approach velocity and head loss in the table are plotted in Figure 3.f.10-1. The data was curve fit to a second order polynomial. Note that the curve is forced through the origin because the head loss would be 0 ft at an approach velocity of 0 ft/s.

3.0 y = 168,229.84x2 + 489.78x 2.5 2.0 Head Loss (ft) 1.5 1.0 0.5 0.0 0 0.0005 0.001 0.0015 0.002 0.0025 0.003 Average Approach Velocity (ft/s)

Figure 3.f.10-1: Curve Fit of Conventional Debris Flow Sweep Data of FDL1 Test E3-98

Enclosure 3 Updated Final Responses to Generic Letter 2004-02 Using the coefficients of the curve fit shown in the figure above, and water density 61.89 lbm/ft3 and viscosity 0.000375 lbm/(ft-s) at the average flow sweep temperature of 120.53qF, the a and b parameters are calculated as follows:













PHL is taken to be the head loss measured during the flow sweep at the test approach velocity and temperature, 2.59 ft, as shown in Table 3.f.10-1.

The same approach is used to calculate the head loss correction parameters for FDL1 chemical debris flow sweep, and the conventional and chemical debris flow sweeps of the three other tests: FDL2, TB and CT. Table 3.f.10-10 shows the resulting correction parameters as inputs into NARWHAL.

Table 3.f.10-4: Head Loss Correction Parameters Debris Head Test a b PHL (ft)

Loss Type FDL1 Conventional 1306103.95 2718.22 2.59 FDL1 Chemical 1913299.26 3916.00 4.15 CT Conventional 479702.18 2235.38 1.55 CT Chemical 856890.73 4297.01 2.88 FDL2 Conventional 1043863.36 3817.12 2.81 FDL2 Chemical 2276519.16 5507.26 4.85 TB Conventional 530661.43 1401.60 1.15 TB Chemical 1022901.25 3956.69 2.78 By substituting the values of a, b, and  into the formula above, a head loss correction factor  can be calculated for each set of flow sweep data at each time step. The correction parameters for conventional debris head loss are applied to their respective test. For example, during a time step, the plant strainer debris loads are bounded by the FDL1 test. The conventional debris head loss for this time step is taken to be the peak head loss of the FDL1 test (2.640 ft, see Table 3.f.10-1) multiplied by the scaling factor calculated using the values of a, b, and  from the FDL1 test for conventional debris.

For chemical debris head loss, the scaling factor was calculated from values of a, b, and  for all four tests, and the greatest scaling factor was applied.

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02

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 net positive suction head (NPSH) margin were applied to address potential inability to pass the required flow through the strainer.

Response to 3.f.11:

As shown in the Response to 3.g.1, the strainer is submerged for all breaks at the beginning of recirculation and for the duration of recirculation.

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 the PBN head loss testing. Sufficient turbulence was maintained in the mixing section of the test 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. The turbulence was created by five mixing nozzles in the test tank as shown in Figure 3.f.12-1. The placement and size of the five mixing nozzles was carefully chosen to achieve the desired level of turbulence in the test tank without disturbing the debris bed formed on the test strainer.

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02 Figure 3.f.12-1: Mixing Nozzles in Test Tank

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 adjusted 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 PBN-specific flow sweep data.

During the 2016 head loss testing, flow sweeps were conducted during 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.

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02

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:

As stated in the Response to 3.f.7, a failure of the strainer would be recorded if flashing occurs due to the pressure downstream of the strainer being lower than the vapor pressure at the sump temperature. This evaluation is performed in NARWHAL, as discussed below.

Analysis of Flashing 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.

A conservative containment pressure curve was used in NARWHAL. For sump temperatures equal to or below 212°F, a containment pressure of 14.7 psia was used.

For sump temperatures above 212°F, containment pressure was set equal to the vapor pressure corresponding to the sump temperature.

A small amount (2 psi) of containment accident pressure was credited for the first 200 minutes of the accident. To demonstrate the margin in available containment pressure to prevent flashing, the containment pressure curve used in the NARWHAL evaluation, including the 2 psi credited for the first 200 minutes, is compared to the containment pressure curve from the PBN containment analysis in Figure 3.f.14-1.

The containment analysis curve is for a double-ended pump suction (DEPS) break with minimum safety injection. Figure 3.f.14-1 below shows the curve used in the NARWHAL evaluation in blue, and the containment analysis curve in orange. Note that the discontinuity in the blue curve at 12,000 seconds (200 minutes) corresponds to the time when the 2 psi of containment accident pressure is no longer credited.

As shown in the figure, even with the 2 psi of containment accident pressure credited, the containment pressures used in NARWHAL are at least 6 psi lower than those from the containment analysis. Therefore, crediting 2 psi of additional pressure for a short duration of time is conservative and acceptable.

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02 45 Containment Pressure in NARWHAL Base Case with Credited CAP 40 Containment Pressure from Accident Analysis 35 Containment Pressure (psia) 30 25 20 15 RHR starts CS starts in switchover to recirculation recirculation mode 10 1.0E+03 1.0E+04 1.0E+05 1.0E+06 1.0E+07 Time after Initiation of Accident (sec)

Figure 3.f.14-1: Containment Pressure Margin Analysis of Degasification Degasification was evaluated for PBN1 and PBN2 in NARWHAL. For each time step, the amount of air that could be released due to the pressure drop across the debris laden strainer was quantified by determining the air solubility decrease as flow travels through the strainer. This information was then used to calculate the void fraction, which was compared with the 2% acceptance limit given in NEI 09-10 (Reference 24

p. 28).

No containment accident pressure was credited for the degasification evaluation. The containment pressure was assumed to be 14.7 psia for sump temperature at or below 212 qF, and equal to water saturation pressure at the sump temperature for sump temperatures above 212 qF. The evaluation used a strainer submergence calculated from the midpoint of the strainer. Additionally, it was assumed that air released through degasification at the strainer transports to the pump suction, conservatively neglecting compression of the air bubbles due to increase in hydrostatic pressure as it travels to the pump at a lower elevation.

For any break evaluation in which the 2% void fraction acceptance limit during the sump recirculation phase is exceeded, a failure of the RHR pump is recorded for that break.

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02 Additionally, in the NPSH evaluation, the NPSH required obtained from the vendor curve was adjusted to account for impact on pump performance due to void fraction at the pump suction. See Section 3.g.3 for more details.

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02

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 loss-of-coolant accident (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 PBN analyzed the flow rates across the ECCS strainer and corresponding NPSH margins for the operating pumps using hydraulic modeling for different recirculation modes of operation. The recirculation mode that provides the most significant challenge to RHR pump NPSH is when the RHR and SI pumps are aligned in piggyback operation, with RHR pumps providing suction to the SI pump, and the two providing simultaneous upper plenum and cold leg injection. Per the results of this analysis, a procedural change was implemented to limit the RHR flow rate to 2,000 gpm. For conservatism, a strainer flow rate of 2,100 gpm was used when evaluating pump NPSH margin in the NARWHAL evaluation.

Minimum Water Level Minimum sump pool levels were calculated in NARWHAL and in a bounding hand calculation. The NARWHAL calculation performs comprehensive evaluation of GSI-191 phenomena in a self-consistent and time-dependent manner. For each break 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 holdup volumes from the total quantity of water in containment. The water level is then used to evaluate other failure criteria, including pump NPSH margin, for the same time step.

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). These evaluations informed the PBN strainer head loss and fiber penetration testing.

Table 3.g.1-1 summarizes the results of the minimum containment water level hand calculation for PBN1. The pool height values in the table are water depth above the containment floor at the start of RHR recirculation. The submergence values in the table were calculated by subtracting the top elevation of the strainers from the pool height. The elevation at the top of the strainer is 3 ft above the containment floor.

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02 Table 3.g.1-1: Minimum Sump Pool Water Levels (PBN1)

Pool Strainer Break Break Height Submergence Size Elevation (ft) (ft)

SBLOCA Top of the pressurizer 3.17 0.17 Below the elevation of the SBLOCA 3.29 0.29 top of the hot leg nozzles LBLOCA Top of the pressurizer 3.17 0.17 Below the elevation of the LBLOCA 3.72 0.72 top of the hot leg nozzles A similar summary is shown in Table 3.g.1-2 for PBN2.

Table 3.g.1-2: Minimum Sump Pool Water Levels (PBN2)

Pool Strainer Break Break Height Submergence Case Elevation (ft) (ft)

SBLOCA Top of the pressurizer 3.19 0.19 Below the elevation of the SBLOCA 3.31 0.31 top of the hot leg nozzles LBLOCA Top of the pressurizer 3.19 0.19 Below the elevation of the LBLOCA 3.74 0.74 top of the hot leg nozzles When determining the minimum sump pool water levels, the NARWHAL model uses many inputs and assumptions that are consistent with the hand calculation. However, the hand calculation includes some additional conservatisms that are not included in the NARWHAL model. For example, the hand calculation assumed a pool temperature of 130°F to maximize the pool water density and minimize the pool water level.

NARWHAL calculates the pool water density as a function of the pool temperature, which is significantly higher than 130°F early in the event.

An example break at PBN1 is shown below to compare the minimum water levels calculated in NARWHAL and the hand calculation. The NARWHAL calculated pool levels are shown in Table 3.g.1-3 for an 11.188-inch break at Weld RC-10-PZR-1001-

01. As shown in the table, the NARWHAL results and hand calculation results are similar, and only vary due to simplifications made in the hand calculation for the time-dependent inputs.

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02 Table 3.g.1-3: Comparison of NARWHAL Water Level Results with Hand Calculation NARWHAL Hand Calculation Time Calculation Pool Minimum Pool Height (ft) Level (ft)

Start of RHR Recirculation 3.28 3.17 End of CS Injection 4.38 4.20 Sump Temperature The PBN risk quantification in NARWHAL used a maximum sump temperature profile.

For each time step, the higher temperature between the double-ended pump suction leg break case with minimum safeguards and the double-ended hot leg break case with minimum safeguards was used, as shown in Figure 3.g.1-1. Note that the minimum safeguards temperature profile is conservatively used because it is higher than the temperature profile for the maximum safeguards case.

280 260 240 Sump Temperature (°F) 220 200 180 160 140 120 100 1.0E+00 1.0E+01 1.0E+02 1.0E+03 1.0E+04 1.0E+05 1.0E+06 1.0E+07 Time (seconds)

Figure 3.g.1-1: Sump Temperature for Double-Ended Pump Suction Break with Minimum Safeguards E3-107

Enclosure 3 Updated Final Responses to Generic Letter 2004-02

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 Rates As discussed in the Response to 3.g.1, the RHR flow rate was assumed to be 2,100 gpm, which was rounded up from the maximum calculated flow rate. Additionally, this flow rate is higher than the RHR pump flow rate which is procedurally limited to 2,000 gpm. Using a higher flow rate is generally conservative in terms of strainer head loss, flashing and degasification evaluations.

The CS and SI pumps piggyback off a RHR pump during recirculation. Therefore, specific CS pump and SI pump flow rates do not impact strainer failure analyzed in NARWHAL. The impact of the CS and SI pump flow rates on the reactor core failure due to in-vessel downstream effects is addressed in a bounding evaluation, as discussed in the Response to 3.n.

In the NARWHAL model, the same pump flow rates were used consistently for all breaks regardless of break size.

Minimum Water Level As stated in the Response to 3.g.1, minimum sump pool water levels were calculated in both NARWHAL and a hand calculation. The significant assumptions used in the water level evaluation are listed as follows.

1. The CAD model is not all-inclusive. That is, various mechanical items are not modeled (e.g., vents, equipment supports, annulus piping, etc.). Other items (e.g., I-beams) were simplified to facilitate more robust CFD analyses.

Therefore, the pool level, as a function of pool volume, is lower than if these items were included or resolved in more detail. This is conservative.

2. It was assumed that differences between PBN1 and PBN2 are small enough that a single generic water level analysis can adequately address both units.

This is reasonable since the significant dimensions (e.g., the diameters of the RWSTs, containment buildings, depths of sumps, etc.) are identical. The analyzed flow rates between the two units containment spray and ECCS have negligible differences. Therefore, using conservative inputs, assumptions, and methods ensure that using data from one of the two units will reasonably reflect the conditions for both units.

3. Fluid densities were calculated as pure water with the exception of solutions composed of NaOH. In the cases where fluid inventories are composed of solutions of water and other constituents (i.e., boric acid, etc.), the resulting density using the assumption of a pure water volume is slightly lower, leading to a negligibly larger final pool volume. Additionally, the masses of the various solutes are negligible with respect to the total mass of the water.

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4. It was assumed that the bounding containment pressure, temperature, and sump water temperature values are applicable to SBLOCAs. This is a reasonable assumption when used to calculate the density of the post-LOCA pool inventory and the vapor in containment hold-up, as the pressure and temperatures are expected to be considerably elevated for all LOCA sizes.

5. The inventory of the RCS is assumed to remain relatively constant throughout the operating cycle. This is a reasonable assumption because the RCS is a fixed volume that remains at constant temperature and pressure during full power operation, and any variation in the RCS liquid volume is negligible, considering the magnitude of the RCS liquid volume compared to the RWST liquid volume.

6. It was assumed that LBLOCAs will result in full depressurization of the RCS; therefore, during recirculation, the RCS will retain water up to the elevation of the break.

7. It was assumed that the time duration for manual actions necessary to realign the RHR, SI, and CS pumps from injection from the RWST to recirculation from the containment sump is zero. This is a conservative assumption because it minimizes the amount of water injected into containment, thereby reducing the containment water level.

8. For determining the amount of inventory held up as steam in the containment atmosphere, the pre-LOCA mass of steam in the atmosphere was determined to be 0 lbm by assuming a humidity of 0%, and the post-LOCA mass of steam was determined by assuming a humidity of 100%. This conservatively maximizes the atmospheric steam hold-up, thereby reducing the pool water level.

Sump Temperature To ensure a bounding sump temperature profile is used, the maximum temperature for each time step between the double-ended pump suction leg break case with minimum safeguards and the double-ended hot leg break case with minimum safeguards was used.

3. Provide the basis for the required NPSH values, e.g., 3 percent head drop or other criterion.

Response to 3.g.3:

The pump performance curve provided by the pump vendor was used to determine the value of NPSH required (NPSHr). The methods used by the vendor were in accordance with the test procedures outlined in the Standards of the Hydraulic Institute.

In the PBN NARWHAL model, NPSH available is not calculated in detail. Instead, a bounding evaluation of pump NPSH in a hand calculation is adjusted to account for differences in inputs. This hand calculation used linear interpolation of the NPSHr data E3-109

Enclosure 3 Updated Final Responses to Generic Letter 2004-02 points from the vendor curve to calculate NPSHr. The limiting case in the hand calculation used a pump NPSHr value of 12.75 ft.

The same NPSH required of 12.75 is used in the NARWHAL model. This value is increased to account for the impact on pump performance due to void fraction at pump suction. This adjustment uses the methodology recommended in RG 1.82 as follows:



Here, Dp is volumetric percentage of air in the fluid at pump suction. For example, if the void fraction at the pump suction is calculated to be 1%, Dp is equal to 1.

4. Describe how friction and other flow losses are accounted for.

Response to 3.g.4:

The bounding pump NPSH margin was analyzed in a hand calculation. Using the as-built isometric drawings, a model of the ECCS was defined in PROTO-FLO as a network of connecting node points. Pipe data such as length, material, schedule, components, and fittings was gathered from the drawings and other references and entered for each section of pipe in the system.

The piping losses were 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.

The NPSH analysis in NARWHAL used the piping losses from the limiting case in the hand calculation. The strainer head loss was not considered when calculating pump NPSH available or NPSH margin. Therefore, a RHR pump failure due to inadequate NPSH was recorded in the NARWHAL evaluation for any break in which the total strainer head loss exceeds the calculated NPSH margin.

5. Describe the system response scenarios for LBLOCA and SBLOCAs.

Response to 3.g.5:

For an LBLOCA, the RCS undergoes rapid depressurization due to the size of the break. Safety injection is automatically initiated upon an SIAS and the reactor is tripped. The following equipment is activated: SI pumps, RHR pumps, and all injection valves open. Additionally, the charging pumps are started to augment flow of the safety injection system. These pumps take suction from the RWST and inject to the RCS cold legs and reactor. This system line-up is referred to as the ECCS injection phase.

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02 As the energy is released into containment, the containment pressure will increase, and the containment spray actuation system (CSAS) will start the CSS pumps.

One RHR pump is required to inject borated water to the core. The high and low head injection flows during the injection phase are sufficient to prevent boric acid precipitation (BAP). Cold leg injection flow from the high head pumps is secured prior to the transfer to sump recirculation, but is reinitiated prior to the occurrence of BAP in the reactor vessel.

For small breaks, RHR injection into the core is not required. RHR pumps are isolated after a LOCA is determined to be a small break, while SI pumps remain in operation for the injection phase. Atmospheric dump valves are opened to reduce RCS pressure enough to allow low head injection within 6 to 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> of the event.

Before the RWST inventory is depleted, the suction source of the pumps must be switched. The RHR system is lined up to take suction from the containment sump when RWST level is less than or equal to 34% and the containment sump contains enough water to provide sufficient NPSH for the RHR pumps. CS pumps are manually realigned to take suction from the RHR pumps in recirculation. The switchover is complete when the suction valves from the RWST for all pumps are manually closed. Containment spray is not necessary for containment cooling after the injection phase, but continues to run for iodine removal.

Approximately 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> following switchover of CS to recirculation, the ECCS line-up is modified for simultaneous cold leg and upper plenum injection. For this operating mode, the SI pump takes suction from the RHR pump discharge and supplies flow to the cold leg.

6. Describe the operational status for each ECCS and CSS pump before and after the initiation of recirculation.

Response to 3.g.6:

Prior to the initiating event, the ECCS and CSS pumps will be in a state of stand-by readiness. Operation of each pump is described in the sections below.

Residual Heat Removal Pumps During the injection phase, the RHR pumps are active, drawing suction from the RWST and injecting into the reactor vessel and core barrel. Prior to switchover to recirculation, one RHR pump is secured. That pump is realigned to take suction from the recirculation sump and is restarted to deliver flow to the core and/or to the CS or SI pump suction. The RHR pump injecting from the RWST is then secured. This ensures flow to the core is maintained throughout the switchover process.

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02 Containment Spray System Pumps During the injection phase, the CS pumps are active, drawing suction from the RWST.

Prior to switchover to recirculation, CS A pump is secured. After CS switchover to recirculation, the CS A pump is re-aligned to take suction from RHR A pump discharge and is restarted. The second CS pump injecting from the RWST is then secured. The CS A pump is run for at least two hours in recirculation. After two hours, the CS A pump is secured.

High Head Safety Injection Pumps During the injection phase, the SI pumps are active, drawing suction from the RWST and injecting into the RCS cold leg and/or the reactor vessel. The SI pumps continue drawing from the RWST until after the RHR pumps are switched over to recirculation, at which time the SI pumps are secured. The SI pumps remain secured until the CS pump is operated for two hours in recirculation. Afterwards, the CS pump is secured, and the SI A pump is aligned to take suction from the RHR A pump, with the RHR A pump taking suction from the sump. The SI A pump injects into the RCS cold legs.

When both RHR pumps are in operation, the SI pump may be aligned to take suction from the RHR pump while the CS pump is still injecting from RWST. This alignment will not affect debris build-up on the strainers differently than the analyzed alignment as the strainer flow rate stays unchanged.

7. Describe the single failure assumptions relevant to pump operation and sump performance.

Response to 3.g.7:

As described in Enclosure 4, Section 4.3.3, the PBN risk quantification considered many different equipment configurations and was not limited to the worst single failure.

The PBN probabilistic risk assessment (PRA) model was used to identify the likelihood of different equipment configurations that could occur in response to a LOCA. All potential equipment failure combinations of the RHR, SI and CS pumps were screened and two unique equipment failure scenarios were analyzed in detail in NARWHAL. Refer to Enclosure 4, Section 4.3.3 for details.

x No Equipment failure x Single Train failure In the PBN NARWHAL models, the effects of random pump failures were addressed by assuming that the failures occur at the start of recirculation.

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02

8. Describe how the containment sump water level is determined.

Response to 3.g.8:

The water volume calculation used the methodology described below:

x A correlation was first developed for the relationship between the containment water level and the water volume using a 3-D CAD model.

x The quantity of water added to containment from the RWST, RCS, SI accumulators, and spray additive tank (SAT) was calculated.

x The quantity of water that is diverted from the containment sump by the following effects was evaluated:

o Water volume required to fill the CS discharge piping that is empty pre-LOCA.

o Water in transit from the containment spray nozzles to the containment floor.

o Water held up on containment surfaces exposed to containment spray and steam condensation.

o Steam held up in the containment atmosphere.

o RCS re-flood hold-up.

o Water held up in the pressurizer cubicle.

o Water held up in the sump A keyway tower.

o Hold-up in the refueling canal.

x Given the net mass of water added to the containment floor based on the second and third bullets listed above, the post-LOCA containment water level was calculated using the correlation developed in the first bullet.

As discussed earlier, the risk quantification in NARWHAL 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 and SBLOCA and provided inputs for evaluations that informed strainer head loss and fiber penetration testing. The hand calculation only reported water levels at the start of RHR recirculation and the end of CS injection.

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 inputs provided in the Minimum Water Level section of the Response to 3.g.1 and the assumptions provided in the Minimum Water Level section of the Response to 3.g.2 ensure that minimum (conservative) containment water levels are calculated.

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02

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:

The Response to 3.g.8 listed the holdup volumes that were accounted for in the PBN water level analysis in both the water level hand calculation and the NARWHAL models.

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 such as concrete and structural steels will displace water. These volumes were accounted for in the containment water volume calculation. The 3-D CAD model of containment was used to determine the correlation between the containment pool volume and water level.

Smaller equipment, cables, and instruments were excluded from the CAD model and therefore provide some conservatism in the resulting water levels, as stated in the Response to 3.g.2. Figure 3.g.11-1 shows the level of detail of structures and components credited for water displacement in the containment water volume calculation.

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02 Figure 3.g.11-1: PBN PBN1 Containment CAD Model

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:

x The TS minimum initial RWST level was used for the initial RWST water level.

The low level (plus an amount to account for uncertainty) was used for the final RWST water level. The minimum RWST injection volume at the start of recirculation is 165,787 gal. This was input into the NARWHAL model as an initial mass, calculated using the density at the maximum TS allowable RWST temperature of 97.5qF, as shown in the table below. The table also shows the final RWST mass values at the start of RHR recirculation and CS recirculation.

x There are two SI accumulators for each unit, and the minimum combined volume of the SI accumulators is 16,458 gal. This was input into the NARWHAL model a E3-115

Enclosure 3 Updated Final Responses to Generic Letter 2004-02 total mass, calculated using the density at the maximum temperature of 120qF, as shown in the table below.

x The initial RCS liquid volume is estimated to be 42,003 gal. This was input into the NARWHAL water level calculation as a mass, as shown in the table below.

x The RCS represents both a source of water and a hold-up volume. The mass of water held up in the RCS after the accident may be more or less than the initial RCS mass depending on the size and elevation of the break, as shown in the table below.

x The minimum volume of NaOH solution provided by the SAT is 541 gal, which was converted to mass as input into the NARWHAL, using the density of 30%

NaOH by weight solution at the maximum SAT temperature of 85°F.

Table 3.g.12-1: Water Source Inputs for Water Level Analysis Variable Value Units Initial RWST Mass 2,231,809 lbm Final RWST Mass (start of RHR recirculation) 862,648 lbm Final RWST Mass (start of CS recirculation) 460,480 lbm SAT Mass 5,968 lbm SI Accumulators 136,086 lbm Initial RCS Mass 254,078 lbm RCS Hold-up: SBLOCA E1 42,003 gal RCS Hold-up: SBLOCA E2 47,629 gal RCS Hold-up: MB/LBLOCA E1 22,219 gal RCS Hold-up: MB/LBLOCA E2 47,629 gal

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:

No credit was taken for containment accident pressure in determining NPSH available.

Containment pressure is further described in the Response to 3.g.14.

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02

14. Provide assumptions made which minimize the containment accident pressure and maximize the sump water temperature.

Response to 3.g.14:

Containment Pressure As mentioned in the Response to 3.g.13, no containment accident pressure was credited for NPSH evaluation. For sump temperatures equal to or below 212°F, a containment pressure of 14.7 psia was used. For sump temperatures above 212°F, containment pressure was set equal to the vapor pressure corresponding to the sump temperature.

Sump Temperature The NPSH evaluation was performed in NARHWHAL using a maximum sump temperature profile from the containment analysis as discussed in the Responses to 3.g.1 and 3.g.2.

15. Specify whether the containment accident pressure is set at the vapor pressure corresponding to the sump liquid temperature.

Response to 3.g.15:

As discussed in the Response to 3.g.14, the containment pressure was set at 14.7 psia for sump temperatures below or equal to 212°F. For sump temperatures above 212°F, the containment pressure was set equal to the vapor pressure corresponding to the sump liquid temperature.

16. Provide the NPSH margin results for pumps taking suction from the sump in recirculation mode.

Response to 3.g.16:

The RHR pump NPSH margin was evaluated using the PBN NARWHAL model in a time-dependent and break-specific manner. The RHR pump NPSH margin results as a function of sump temperature are shown in the tables below for two example breaks.

Table 3.g.16-1 shows the NPSH margins vs. sump temperature for the PBN1 worst fiber break that does not result in a strainer failure. As shown in the table, the minimum NPSH margin is over 2 ft occurring during the early phase of the recirculation. The pump NPSH margin increases rapidly after the sump temperature drops below 212qF.

Similar results are shown in Table 3.g.16-2 for the PBN2 worst Cal-Sil break that does not result in a strainer failure.

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02 Table 3.g.16-1: RHR Pump NPSH Margin for PBN1 Worst Fiber Break that Does Not Result in Strainer Failure Sump NPSH Margin Before Net NPSH Margin After Strainer Head Loss Temperature Subtracting Strainer Subtracting Strainer (ft)

(°F) Head Loss (ft) Head Loss (ft) 242 4.12 1.43a 2.69 222 4.94 1.47a 3.47 212 4.83 1.49a 3.34 200 12.27 2.48a 9.79 182 21.03 2.58a 18.45 160 28.07 2.73a 25.34 138 31.84 4.95b 26.89 120 34.15 5.29b 28.86 a This head loss includes clean strainer head loss and conventional debris (fiber and particulate) head loss.

b This head loss includes clean strainer head loss, conventional debris (fiber and particulate) head loss and chemical debris (sodium aluminum silicate) head loss.

Table 3.g.16-2: RHR Pump NPSH Margin for PBN2 Worst Cal-Sil Break that Does Not Result in Strainer Failure Sump NPSH Margin Before Net NPSH Margin After Strainer Head Loss Temperature Subtracting Strainer Subtracting Strainer (ft)

(°F) Head Loss (ft) Head Loss (ft) 242 3.61 1.43a 2.18 222 4.45 1.47a 2.98 212 4.33 1.49a 2.84 200 11.78 1.51a 10.27 182 20.54 1.55a 18.99 160 27.57 1.61a 25.96 138 31.63 3.73b 27.9 120 33.98 3.95b 30.03 a This head loss includes clean strainer head loss and conventional debris (fiber and particulate) head loss.

b This head loss includes clean strainer head loss, conventional debris (fiber and particulate) head loss and chemical debris (sodium aluminum silicate) head loss.

Since the CS and SI pumps take suction from the RHR pump discharge during recirculation, the NPSH margins for the RHR pumps are more limiting.

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02

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, Ameron 90 epoxy finish coat.

Response to 3.h.1:

The types of coating and systems used in PBN1 and PBN2 containment are presented in Table 3.h.1-1 and Table 3.h.1-2, respectively.

Qualified Coatings Table 3.h.1-1: PBN1 Qualified Coatings Systems Used in Debris Generation Analyses DFT Density Substrate Layer Type (mil) (lbm/ft³)

1st Coat Dimetcote 6 - IOZ 7 300 nd Steel Surfaces 2 Coat Amercoat 66 - Epoxy 10.5 97.1 Total 17.5 --

1st Coat Carboline 195 - Epoxy 27.4 109 Concrete 2nd Coat Phenoline 305 - Epoxy 5.0 101.3 Walls Total 32.4 --

Concrete 1st Coat Phenoline 305 - Epoxy 10.9 101.3 Floors Table 3.h.1-2: PBN2 Qualified Coatings Systems Used in Debris Generation Analyses DFT Density Substrate Layer Type (mil) (lbm/ft³)

1st Coat Dimetcote 6 - IOZ 7 300 nd Steel Surfaces 2 Coat Amercoat 66 - Epoxy 10.5 97.1 Total 17.5 --

1st Coat Carboline 195 - Epoxy 33.4 109 Concrete 2nd Coat Phenoline 305 - Epoxy 5.0 101.3 Walls Total 38.4 --

Concrete 1st Coat Phenoline 305 - Epoxy 10.9 101.3 Floors E3-119

Enclosure 3 Updated Final Responses to Generic Letter 2004-02 Unqualified Coatings Unqualified coatings are those that fail under design basis accident conditions and create debris that could be transported to the containment recirculation strainers.

There are several types of unqualified coatings applied over numerous substrates within containment. The quantities of these unqualified coatings are shown in Table 3.h.1-3 for PBN1 and Table 3.h.1-4 for PBN2.

Table 3.h.1-3: PBN1 Unqualified Coatings Quantities Used in Analyses Coating Type Volume (ft3)

IOZ 1.49 Alkyd 4.11 Epoxy (Unqualified) 2.41 Table 3.h.1-4: PBN2 Unqualified Coatings Quantities Used in Analyses Coating Type Volume (ft3)

IOZ 1.73 Alkyd 5.78 Epoxy (Unqualified) 2.93 The higher loads for Unit 2 were increased by 10% for future operating margins before being used as inputs for both units in the NARWHAL models, as shown in the table below. These volumes were applied for all breaks.

Table 3.h.1-5: Quantities of Unqualified Coatings used in NARWHAL Model Volume Coating Type (ft3)

Unqualified IOZ 1.90 Unqualified Alkyd 6.36 Unqualified Epoxy 3.22 Total 11.48 Actively Delaminating Qualified Epoxy Coatings The actively delaminating qualified (ADQ) epoxy coatings was assumed to fail as a combination of particulate and chips of different sizes. See the Response to 3.h.6 for the size distribution used. The table below shows the quantities of ADQ epoxy coatings used in the PBN risk quantification.

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02 Table 3.h.1-6: Quantity of ADQ Epoxy Coatings with Operating Margin for Units 1 and 2 Size Disignation Volume (ft3)

Fine Particulate 0.40 Flat Fine Chips 1.21 Flat Small Chips 0.31 Flat Large Chips 0.67 Curled Large Chips 0.67 Total 3.26

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 PBN1 and PBN2 debris transport analyses:

x It was conservatively assumed that all unqualified coatings are located in lower containment. This is conservative since it results in 100% of unqualified coatings being present in the pool at the start of recirculation and results in 100% transport of this debris type.

x It was assumed that the settling velocity of particulate debris (insulation, dirt/dust, and coatings) can be calculated using Stokes Law. This is a reasonable assumption since the particulate debris is generally spherical, small in size, and would settle slowly (within the applicability of Stokes Law). This assumption has been addressed in the San Onofre (Reference 25) and Indian Point (Reference 26) Audit Reports, and it has been concluded that it is not a significant factor with respect to debris transport since no credit is taken for debris settling using this approach.

x Unqualified coatings outside the ZOI were assumed to fail after pool fill has occurred so that no unqualified coatings transport to inactive cavities during pool fill-up.

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:

PBN has qualified coatings (IOZ and epoxy), unqualified coatings (IOZ and epoxy),

and ADQ epoxy coatings. Silica flour, with a median size distribution of approximately 13.5 microns, was used as a surrogate for qualified coatings, unqualified coatings, and the particle portion of the ADQ epoxy coatings. Pressure washed paint chips, with E3-121

Enclosure 3 Updated Final Responses to Generic Letter 2004-02 a nominal size of approximately 0.125, were used as a surrogate to model the flat fine chip portion of the ADQ epoxy coatings. Paint chips were used as a surrogate for the small chip portion of the ADQ epoxy coatings. 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.h.3 and 3.f.4.

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 calculations:

x Epoxy and alkyd unqualified coatings were assumed to have properties as listed in Table 3.h.1-1 and Table 3.h.1-2.

x Unqualified IOZ was assumed to have a particulate size of 10 m (Ref. 2, p.

51) and the density of Carbozinc 11 - a typical IOZ used in nuclear power plants

- 208 lb/ft3 (27.81 lb/gal) .

x Qualified coatings were analyzed within a 4.0D ZOI. This ZOI has been previously accepted by the NRC (Reference 14 p. 2).

The quantities of unqualified coatings in containment were based on detailed logs maintained by the plant and are contained in Table 3.h.1-3 and Table 3.h.1-4 for PBN1 and PBN2, respectively.

The quantities of qualified coatings generated by selected breaks are shown in Tables 3.b.4-1 through 3.b.4-4 along with the rest of the ZOI debris for both PBN1 and PBN2.

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 11, pp. 3-12 through 3-13) and the associated NRC SE (Reference 8 p. 22), the qualified coatings debris within the ZOI and the unqualified coatings were treated as 10 micron particulate. See the Responses to 3.h.1, 3.h.2, and 3.h.3 for additional debris characteristics description.

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02 A portion of the failed ADQ epoxy coatings were treated as chips. The table below shows the size distribution applied to the ADQ epoxy coatings. Note that the flat large chips and curled chips were not included in the head loss testing or risk quantification because they have been shown not to transport even with agitation (see the Response to 3.f.4).

Table 3.h.6-1: ADQ Coatings Size Distribution Size Size Range Percentage of Designation (inch) Total Mass Fines (particles) 0.006 12.38%

Flat Fine Chips 0.015 37.13%

Flat Small Chips 0.125-0.5 9.43%

Flat Large Chips 0.5-2.0 20.53%

Curled Chips 0.5-2.0 20.53%

7. Describe any ongoing containment coating conditions assessment program.

Response to 3.h.7:

PBN performs coatings assessments in containment on a refueling interval frequency to ensure the total inventory of coatings debris remains bounded by the design basis for the sump screens. The coatings assessments are controlled by procedure under the PBN protective coatings program. The assessment procedure conforms to the intent of ASTM (Reference 27, Enclosure 1, pg. 28-29).

The coating assessment procedure requires a general visual inspection of all accessible surface areas inside containment, with thorough inspections performed as needed in areas exhibiting degradation including such conditions as flaking, blistering, delamination, cracking, checking, pinholes, rust, or damaged or abraded areas.

Coating assessment walkdowns are performed by at least two qualified individuals, including the coating program owner and a quality control inspector. The qualifications of these individuals meet the intent of EPRI and ASTM. The general visual inspection involves comparison of the as-found condition to the previously documented condition, and documenting changes or new conditions that are observed. Where new or further degradation of coatings is noted, a more thorough inspection may be performed to better define the extent and cause of degradation.

Inspections may involve several different techniques including visual inspection, non-destructive tests for dry film thickness, destructive tests for adhesion, and destructive sampling for subsequent chemical analysis. Supplemental inspections and tests are performed in accordance with current industry guidance. Where nonconforming conditions are noted that have not been previously evaluated, or where the condition has further degraded as compared to previous results, the corrective action program is used to identify and evaluate the condition.

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02 The general condition of the containment coatings is summarized in a report, which is issued following each refueling outage. The most recently issued report for each unit contains the log of the total surface area and volume of all unqualified and degraded coatings within the containment, as of the end of the most recent refueling outage.

The report also contains a computation of the current operating margin as compared to the volumes of coating debris used in the design and testing of the containment sump strainers.

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02

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:

PBN has implemented a number of actions to enhance containment cleanliness as documented in the response to Bulletin 2003-01. Detailed containment cleanliness procedures exist for unit restart readiness and for containment entry at power. These procedures incorporate the guidance of Nuclear Energy Institute (NEI) 02-01 to minimize miscellaneous debris sources within the containment and ensure the operational readiness of the sump strainers. At the end of each outage, a thorough inspection of containment is performed to ensure the containment is free of loose debris and fibrous material, remove items not approved for storage in containment, and ensure the containment sump strainers and strainer piping can perform their design function.

Additionally, these procedures also satisfy Technical Specification Surveillance 3.5.2.6, Verify by visual inspection that the suction inlet to the containment sump is not restricted by debris and that the debris strainers show no evidence of structural distress or abnormal corrosion. Lastly, the maintenance director is in charge of maintaining the general housekeeping of containment, which includes tracking the overall cleanliness of containment and promptly correcting identified deficiencies.

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02

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:

Foreign material exclusion programmatic controls are in place at PBN that consider the containment a plant system. This ensures that proper work control is specified for debris-generating activities within containment to prevent introduction of foreign material into containment that could challenge the containment recirculation function.

Additionally, the foreign material exclusion program requires that engineering be consulted anytime foreign material covers are placed on or modifications are performed on the containment sump strainers. Note that the foreign material exclusion controls are applicable in Modes 1-4 only. However, the close-out inspection discussed in Response to 3.i.1 is required prior to leaving Mode 5, which works in concert with the foreign material exclusion controls to limit foreign material in containment.

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:

NextEra engineering change processes and procedures ensure modifications that may affect the ECCS, including sump performance, are evaluated for GL 2004-02 compliance. During engineering change preparation, the process requires specific critical attributes be listed, evaluated and documented when affected. This includes the introduction of materials into containment that could affect sump performance or lead to equipment degradation (e.g., GSI-191), including insulation, coated equipment and components, and exposed aluminum. It also includes repair, replacement, or installation of coatings inside of primary containment.

NextEra adopted the industrys standard design change process, including the industry procedure IP-ENG-001 (Reference 28). The standard process and tools are intended to facilitate sharing of information, solutions and design changes throughout the industry. This process requires activities that affect UFSAR described structure, system, or component (SSC) design functions to be evaluated as a design change in accordance with PBNs 10 CFR 50 Appendix B program. This includes modifications that would impact the containment sump. Design changes require a final impact review meeting (i.e., final design workshop) and assessment in accordance with 10 CFR 50.59. Additional meetings may be required based on complexity and risk of the change. A failure modes and effects analysis is required if the design change introduces any new failure modes or changes failure modes for the affected SSCs.

This guidance has been enhanced by an engineering specification that brings together, in one document, the insulation design documents that determine the design E3-126

Enclosure 3 Updated Final Responses to Generic Letter 2004-02 basis for the insulation debris component of the containment recirculation strainer design. This specification provides guidance for evaluating and maintaining piping and component insulation configuration within the containment buildings at PBN1 and PBN2.

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:

Temporary configuration changes are controlled by plant procedure. This process maintains configuration control for non-permanent changes to plant structures, systems, and components while ensuring the applicable technical reviews and administrative reviews and approvals are obtained. If, during power operation conditions, the temporary alteration associated with maintenance is expected to be in effect for greater than 90 days, the temporary alteration is screened, and if necessary, evaluated under 10 CFR 50.59 prior to implementation.

In accordance with 10 CFR 50.65 (Maintenance Rule), an assessment of risk resulting from the performance of maintenance activities is required. Prior to performing maintenance activities (including but not limited to surveillance, post-maintenance testing, and corrective and preventive maintenance), the licensee assesses and manages the increase in risk that may result from the proposed maintenance activities. The scope of the assessment may be limited to those SSCs that a risk-informed evaluation process has shown to be significant to public health and safety.

In general, the risk assessment ensures that the maintenance activity will not adversely impact a dedicated/protected train. The dedicated/protected train ensures a system is capable to perform its intended safety function. PBN implements the requirement via procedures.

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:

At PBN, insulation modifications included replacing the mineral wool on the pressurizer from each unit with RMI and replacing the fibrous insulation on both RCPs in each unit with RMI. Additionally, the fibrous insulation on the PBN2 main RCS loop piping has been replaced with RMI.

b. Any actions taken to modify existing insulation (e.g., jacketing or banding) to reduce the debris burden at the sump strainer.

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02 Response to 3.i.5.b:

There were no additional actions taken (e.g., jacketing or banding) to reduce the debris burden at the sump strainer.

c. Modifications to equipment or systems conducted to reduce the debris burden at the sump strainers.

Response to 3.i.5.c:

Debris interceptors were installed in the PBN1 containment. Specific credit was not taken for the reduction of problematic debris transport to the strainer (e.g., fiber fines, particulate fines). In reality, these debris interceptors would likely reduce the potential transport of the debris sources.

d. Actions taken to modify or improve the containment coatings program.

Response to 3.i.5.d:

Significant quantities of degraded or unqualified coatings have been remediated by removal, replacement, or qualification by a combination of testing and analysis.

Containment coatings are discussed in greater detail in the Response to 3.h.

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02

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 intent of the modification was to perform the hardware changes required to bring PBN into conformance with GSI-191 by replacing the original small area screens with screens having a substantially increased surface area.

Original Screens The original PBN ECCS screens consisted of a single vertical cylindrical screen for each train of ECCS. The screens were fabricated from stainless steel with 1/8 diameter perforations. The screens were each 13.5 in diameter, and 71 tall for each train of ECCS, and were completely enclosed in a single, larger trash rack fabricated from 1/2 thick stainless steel. 1 wide vertical slots were cut in the surface of the trash rack to admit sump water while excluding larger debris. There were approximately 256 slots that were 6 tall, and approximately 32 slots that were 5 tall. The solid top of the rack served to close off the top of the screens and prevent debris intrusion should the screens become totally submerged (Reference 27, Enclosure 1, pg. 30).

The effective area of each of the original screens was approximately 21 square feet per train if fully submerged. At the time that sump recirculation would have been initiated, the screens would have been only partially submerged with a minimum of

~38 in the sump. The effective area would have then been approximately 11 square feet per train (Reference 27, Enclosure 1, pg. 30).

Replacement Screens The modification installed a passive, safety-related Sure-Flow Strainer assembly, engineered and manufactured by Performance Contracting Incorporated (PCI).

Originally, each strainer train at PBN1 and PBN2 consisted of 11 strainer modules connected to the respective trains sump outlet pipe. The installations were performed during the spring 2006 and 2007 refueling outages. An additional 3 modules were added to each train in the Fall 2008 and Fall 2009 outages.

Figure 3.j.1-1 and Figure 3.j.1-2 show the general arrangement of the strainer installation at each unit. The effective surface area of each replacement strainer train is 1,904.6 ft2, more than a 90-fold increase over the area of the original screens if the screens would have been fully submerged. The replacement screens are designed to draw a flow rate of 2,200 gpm evenly across the entire active surface (Reference 27, Enclosure 1, pg. 30), reducing the screen approach velocity to just 0.0026 fps.

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02 The replacement strainers would be fully submerged by the time that sump recirculation initiates.

Figure 3.j.1-1: PBN1 ECCS Strainer General Arrangement E3-130

Enclosure 3 Updated Final Responses to Generic Letter 2004-02 Figure 3.j.1-2: PBN2 ECCS Strainer General Arrangement The 14 modules in each strainer train consist of a core tube and mounting tracks. The modules are nearly identical with the only difference being the flow control hole sizes in the core tube. Each module is independently supported by pinned connections to a mounting track. The modules are connected with thin gauge stainless steel bands that are used to prevent debris from entering the system between adjacent modules.

The bands are secured with a seismic latch. This connection permits relative motion in the axial direction as the core tube can slide relative to the stainless steel bands, and accommodates disassembly for inspection, repair, replacement, or installation of E3-131

Enclosure 3 Updated Final Responses to Generic Letter 2004-02 additional modules to extend the assemblies, or strings, of strainer modules (Reference 27, Enclosure 1, pg. 31).

Figure 3.j.1-3 shows the replacement strainer module. Each module is made of stainless steel perforated sheet with a nominal hole diameter of 0.066. The perforated sheets are riveted together along the outside edge and fitted to the core tube along the inner edges. Because of the convoluted configuration, and internal and external cross bracing, the modules are inherently rugged and do not require an external trash rack to provide protection from larger debris or incidental damage. The bottom active strainer surfaces on the modules are located approximately 3 above the containment floor (Reference 27, Enclosure 1, pg. 31).

Figure 3.j.1-3: Replacement Strainer Modules (Reference 27, Enclosure 3)

Figure 3.j.1-4 shows the typical strainer installation at PBN. The mounting tracks are secured to the containment floor by anchor bolts. The strainer module strings are connected to the containment outlets by 16 diameter stainless steel piping anchored and supported against the same loading conditions (Reference 27, Enclosure 1, pg.

31).

At the point that the 16 diameter piping turns downward to connect to the containment outlets, the piping transitions to an 18 diameter elbow. The large diameter elbow maximizes the annular flow area between the existing sump outlet valve disk and the elbow wall. The slower velocity also serves to minimize the frictional head loss through this transition into the piping (Reference 27, Enclosure 1, pg. 31).

The strainer core tubes were fabricated from 16 stainless steel pipe. The core tubes have variable sized windows cut in the walls to admit flow of strained water from the inside of the perforated strainer sheets. The windows are sized to ensure an even E3-132

Enclosure 3 Updated Final Responses to Generic Letter 2004-02 distribution of flow through the entire strainer surface. This provides maximum assurance of even debris loading, while minimizing total head loss and potential for air entrainment (Reference 27, Enclosure 1, pg. 31).

Figure 3.j.1-4: Replacement Strainer Typical Install (Reference 27, Enclosure 3)

2. Provide a list of any modifications, such as reroute of piping and other components, relocation of supports, addition of whip restraints and missile shields, etc.,

necessitated by the sump strainer modifications.

Response to 3.j.2:

There were no plant modifications that were necessitated by the sump strainer modifications.

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02

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:

Strainer Modules In PBN1 and PBN2, the containment recirculation sumps and debris interceptors provide filtered suction intake for the RHR pumps. Each strainer assembly is a passive unit (i.e., there are no active components). The strainer assemblies are considered safety related. (See the Response to 3.j for additional description.)

There are two independent strainers at each unit. Each strainer consists of 14 modules. The modules are connected with stainless steel bands that are used to prevent debris from entering the system between adjacent modules. The bands are secured with a seismic latch. This connection permits relative motion in the axial direction as the core tube can slide relative to the stainless steel bands. Each module is made of stainless steel perforated sheet. The perforated sheets are riveted together along the outside edge and fitted to the core tube along the inner edges. The strainer core tubes and extension sleeve are fabricated from 16 inch diameter stainless steel pipe. The end cover is made of solid stainless steel plate.

The loads on the strainer are comprised of weight, pressure, and dynamic loads. The dynamic loads come from two sources, seismic inertia and hydrodynamic drag loads due to sloshing. The strainers are loaded due to the inertia effect from the motion of the containment floor during an earthquake. Hydrodynamic loads on the strainer are due to the motion of the water surrounding the strainer during a seismic event. The weight loads include the weight of the strainer components themselves, and the weight of the debris that accumulates on the strainer. The weight of debris per strainer module is taken as 100 lb per module. The normal operating pressure load is simply the pressure drop across a clean strainer. There are no thermal expansion loads since E3-134

Enclosure 3 Updated Final Responses to Generic Letter 2004-02 the strainers are free to expand without restraint. The piping is not rigidly attached to the strainer modules. Therefore, the piping is also free to expand without imposing any thermal loads on the strainers.

Sump Strainer Structural Analysis The Sump strainers were qualified using a combination of manual calculations generated in Mathcad, as well as finite element analyses using the GTSTRUDL software and the ANSYS software. The strainer frame and assembly was qualified using GTSTRUDL while the perforated strainer plates were qualified using ANSYS.

Applicable Strainer Codes The detailed evaluations were performed using the rules, as applicable, of ANSI/ASME B31.1 Power Piping 1998 Edition through 1999 Addenda. The use of the ASME Boiler and Pressure Vessel Code is primarily for the qualification of pressure retaining parts of the strainer which are not covered in B31.1 (perforated plate, and internal wire stiffeners). Some parts of the strainers (radial stiffeners, connecting rods, edge channels, seismic stiffeners, etc.) are classified as part of the support structure.

These types of components are covered under the AISC 9th Edition. ANSI/AISC N690-1994,"Specification for the Design, Fabrication, and Erection of Steel Safety Related Structures for Nuclear Facilities was used to supplement the AISC in any areas related specifically to the structural qualification of stainless steel. The strainer also has several components made from thin gage sheet steel, and cold formed stainless sheet steel. Therefore, SEI/ASCE 8-02, "Specification for the Design of Cold-Formed Stainless Steel Structural Members", was used for certain components where rules specific to thin gage and cold form stainless steel should be applicable. The rules for Allowable Stress Design (ASD) as specified in Appendix D of this code were used.

This was further supplemented by the AISI Code where the ASCE Code is lacking specific guidance. Finally, guidance was also taken from AWS D1.6, "Structural Welding Code - Stainless Steel as it relates to the qualification of stainless steel welds. The analysis of the anchorage to the containment concrete slab was in accordance with the Hilti technical Guide.

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02 Load Combinations for the Strainer The applicable load combinations for the strainers are:

Table 3.k.1-1: Load Combinations for the Strainer Load Condition Combination (1a) Normal Operating DP + DW (1b) Normal Operating (Outage/Lift Load) DW + LL (2) Upset DP + DW + WD + OBE (3) Emergency/Faulted DP + DW + WD + SSE Where, DW= Dead Weight Load LL= Live Load (additional loads on strainers during outages or during installation, live load is not applicable during operation)

WD= Weight of Debris DP= Differential Pressure OBE= Operating Basis Earthquake SSE= Safe Shutdown Earthquake Note that combination (3) was classified as Emergency Condition for all ASME Code evaluations and Faulted for all components governed by AISC and ACI Codes. Also note that wind, snow, tornado, and jet force loads are not applicable. Flood loads are considered for Load Combinations 2 and 3. Flood loads consist of the effects due to an earthquake in a submerged condition (sloshing and added mass). There are no hydrostatic pressure loads associated with flooding since the flood waters are present on all sides. Thermal expansion stresses were considered negligible.

Core tube Combinations The core tube was evaluated as piping per B31.1 Paragraph 104.8 as applicable.

Since the B31.1 does not explicitly identify how to incorporate the Emergency SSE loads, PBN used ASME Section III as a guide as discussed in site-specific design requirements.

Table 3.k.1-2: Load Combinations for the Core Tube B31.1 Eq. No Load Condition Load Combination Allowable Stress Normal DW 1.0 Sh 12 (OBE) Upset DW + OBE 1.2 Sh 12 (SSE) Emergency DW + SSE 1.8 Sh E3-136

Enclosure 3 Updated Final Responses to Generic Letter 2004-02 Strainer Pressure Retaining Plates Combinations For the pressure retaining plates, such as the perforated plate and the core tube end cover stiffener plate, the B31.1 Code does not provide any design guidelines as discussed above. For the perforated plate, the equations from Appendix A, Article A-8000 of the ASME B&PV Code,Section III, 1998 Edition through 1999 Addenda was used to calculate the stresses. Note that Article A-8000 refers to Subsection NB for allowable stresses, which are defined in terms of stress intensity limits, Sm. However, in keeping with the B31.1 maximum principal stress design philosophy, principal stresses were calculated and compared to the allowables based on the ASME allowable stress limit, S.

Stress limits for the pressure retaining plates were taken from ASME Section III, subsection NC-3321.

Table 3.k.1-3: Load Combinations for the Pressure Retaining Plates Load Condition Stress Type Allowable Stress Design Level Normal/Upset* Primary Membrane Stress 1.0 Sh Level A Primary Membrane (or Local) +Bending 1.5 Sh Emergency Primary Membrane Stress 1.5 Sh Level C Primary Membrane (or Local) +Bending 1.8 Sh

• Allowable stresses for Upset condition may be increased by 10% as permitted by NC-3321 (Reference 29)

Strainer Structural Components Combinations Based on the discussion provided earlier in this section, the allowable stresses on the strainer structural components was based on the AISC 9th Edition. The allowable stress for the SSE Load Combinations was taken from site specific design requirements.

Table 3.k.1-4: Load Combinations for the Strainer Structural Components Load Condition Load Combination Allowable Stress Normal Operating 1a, 1b 1.0 AISC Upset 2 1.0 AISC Faulted 3 1.5 AISC but not to exceed 0.9 Sy E3-137

Enclosure 3 Updated Final Responses to Generic Letter 2004-02 Debris Interceptors (Perforated Flow Diverters)

Like the strainers, the debris interceptors are passive units and intended as pre-filters to the strainers. There are two types designated as B and C. Previously installed Type A debris interceptors were removed by EC 284675. Type B debris interceptors consist of structural steel beams supporting grating and a steel perforated plate. Type B interceptors are located in the windows of the steam generator cubicles at El. 10.

Type C interceptors are 6 high curbs made from stainless steel plate, which is bent into a circular shape and attached to the concrete floor at El. 66.

Debris Interceptors Structural Analysis Type B and C debris interceptors were designed using manual calculations where required for design.

Applicable Debris Interceptor Codes The detailed evaluations were performed using the rules, as applicable, under the AISC 9th Edition Code and ANSI/AISC N690-1994 Specification.

Table 3.k.1-5: Load Combinations for the Debris Interceptors

1. DL + LL1 + OBE

2. DL + LL2 + OBE

3. DL + LL1 + SSE (Conservative)

4. DL + LL2 + SSE (Conservative)

Where:

DL= Dead load LL1= Initial differential head across the debris interceptors LL2= High water level differential head across the debris interceptors Connecting Piping and Supports Each sump strainer has two 16-inch diameter pipes that exit the A and B strainer trains and anchor into the floor. Each section of pipe (the pipe run between the strainer and floor) is connected via flanged sections of pipe up to the strainer assembly.

Piping and Pipe Support Analysis Mathcad was used to perform the manual calculations for the supports and various other associated piping calculations. AutoPlPE was used for the piping analysis.

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02 Load Combinations for Piping The piping was evaluated in accordance with ANSI/ASME B31.1 Power Piping 1998 Edition. The Piping supports, baseplates other mounting hardware was evaluated to AISC 9th Edition as permitted in Paragraph 120.2.4 of the B31.1 Code. Additional guidance was also taken from ANSI/AISC N690-1994, SEI/ASCE 8-02 and AWS D1.6 to supplement the AISC in any areas related to the structural qualification of stainless steel. Since the B31.1 does not explicitly identify how to incorporate the emergency SSE loads, PBN used ASME Section III as a guide, as discussed in Section 6.0 of DG-M09.

Table 3.k.1-6: Load Combinations for Piping B31.1 Eq. No Load Condition Load Combination Allowable Stress 11 Normal DP + DW 1.0 Sh 12 (OBE) Upset DP + DW + OBE 1.2 Sh 12 (SSE) Emergency DP + DW + SSE 1.8 Sh 13 Thermal T1 1.0 SA Where, DW= Dead Weight Load DP= Differential Pressure OBE= Operating Basis Earthquake SSE= Safe Shutdown Earthquake T1= Thermal Expansion The thermal expansion stresses were based on a stress range from the ambient condition of 70 °F to the maximum operating condition of 250 °F (T=180 °F) .

Piping Support Structural Components Load Combinations The allowable stresses on the piping support components were based on the AISC 9th Edition. Also, the allowable stresses for the sump sole plate tabs, bolts, and welds were based on the AISC 9th Edition. The allowable stress for the SSE Load Combinations was taken from Section 6.9 of DG M10.

Table 3.k.1-7: Load Combinations for the Piping Support Structural Components Load Condition Load Combination Allowable Stress Normal DW + T1 1.0 AISC Upset DW + OBE + T1 1.0 AISC Faulted DW + SSE + T1 1.5 AISC but not to exceed 0.9 Sy E3-139

Enclosure 3 Updated Final Responses to Generic Letter 2004-02

2. Summarize the structural qualification results and design margins for the various components of the sump strainer structural assembly.

Response to 3.k.2:

The structural margin of the strainer components were calculated using GTSTRUDL structural analysis software.

Strainer Modules for Units 1 & 2 As documented below, the interaction ratio of each subcomponent is less than 1, and therefore is acceptable.

Table 3.k.2-1: Interaction Ratio for the Strainer Modules (Units 1 and 2)

Governing Interaction Strainer Component Calculated Allowable load case Ratio External Radial Stiffener 10.60 ksi 21.24 ksi (1) SSE 0.95 (Including Debris Stops) (flexure) (1)

Tension (Connecting) 11.20 ksi 17.00 ksi (1) SSE 0.94 Rods (axial) (1) (2) 5.29 ksi Edge Channels (max) 14.16 ksi (1) SSE 0.80 (flexure) (1)

Seismic Stiffeners 3.24 ksi 17.70 ksi (1) OBE 0.92 (3)

(including Support legs) (flexure) (1) 4.97 ksi Spacers 12.95 ksi (1) OBE 0.54 (axial) (1)

Core Tube (Biggest 0.69 ksi 20.64 ksi OBE 0.03 Holes)

Perforated Plate (DP 24.48 ksi 25.80 ksi OBE 0.95 Case)

Perforated Plate (Seismic 10.78 ksi 30.96 ksi SSE 0.35 Case)

Perforated Plate (Edge 3.82 ksi 25.80 ksi OBE 0.14 Channels)

Perforated Plate (Inner 11.54 ksi 25.80 ksi OBE 0.45 Gap)

Wire Stiffener 33.75 ksi 48.75 ksi OPR 0.69 Perforated Plate (Core 6.88 ksi 14.16 ksi OBE 0.49 Tube End Cap DP Case) 433 lb 2853 lb Perforated Plate (Core (shear) (shear)

Tube End Cap Seismic SSE 0.72 1329 lb 2351 lb Case)

(tension) (tension)

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02 Governing Interaction Strainer Component Calculated Allowable load case Ratio Radial Stiffening Spokes of the End Cover 1105 lb/in 3540 lb/in OBE 0.31 Stiffener Circumferential Rings of 17.591 ksi 21.24 ksi SSE 0.82 the End Cover Stiffener 215 lb 2853 lb (shear) (shear)

End Cover Sleeve SSE 0.64 1329 lb 2351 lb (tension) (tension)

Welds of End Cover 6846 lb/in 7965 lb/in SSE 0.86 Weld of Radial Stiffener 4.46 ksi 9.46 ksi SSE 0.47 to Core Tube Weld of Radial Stiffener 1.34 ksi 2.65 ksi OBE 0.51 to Seismic Stiffener Edge Channel Rivets 95.75 lb 782 lb OBE 0.13 Inner Gap Hoop Rivets 75 lb 782lb OBE 0.09 Mounting Pins 2.36 ksi 12.26 ksi SSE 0.19 Clevis Hitch Pins 8.05 ksi 12.26 ksi SSE 0.66 Angle Iron Tracks 18.54 ksi 21.24 ksi SSE 0.87 1850 lb 2381 lb Expansion Anchors to (tension) (tension)

SSE 0.97 Floor 834 lb 4294 lb (shear) (shear)

Angle Iron-to-Angle Iron 1.0 ksi 3.98 ksi SSE 0.25 Track Weld Module-to-module Sleeve 3.96 ksi 21.24 ksi SSE 0.19 Module-to-module Latch 812 lb 987 lb SSE 0.82 Connection 3.82 ksi Lift Case 14.16 ksi (1) LIFT 0.26 (flexure) (1) 2.42 ksi Outage Case 14.16 ksi (1) Outage 0.19 (flexure) (1)

Notes:

(1)

Calculated and allowable values were calculated for the most governing stress components (i.e., axial, flexure, etc.) per AISC manual 9th Edition; (2)

The 10% over-torque effect was not included in the calculated value; (3)

Interaction ratio was calculated based on the slenderness ratio, not by the strength.

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02 Connecting Piping and Supports for Unit 1 As documented below, the interaction ratio of each subcomponent is less than or equal to 1, and therefore is acceptable.

Table 3.k.2-2: Interaction Ratio for the Connecting Piping and Supports (Unit 1)

Governing Interaction Component Calculated Allowable Load Case Ratio B Strainer Pipe 2.813 ksi 20.64 ksi SSE 0.14 Flanges Flange Bolting (At Sole 1.738 in2 2.27 in2 SSE 0.77 Plate)

Flange Bending (At Sole 16.34 ksi 17.20 ksi OBE 0.95 Plate)

Flange Weld to Pipe 2.31 ksi 9.46 ksi SSE 0.24 Missing Bolts Flange Bolts 0.133 in2 0.14 in2 SSE 0.93 Flange Bending 17.20 ksi 17.20 ksi SSE 1.00 Sole Plate Connection LLRT Sole Plate 4.82 ksi 17.70 ksi 0.27 Testing 2322 lb 3133 lb Sole Plate Expansion (tension) (tension) LLRT 0.84 Anchors 156 lb 1546 lb Testing (shear) (shear)

Type PS1/PS2 Restraint 15.721 ksi 21.24 ksi (flexure) (flexure)

Angle Normal Stress SSE 0.80 0.522 ksi 8.06 ksi (axial) (axial)

Angle Shear Stress 1.484 ksi 11.80 ksi SSE 0.13 1583 lb 2160 lb Expansion Anchors (Type (tension) (tension)

SSE 0.78 PS1) 209 lb 4157 lb (shear) (shear) 1860 lb 2344 lb Expansion Anchors (Type (tension) (tension)

SSE 0.87 PS2) 225 lb 2614 lb (shear) (shear)

Baseplate 15.045 ksi 17.70 ksi SSE 0.85 Weld of Angle to Baseplate 2352 lb/in 3977 lb/in OBE 0.59 E3-142

Enclosure 3 Updated Final Responses to Generic Letter 2004-02 Governing Interaction Component Calculated Allowable Load Case Ratio Saddle Plate Bending 2.91 ksi 21.24 ksi SSE 0.14 Saddle Plate Shear 8.685 ksi 11.80 ksi SSE 0.74 Saddle Plate Welds 81 lb/in 497 lb/in OBE 0.16 Saddle Plate Pins 5.58 ksi 18.73 ksi SSE 0.30 Shear Lugs 0.928 ksi 11.80 ksi SSE 0.08 Integral Welded Attachments 6.045 ksi 20.64 ksi OBE 0.29 Type PS3 Restraint 2.798 ksi 9.814 ksi W6x15 Normal Stress OBE 0.22 0.180 ksi 14.16 ksi W6x15 Shear Stress 0.683 ksi 9.44 ksi SSE 0.07 1063 lb 2632 lb (tension) (tension)

Expansion Anchors SSE 0.49 235 lb 2604 lb (shear) (shear)

Baseplate 7.795 ksi 17.70 ksi SSE 0.44 Weld of W6x15 to Baseplate 393 lb /in 3977 lb/in SSE 0.10 Angle Normal Stress 11.871 ksi 15.58 ksi SSE 0.76 Angle Shear Stress 2.155 ksi 9.40 ksi SSE 0.23 Weld of Angle to W6x15 1329 lb/in 2983 lb/in SSE 0.45 U-Bolt Normal Load 3.651 ksi 12.90 ksi SSE 0.28 Type PB1 Restraint Stanchion Plate Bolts 1.574 ksi 18.73 ksi SSE 0.08 Integral Welded Attachments 2.37 ksi 20.64 ksi OBE 0.11 Other Pipe Components Slip Joint 313 lb 438 lb OBE 0.71 Connecting Piping and Supports for Unit 2 As documented below, the interaction ratio of each subcomponent is less than 1, and therefore is acceptable.

Table 3.k.2-3: Interaction Ratio for the Connecting Piping and Supports (Unit 2)

Governing Interaction Component Calculated Allowable Load Case Ratio A Strainer Pipe 1.152 ksi 17.20 ksi NORM 0.07 B Strainer Pipe 1.434 ksi 20.64 ksi OBE 0.07 Flanges E3-143

Enclosure 3 Updated Final Responses to Generic Letter 2004-02 Governing Interaction Component Calculated Allowable Load Case Ratio Flange Bolting (At Sole 1.734 in2 2.18 in2 SSE 0.79 Plate)

Flange Bending (At Sole 15.308 ksi 17.20 ksi SSE 0.89 Plate)

Flange Weld to Pipe (In-line 3.85 ksi 9.46 ksi SSE 0.41 Flanges)

Missing Bolts Flange Bolts 0.134 in2 0.136 in2 SSE 0.98 Flange Bending 16.33 ksi 17.20 ksi SSE 0.95 Sole Plate Connection LLRT Sole Plate 4.82 ksi 17.70 ksi 0.27 Testing 2322 lb 3133 lb Sole Plate Expansion (tension) (tension) LLRT 0.89 Anchors 203 lb 1381 lb Testing (shear) (shear)

Pipe Supports 7.7 ksi 21.24 ksi (flexure) (flexure)

Angle Normal Stress SSE 0.45 0.689 ksi 8.804 ksi (axial) (axial)

Angle Shear Stress 0.565 ksi 11.80 ksi SSE 0.05 1614 lb 1850 lb (tension) (tension)

Expansion Anchors SSE 0.93 118 lb 2209 lb (shear) (shear)

Baseplate 12.397 ksi 21.24 ksi SSE 0.58 Weld of Angle to Baseplate 891 lb/in 2983 lb/in OBE 0.30 Saddle Plate Bending 2.442 ksi 21.24 ksi SSE 0.11 Saddle Plate Shear 0.802 ksi 11.80 ksi SSE 0.07 Saddle Plate Welds 156 lb/in 497 lb/in OBE 0.31 Saddle Plate Pins 6.293 ksi 18.73 ksi SSE 0.34 E3-144

Enclosure 3 Updated Final Responses to Generic Letter 2004-02 Debris Interceptors As documented below, the interaction ratio of each subcomponent is less than 1, and therefore is acceptable.

Table 3.k.2-4: Interaction Ratio for the Debris Interceptors Interaction Component Type Calculated Allowable Ratio Debris Interceptor Panel B 125.63 lb*ft/ft 343.75 lb*ft/ft 0.37 2335 psi 17.7 ksi (bending - y) (bending -y)

Support Beams (W6x9) OBE B 0.92 12345 psi 15.576 ksi (bending - x) (bending - x) 2420 psi 21.24 ksi (bending - y) (bending - y)

Support Beams (W6x9) SSE B 0.77 12913 psi 21.24 ksi (bending - x) (bending - x)

Beam Splice Connection B 18977 psi 21.24 ksi 0.89 F879 Type 302HQ, CW B 871 lb 3408 lb 0.26 Machine screws (shear)

Flange Splice Weld B 7.07 in 11.25 in 0.63 661 lb 1359 lb (Tension) (Tension)

Anchor Bolts (1/2 dia. HKB3) B 0.82 661 lb 1987 lb (Shear) (Shear)

Angle 4x4x3/8 OBE B 15112 psi 17700 psi 0.85 Angle 4x4x3/8 SSE B 17033 psi 21240 psi 0.80 Curb Plate (11 GA) C 17013 psi 17700 psi 0.96 Support Angle (L6x6x1/4) C 1298 psi 17700 psi 0.07 20.23 lb 633 lb (Shear)

(Shear)

Anchor Bolts (1/4 dia. HKB3) C 660 lb 0.06 17.4 lb (Tension)

(Tension)

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:

Containment sump recirculation is used when makeup to the RCS is required, and other sources are not available or are of such small volume as to be insufficient. The E3-145

Enclosure 3 Updated Final Responses to Generic Letter 2004-02 license and design bases of PBN only credit containment sump recirculation following a LOCA (Reference 27, Enclosure 1, pg. 34). Since sump recirculation is not credited following other potential high energy line breaks (HELBs) such as feedwater or main steam line breaks, the potential dynamic effects of a HELB were not evaluated for the replacement strainers.

In Safety Evaluations dated June 6, 2005, (Reference 30) November 7, 2000, (Reference 31) (and supplemented on February 7, 2005, (Reference 32)), December 15, 2000, (Reference 33) (also supplemented on February 7, 2005), and December 18, 2000, (Reference 34) the NRC reviewed and accepted analyses demonstrating that a rapidly propagating failure of the large bore RCS piping components at PBN is highly unlikely (leak before break analyses). These analyses included the RCS primary loop piping, SI accumulator discharge lines to the RCS, the pressurizer surge line, and the high pressure RHR piping connections to the RCS. As such, consideration of missile impacts or other dynamic effects of a LOCA per 10 CFR 50 General Design Criterion 4 (Plant specific GDC 40) are no longer part of the design bases for PBN.

The replacement screens have been located outside of the thick walled reactor coolant loop compartments and are away from openings in the walls to the extent practicable. The strainers are also inherently robust, owing to the tough and relatively thick material used for the strainer active surfaces (18 gauge stainless steel), the internal reinforcements to prevent deformation under the design differential pressure, the convoluted form that precludes large, unbroken diaphragm surfaces, and the external bracing for seismic loading. As such, they are unlikely to tear or be perforated by incidental impacts from debris or rebounding missiles, tending rather to deform or dent. The strainers in each unit are routed away from each other such that no single missile would be capable of impacting both strainers.

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:

Back flushing of the sump strainers is not credited in the PBN analysis; therefore, no structural analysis considering reverse flow was performed.

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02

l. 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)(iv)

The basis for concluding that the water inventory required to ensure adequate ECCS or CSS 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.l.1:

The following areas / items were considered as part of the evaluation to determine potential choke points for flow upstream of the sump:

x Refueling Canal x Steam Generators x Annulus in Lower Containment x Reactor Cavity (reactor cavity breaks only) x Containment Spray Washdown Refueling Canal The refueling canal at both PBN1 and PBN2 is drained by one 4-inch pipe that exits the refueling canal at the floor of the canal.

The entrance to the drain at each unit is covered by a strainer that has two hundred 1-inch diameter holes and sits over the cavity outlet. Figure 3.l.1-1 shows the construction of the refueling canal drain strainer. The strainer consists of a vertical cylinder with a 10 nominal diameter and two horizontal cylinders with a 6 nominal diameter.

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02 Figure 3.l.1-1: Refueling Canal Strainer Any sprays falling directly in the refueling canal must flow through the refueling canal drain. It is possible that some debris could accumulate on and around the strainer over the drain. An evaluation was performed that verified water will flow freely through the refueling canal drain and the drain screen would not become obstructed with debris.

Further supporting this evaluation is the refueling canal drain blockage testing that was performed for Turkey Point Nuclear Plant. This testing not only tested the refueling cavity drains to determine if they would become blocked by post-accident debris, but also tested the behavior of the debris that was assumed to be blown into the refueling canal as a result of the LOCA. This portion of the testing demonstrated that it requires containment spray flow rates significantly greater than the expected flow rates to cause debris to move of a size that would challenge the ability of the PBN refueling canal drain strainer to provide the assumed flow out of the refueling canal.

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02 Even if debris built up around the bottom of the strainer, there would still be sufficient flow area to meet containment water volume analysis assumptions for containment water level.

Steam Generator Compartments The steam generator compartments for both units do not have a significant amount of grating and have significant open area between potential break locations and the containment floor. Therefore, break and spray water in the steam generator compartments would drain down to the containment pool with limited obstruction.

At the base of each steam generator compartment in PBN1, there are five passageways that communicate with the annulus, three of which have debris interceptors installed. If debris were to block these passageways, water would simply flow out of the steam generator compartments into the annulus through the two open passageways where debris interceptors are not installed (see the Response to 3.l.3 for more information).

Annulus in Lower Containment In the annulus compartment in lower containment in each unit, the containment geometry is not compartmentalized. Therefore, there are no potential upstream blockage points in the annulus.

Reactor Cavity For breaks in the reactor cavity (at the reactor vessel nozzles), the reactor cavity would fill first before the recirculation sump. A modification was performed that bored a 16 diameter hole in each unit to allow the two sumps to communicate with each other, which eliminates the potential chokepoint and hold up of water for reactor cavity breaks.

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, and there are unobstructed stairways that allow the water to pass.

A complete evaluation of the containment CAD model, along with a review of the CFD model, indicated no significant areas that would become blocked with debris and hold up water during the sump recirculation phase.

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02

2. Summarize measures taken to mitigate potential choke points.

Response to 3.l.2:

There have been several modifications at PBN1 and PBN2 to mitigate potential choke points. The installation of the refueling canal drain strainer was completed prior to Generic Letter 2004-02. In PBN1, two debris interceptors have been removed from the annulus at the 8 Elevation to mitigate potential upstream blockage points in the annulus. A 16 diameter hole was bored through the walls of the incore wall in each unit to allow communication between the reactor cavity and the recirculation sump for reactor cavity breaks.

3. Summarize the evaluation of water holdup at installed curbs and/or debris interceptors.

Response to 3.l.3:

Debris interceptors were installed for PBN1 only. These debris interceptors are comprised of stainless steel bar grating with attached stainless steel perforated plate on one side (16 gage ASTM A240, Type 304 stainless steel with 1/4 diameter holes, approximately 58% open area). The bar grating consists of 1 x 1/8 bearing bars, spaced 1-3/16 center to center, and cross bars spaced 4 center to center. If debris were to block these passageways, water will simply flow out of the steam generator compartments into the annulus through the two open passageways where debris interceptors are not installed.

Holdup was evaluated in the pressurizer cubicle. There are twelve vent holes located in the pressurizer bottom skirt, whose centerline lies 2-3/8 above the pressurizer cubicles floor elevation. The amount of hold-up in the pressurizer compartment is 313 ft3.

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.l.4:

As discussed in the Response to 3.l.1, the entrance to the refueling canal drains at each unit is covered by a strainer that has 200 1-inch diameter holes and sits over the cavity outlet. Any sprays falling directly in the refueling canal will flow through the refueling canal drain. In the evaluation of the drain strainer, it was assumed that some blockage of the lower holes will occur resulting in a ponded volume of water that is held up. The calculated amount of holdup volume is 438 gallons.

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02

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 ECCS and CSS 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 ECCS and CSS 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 screens 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 NRC-approved methods were used (e.g., WCAP-16406-P-A with accompanying NRC 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:

PBN developed a calculation to address ex-vessel (i.e., component and systems) downstream effects. The calculation was developed in accordance with PWROG WCAP-16406-P-A, Revision 1. The limitations and conditions provided in the NRC SER were addressed as part of the evaluations and it was shown that the WCAP-16406-P-A methodology was appropriate for use at PBN.

The following methodology was employed in the ex-vessel downstream effects evaluation. The evaluation did not use any unapproved methods or take any exceptions to NRC-approved methods.

Maximum Debris Ingestion Determination Blockage and wear of the ECCS and CSS components and piping in the post-LOCA recirculation flowpaths downstream of the sump screen were addressed within the downstream effects evaluations. The adequacy of the sump screens mesh spacing E3-151

Enclosure 3 Updated Final Responses to Generic Letter 2004-02 or strainer hole size (nominal hole diameter of 0.066 inches as described in the Response to 3.j.1) was conservatively addressed by assuming that the maximum amount of particulate debris transports to the strainers and 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 actual plant value of 86 g/FA, evaluated from the large scale fiber penetration testing data. The ex-vessel downstream effects evaluations were based on this maximum amount of ingested debris (see Initial Debris Concentrations below).

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 amounts of particulate debris transported to the strainer that are small enough to penetrate the strainer were assumed to pass through the strainer. The total maximum initial debris concentration was determined to be 1,991.9 ppm, with fiber debris contributing 22.5 ppm, and particulate and coating debris contributing 1,969.4 ppm (1,991.9 ppm - 22.5 ppm) .

Flowpaths and Alignment Review Both trains of the ECCS 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) and other plant design documents as applicable.

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:

x Identifying all components in the ECCS and CSS flowpaths (see Flowpaths and Alignment Review above).

x Applying the appropriate wear models for pumps. Pumps experience erosive wear and abrasive wear due to debris ingestion. Two abrasive wear models were developed in WCAP-16406-P-A including a free flowing abrasive wear model and the Archard abrasive wear model. Each model was used as appropriate in the evaluations.

x Applying the appropriate erosive wear model for heat exchangers, orifices, spray nozzles, system piping, and valves.

x Evaluating the potential for plugging of heat exchanger tubes, orifices, spray nozzles, system piping, and valves by comparing the maximum debris size E3-152

Enclosure 3 Updated Final Responses to Generic Letter 2004-02 expected to be ingested through the sump screen to the clearances within the components.

x Evaluating the potential for debris sedimentation for system piping, heat exchanger tubes, and valves that move or reposition post-LOCA (and must go fully closed) by comparing line velocity to minimum line velocity required to avoid sedimentation (line velocity greater than 0.42 ft/s) .

x Evaluating the potential for debris collection in the instrument sensing lines.

2. Provide a summary and conclusions of downstream evaluations.

Response to 3.m.2:

The following is the summary of results and conclusions of the downstream effects evaluations:

ECCS/CSS Pumps The evaluation for pumps addressed the effects of debris ingestion through the sump screen on three aspects of operability (hydraulic performance, mechanical-shaft seal assembly performance, and mechanical performance). The hydraulic and mechanical performances of the ECCS and CSS pumps were determined to not be negatively affected by the recirculating sump debris. Based on the mechanical shaft seal assembly evaluation, the performance of the RHR and SI pump mechanical shaft seals were determined to be satisfactory with regard to the debris laden fluid following the postulated LOCA for the mission time of 30 days. The mission time of the CSS pumps is 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. The performance of the CSS pump mechanical shaft seals were determined to be satisfactory during this mission time as well.

ECCS/CSS Valves WCAP-16406-P-A provides the criteria for wear and plugging analysis for ECCS and CSS valves due to debris laden fluid. Table 3.m.2-1 and Table 3.m.2-2 contain a summary of the criteria that would necessitate an evaluation. The valves that do not meet these criteria are not critically impacted by wear and plugging due to debris laden fluid.

Table 3.m.2-1: Valve Evaluation Blockage Criteria Valve Type Size (inches) Position During the Event Gate 1 Open Globe 1-1/2 Open Globe > 1 (Cage Guide) Open Check Valves/ Stop Check 1 Open Butterfly <4 Throttled < 20° Globe Valves All Throttled E3-153

Enclosure 3 Updated Final Responses to Generic Letter 2004-02 Hermetically Sealed Valves All Open Table 3.m.2-2: Valve Evaluation Erosive Criteria Valve Type Size (inches) Position During the Event Globe All Throttled Butterfly All Throttled Note that if a valve is intended to be throttled during accident mitigation, it shall be evaluated for erosive wear regardless of valve type. Based on this criteria, the valve population at PBN was reviewed and the individual valves were classified as Not Critical or Evaluation Required. The 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.

Valves were evaluated for blockage in the downstream effects evaluations. It was determined that all valves passed the acceptance criteria for the blockage evaluation.

Valves were evaluated for debris sedimentation. The line velocities for all valves analyzed were found to be greater than 0.42 ft/s; thus, debris sedimentation was not an issue.

Valves were evaluated for erosive wear. The initial debris concentration of 1,991.9 ppm was used to calculate the initial wear rate and was assumed to remain constant for most of the valves. A few valves required that the debris depletion refinement be implemented. The assumed large debris was depleted over time using a depletion coefficient of = 0.07, as recommended by WCAP-16406-P-A. Note that large debris consists of particulates greater than or equal to 100 m, coatings greater than or equal to 400 m, and all fibers. The new wear rate was calculated each hour for a total of 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br />. It was found that the increase in valve flow area due to erosion for all valves that were evaluated is considered negligible. The limiting valves for erosive wear at PBN are the reactor vessel injection gate valves, 1&2-SI-852-A/B. The total wear of 6.76 mil resulted in a maximum change in flow area of 2.75%, which is within the acceptance criteria of 3%.

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 1,991.9 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 PBN heat exchangers, orifices, spray nozzles, and system piping in the ECCS recirculation flow path is 0.375 inches, for the CS Nozzles.

The maximum diameter of downstream debris was conservatively assumed to be E3-154

Enclosure 3 Updated Final Responses to Generic Letter 2004-02 0.0726, which is 110% of the sump screen hole size. Therefore, no blockage of the ECCS flow path is expected with the maximum debris diameter of 0.0726 inches.

System piping and heat exchanger tubing 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 acceptance criteria for plugging due to sedimentation.

ECCS/CSS Instrumentation Tubing Instrumentation tubing (or sensing lines) was evaluated for debris settling. According to WCAP-16406-P-A, Section 8.6.6, instrument tubing is designed to remain water solid without taking flow from the process stream. This prevents direct introduction of debris laden fluid into the instrument tubing. Settling of the debris is the only process by which the debris is introduced into the instrument tubing. Since the sensing lines are water solid and stagnant, the introduction of either fibrous or particulate debris by flow into the sensing lines is not possible. The terminal settling velocities of the debris sources in the process streams are small by comparison to the process fluid velocities; therefore, introduction of debris by settling into the instrument tubing is not expected.

Furthermore, the plant walkdowns showed that all instrument taps into the process piping are from the horizontal position to the upper half of the piping. This excludes the possibility of debris settling in the subjected instrument tubing. Therefore, blockage and wear of ECCS or CSS instrument tubing due to debris laden fluid are not expected.

3. Provide a summary of design or operational changes made because of downstream evaluations.

Response to 3.m.3:

There have been no design or operation changes made because of downstream evaluations.

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02

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:

NextEra performed fiber bypass testing for PBN and applied the test results in the new in-vessel downstream effects analysis. The analysis followed the latest NRC staff review guidance on in-vessel effects (Reference 9) and pressurized water reactor owners group (PWROG) guidance (Reference 35), and used the methodology and acceptance criteria in WCAP-17788-P, Revision 1 (Reference 36; 37) for the resolution of in-vessel effects. A summary of the fiber bypass testing and in-vessel analysis and resolution is provided below. It is concluded that post-accident long-term core cooling (LTCC) will not be challenged by accumulation of debris within the reactor core for all postulated LOCAs at PBN. Note that the discussion on the cold leg breaks and LOCADM analysis has been removed from the submittal in accordance with the NRC review guidance (Reference 9 pp. 2-3).

PBN Fiber Penetration Testing PBN conducted fiber penetration testing for both units in 2014. Because of the similarity of the PBN1 and PBN2 strainers, penetration testing for both units was conducted in a single testing program.

The purpose of the PBN testing was to collect time-dependent fiber penetration data for the plant strainer. Three large-scale tests were conducted with test parameters selected to be representative of the most conservative conditions (temperature, debris quantity and composition, and water chemistry). The test results were used to derive a model to quantify fiber penetration for the PBN strainer at plant conditions.

Test Loop Design The test loop included an acrylic test tank, which housed a test strainer at its downstream end, and various piping and flow components. Water was circulated by a pump through the test strainer, a fiber filtering system, and other loop components (see Figure 3.n.1-1). The fiber filtering system consisted of two parallel in-line filter bag housings, which allowed for one filter bag to be online at all times even during swap of filter housings. The bypass line around the filter bags was isolated for the duration of the test. Downstream of the filter bags, a heat exchanger and control E3-156

Enclosure 3 Updated Final Responses to Generic Letter 2004-02 valves allowed for the loop temperature to be controlled. Likewise, the flow elements downstream of the heat exchanger provided the necessary information to allow the pump frequency to be adjusted to achieve the desired flow rates. The test water was returned to the upstream portion of the test tank through the mixing lines and the debris hopper line. One mixing line was placed at the downstream end of the test tank behind the strainer to prevent debris settling, while the remaining mixing lines and debris introduction line were at the upstream end of the tank.

Figure 3.n.1-1: Penetration Test Loop P&ID The test tank had a rectangular geometry, as shown in Figure 3.n.1-2 below. Debris was introduced in the high-agitation region located at the upstream end of the test tank. This region was equipped with two hydraulic mixing lines to create adequate turbulence and prevent the debris from settling. 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 disk faces and adjacent tank walls imitated the module-to-module spacing at the plant. Enough space was left between the strainer and the rear wall to model the open space on both sides of the strainer modules along the length of most strainer assemblies at PBN 1 and PBN 2.

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02 Mixing Lines Core Tube SStrainer Disks Debris Intro Line Flow High Agitation Transport Region Region Figure 3.n.1-2 General Arrangement of Test Tank (Plan view)

The effectiveness of the agitation region for each of the penetration tests is shown 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 agitation or transport 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 #

Added (g) Fiber (g) Added (g) Transport 1 12,946 1,068 11,878 91.8%

2 12,946 933 12,012 92.8%

3 5,178 0 5,178 100%

Test Strainer The test strainer for penetration testing was a prototypical strainer module that matched the key design parameters (i.e., all disks dimensions, including perforated plate thickness, hole diameter, pitch, etc.) of the plant strainer. The test strainer was similar to that used in head loss testing (as described in the Response to 3.f.4) with the only difference that 3 of the 10 disks in a prototypical module were removed. The overall width of the module between the end disks was preserved, and the remaining disks were relocated to evenly fill the gap. Because the strainer modules are flow-controlled, the disk locations along the length of the core tube does not affect the flow distribution among the disks. This modification nearly doubled the gap between adjacent disks to prevent a fiber bridge from forming across adjacent disks. The core tube slots corresponding to the removed disks were covered by gap rings to prevent a path for flow to bypass the remaining perforated disks. This promoted fiber E3-158

Enclosure 3 Updated Final Responses to Generic Letter 2004-02 penetration by preserving penetrable strainer area. The total surface area of the test strainer was 95.1 ft2.

Debris Types and Preparation Each of the three tests used different debris types, with the percentage of each noted in parentheses: Test 1 included Nukon (40.7%) and Mineral Wool (59.3%); Test 2 included Nukon (28.8%), Mineral Wool (67.7%), and Temp-Mat (3.5%); and Test 3 used only Nukon. Fiber debris types at PBN include each of the tested debris types and low-density fiberglass (LDFG) and latent fiber, which were substituted with Nukon due to similarity in characteristics. Only fiber fines were used in penetration testing.

All fiber fines were prepared according to the NEI protocol following the same procedures used for the head loss tests (Reference 41). All fiber types were heat-treated to simulate aging of fiber from hot pipes. Nukon and Mineral Wool were heat-treated by the debris vendor. Temp-Mat was heat treated by the testing vendor by placing an equal mass of untreated Nukon and Temp-Mat in the oven until the Nukon binder was burnt out. Temp-Mat batches were then created with equal parts heat-treated and non-heat-treated Temp-Mat.

Debris was weighed out into batches and was then pressure-washed with test water following the NEI protocol. For batches with multiple fiber types, each type of fiber was prepared separately. The duration of pressure washing was specific to the fiber type and batch size, and was controlled to achieve a fiber size distribution predominately meeting the Class 2 requirements per NUREG/CR-6224 (Reference 42, Appendix B, Table B-3). Figure 3.n.1-1 shows the prepared debris after pressure washing.

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02 Figure 3.n.1-1: Nukon (Top), Mineral Wool (Center), and Temp-Mat (Bottom)

Fines Prepared for PBN Penetration Testing For batches consisting of multiple fiber types, after each debris type was separately pressure washed, the prepared debris was mixed together in a barrel and stirred to form a homogeneous debris slurry prior to introduction.

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02 Debris Introduction Fine fiber debris was introduced to the high agitation region of the test tank via the debris hopper. The prepared debris slurry was transferred from the barrel to the hopper using 5 gallon buckets. During this process, the debris slurry was stirred to promote a homogeneous mixture in the barrel. Additionally, the debris transported into the tank was mixed by the flow turbulence in the hopper and the mixing region of the test tank to break up any agglomeration of fibers that formed. For each batch, the debris introduction rate was controlled to maintain a prototypical debris concentration in the test tank.

For Test 1 and Test 2, debris was introduced in 8 separate batches of increasing sizes, resulting in a total tested fiber load of 524.3 lbm for Test 1 and 530.2 lbm for Test 2 at plant scale. These are equivalent to a theoretical uniform bed thickness of 1.5 .

For Test 3 (Nukon only), debris was introduced in 5 separate batches of increasing sizes, resulting in a total tested fiber load of 228.5 lbm, equivalent to a theoretical uniform bed thickness of 0.5 .

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 in-line filters downstream of the test strainer, upstream of the pump. All of the flow downstream of the strainer travelled through the 1-micron filter bags before returning to the test tank. The capture efficiency of the filter bags was verified to be above 98 percent. The filtering system allowed the installation of two sets of filter bags in parallel lines such that one set of filter bags 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 during testing was used to quantify fiber penetration. After testing, the debris-laden filter bags were rinsed with deionized (DI) water to remove residual chemicals before being dried and weighed. When processing the filter bags, in either clean or debris laden state, the bags were placed in an oven for at least an hour before being cooled and weighed inside a humidity-controlled chamber. This process was repeated for each bag until two consecutive bag weights (taken at least 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 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 online for a minimum of three pool turnovers (PTOs) to capture the prompt fiber penetration. For batches 1 and 3 and the final batch (batch 8 for Tests 1 E3-161

Enclosure 3 Updated Final Responses to Generic Letter 2004-02 and 2, and batch 5 for Test 3), two additional filter bags were used to capture the fiber penetration due to shedding. For each test, the total test duration exceeded the time after the accident when simultaneous cold leg and upper plenum injection takes place.

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. Before each debris addition, the test tank and debris hoppers were visually checked to verify that all introduced debris had transported to the strainer.

Test Parameters The test water used for fiber penetration testing had a chemical composition prototypical to PBN. The plant condition selected for testing was that of the minimum boron concentration of 0.22 mol/l and the maximum pH of 9.404. This condition was represented in testing with a boron concentration of 0.2 mol/l and a pH of 9.5. This water chemistry corresponds to the maximum pH condition at the plant and was chosen based on small scale testing results which showed that water chemistry was not a significant contributor to penetration quantity. Test water was prepared by adding pre-weighed boron to DI water per the prescribed concentration and then adding sodium hydroxide buffer until the prescribed pH was achieved.

A strainer approach velocity of 0.0027 ft/s was determined from plant operating conditions and used for the PBN fiber penetration testing. This velocity was based on the maximum recirculation flow rate per train through a reduced strainer train surface area (1804.6 ft2 vs. total strainer surface area of 1904.6 ft2). Accounting for a reduction in strainer train surface area led to increased approach velocity, which is conservative.

Note that the full strainer surface area was used when applying the penetration test results to determine the total in-vessel fiber load.

Strainer Penetration Model Development Data gathered from the PBN fiber penetration tests were used to develop a model for quantifying the strainer fiber penetration under plant conditions. The model was developed per the following steps:

x 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 equations contain coefficients whose values were determined separately for each test based on the test results.

x The results of one test were curve fit to the governing equations using various optimization techniques to refine the coefficients. This produced a unique set of equations, and thus a unique penetration model for each test. Figure 3.n.1- compares the fiber penetration results of Test 3 (shown as circles) with the fiber penetration quantities determined by applying the Test 3 model to the test conditions (shown as blue solid line). As Figure 3.n.1- shows, the model results adequately represent the test data.

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02 Figure 3.n.1-3: PBN Test 3 Penetration Model Fit Using the methodology outlined above, one fiber penetration model was derived for each test. Because only three specific debris compositions were tested, a comparison of the three models was performed to determine which model is more conservative and can be used to conservatively assess debris penetration for breaks with intermediate debris compositions. Figure 3.n.1-4 shows a comparison of the fiber penetration percentage of added fiber based on measured fiber penetration quantities for all three tests. The figure shows that Test 3 had the highest average penetration percentages out of the three tests.

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02 Figure 3.n.1-4: Average Penetration % vs Fiber Added (Test Scale)

To verify that the Test 3 penetration model results in the most conservative output, the Test 3 model was applied to Test 1 and Test 2 conditions. The results of this application are shown in Figure 3.n.1-5 and Figure 3.n.1-6. These figures show that use of the Test 3 model results in more fiber penetration than that measured from either of the other tests. Because the Test 3 model results in larger fiber penetration quantities than the other two models under identical conditions, it can be concluded that the Test 3 model is the most conservative model for any debris composition.

Therefore, the Test 3 model may be conservatively applied to breaks of intermediate debris compositions (i.e., with a Nukon percentage greater than Test 1 or Test 2) up to the maximum debris load tested among all penetration tests.

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02 Figure 3.n.1-5: Test 3 Model Applied to Test 1 Conditions Figure 3.n.1-6: Test 3 Model Applied to Test 2 Conditions The penetration models from the previous step can be used to determine the prompt fiber penetration fraction and shedding fraction for a given time and amount of fiber accumulated on the strainer. Coupled with a fiber transport model, a time-dependent evaluation can be performed to quantify the total amount of fiber that could pass through the strainer under certain plant conditions.

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02 Figure 3.n.1-7 through Figure 3.n.1-9 show the results of an example application of the Test 3 model to a set of representative plant conditions. Note that these figures are for demonstration purposes, and the results shown in the figures were not used in the PBN in-vessel analysis. The example application did use a time-dependent approach that is similar to that used in the PBN in-vessel analysis. The recirculation duration was divided into smaller time steps. For each time step, the fiber penetration rates and quantities were calculated. Figure 3.n.1-7 shows the resulting cumulative fiber penetration through the strainer over time.

Figure 3.n.1-7: Test 3 Penetration Model at Plant Scale Figure 3.n.1-8 shows the prompt fiber penetration fraction as a function of fiber quantity on the strainer resulting from the example application. As expected, the prompt penetration fraction decreases as a fiber debris bed forms on the strainer.

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02 Figure 3.n.1-8: PBN Prompt Fiber Penetration Fraction Strainer Model Figure 3.n.1-9 shows the shedding rate as a function of time resulting from the example application. Note that shedding penetration depends on the fiber quantity on the strainer and time. As shown in the figure, the shedding rate decreases over time for a given amount of fiber on the strainer.

Figure 3.n.1-9: PBN Shedding Rate Calculated from High Flow Correlation E3-167

Enclosure 3 Updated Final Responses to Generic Letter 2004-02 In-Vessel Effects Evaluations In-vessel downstream effects concern the accumulation of debris inside the reactor core after passing through the sump strainers. The PBN analysis of in-vessel downstream effects was performed outside of NARWHAL. The evaluation follows the methodology of WCAP-17788 (Reference 10) and NRCs review guidance on in-vessel effects (Reference 9). PBN Units 1 and 2 are Westinghouse 2-loop plants. For this plant type, the NRC review guidance requires the following to resolve the in-vessel effects (Reference 9 pp. 10-11):

1. Confirmation that BAP mitigation measures are taken prior to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

2. Confirmation that maximum combined amount of fiber that may arrive at the core inlet and heated core for hot leg break (HLB) is below the WCAP-17788 fiber limit.

The following steps were used to quantify the combined amount of fiber that may arrive at the core inlet and heated core for HLBs.

1. Calculate initial sump pool fiber debris concentration by dividing the transportable fine fiber load by the pool volume. This approach assumes that the transportable debris is well mixed in the sump pool such that a homogeneous mixture is present during the recirculation. This approach is consistent with WCAP-17788 (Reference 10 pp. 5-1).

2. Divide the recirculation duration into small time steps.

3. Calculate the amount of fine fiber that arrives at the strainer during a time step by multiplying the fine fiber concentration in the pool by the strainer flow rate and time step.

4. Calculate the prompt and shedding penetration fractions using the PBN fiber penetration model derived from Test 3. The model defines the prompt and shedding penetration fractions as functions of fiber load on the test strainer. To use this model, the quantity of fiber accumulated on the plant strainer at the beginning of a time step is first scaled down to the test strainer surface area before being substituted into the model.

5. Calculate the amount of prompt penetration for the given time step by multiplying the prompt penetration fraction from Step 4 by the amount of fine fiber arriving at the strainer during the time step from Step 3.

6. Calculate the amount of shedding penetration for the given time step by multiplying the shedding penetration fraction from Step 4 by the amount of fiber collected on the strainer at the beginning of the time step.

7. Split the fiber that passes through the strainer based on the ratio in flow rate between the in-service RHR and SI/CS pumps of each train. The fiber transported by the RHR pump reaches the reactor. The fiber transported by the SI pump flow reaches the RCS cold leg and reactor core inlet. The fiber carried by the CS pump flow is returned to the sump pool. Some conservative simplifications were applied to model the flow split since the goal is to determine the total amount of fiber reaching the reactor core inlet and heated core.

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02 Following an HLB, the RHR pumps are first switched over to sump recirculation and deliver coolant (and debris carried by the coolant) to the reactor core via upper planum injection. During this process, the majority of the flow passes over the top of the core and spills out of the broken hot leg. Only the flow necessary to make up the boil-off enters the top of the core. Therefore, the debris that reaches the reactor core is proportional to the flow split of the boil-off flow rate to the RHR flow rate. The PBN in-vessel analysis conservatively ignored the flow spilling out of the broken hot leg and assumed that all RHR pump flow that reaches the upper plenum injection point enters the reactor core. This simplified approach added more fiber to the reactor early in the event when the strainer is less covered by debris.

Two hours after start of sump recirculation, one SI pump is started to take suction from the RHR pump discharge. In this configuration, the RHR pumps continue to inject into the upper plenum, and the SI pumps inject into the cold legs. The SI pump flow and debris carried by the flow are directed to the core inlet. This flow split between the RHR and SI pumps is not modeled since the purpose of the analysis is to quantify the total amount of fiber that reaches the core inlet and heated core.

8. Update the pool fiber concentration by adjusting the transportable fiber load in the pool for the quantities of fiber loaded onto the strainer, passing through the strainer and returned to the pool via CS (if applicable) during the current time step. The updated concentration will be the initial condition for the next time step.

9. Repeat Steps 3 to 8 shown above until the fine fiber concentration in the sump pool is less than 1% of the initial value (Reference 10 pp. 6-36). The total in-vessel fiber load is calculated by summing up the amount of fiber that reaches the reactor during all time steps.

10. Divide the total in-vessel fiber load by the number of fuel assemblies, which is 121, and increase the result by 2.72% to account for uncertainties in the fiber penetration model before comparing it with the acceptance limit.

The analysis was conducted for different pump lineups: all pumps in operation, single train failure and failure of both CS pumps. Sensitivity runs were performed to ensure the worst combination of input parameters (e.g., pool volume, transport fiber load, RHR and CS pump flow rates, and CS duration) are used for each case.

The steps shown above were implemented in an Excel spreadsheet. The evaluation followed a deterministic approach using bounding combinations of inputs. Table 3.n.1-2 summarizes the operating scenarios analyzed for in-vessel downstream effects and the resulting total in-vessel fiber loads.

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02 Table 3.n.1-2: Summary of In-Vessel Analysis Results RHR Pump CS Pump Pool Total In-Vessel Case Flow Rate No. of RHR Flow Rate Volume Fiber Load No. Pumps (gpm) (gpm) (gallons) (g/FA) 1 2100 2 0 149,794 84.10 2 2100 2 0 277,528 84.95 3 1560 2 0 277,528 85.48 4 2310 2 0 277,528 84.79 5 2100 2 1200 277,528 83.83 6 2310 2 1200 277,528 83.89 7 2310 1 1200 277,528 49.19 As shown in the above Table, the highest total in-vessel fiber load resulted from the case with both RHR pumps but no CS pumps in operation. The pool volume and RHR pump flow rate had insignificant impact on the results. All cases shown in the table used a total transportable fiber fine debris load of 550 lbm, which bounded all postulated breaks of both PBN units.

The calculated maximum in-vessel fiber load is much lower than the WCAP-17788 limit for Westinghouse 2-loop plants.

Full sump recirculation (with both RHR and CS pumps operating in recirculation mode) starts 5200 to 9200 seconds (1.4 - 2.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />) after initiation of an accident. Two hours after start of full recirculation, the PBN EOP directs the operators to secure the CS pump and start one SI pump to deliver coolant to the reactor core inlet to mitigate any potential accumulation of concentrated boric acid inside the reactor. As a result, PBNs BAP mitigation measures are taken approximately 3 to 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> after the accident, which meets the requirement in the NRC review guidance that BAP mitigation measures are to be taken prior to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

In summary, the deterministic evaluation shows that PBN meets the requirements in the NRC review guidance for resolving the in-vessel downstream effects. Therefore, reactor core failures due to in-vessel effects will not contribute to the PBN risk quantification.

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02

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.

Response to 3.o.1:

The chemical effects strategy for PBN1 and PBN2 includes:

x Quantification of chemical precipitates using the WCAP-16530-NP-A methodology. The limitations and conditions of this WCAP were addressed as part of the evaluation, and it was shown that the WCAP-16530-NP-A methods and values were appropriate to use for PBN.

x Introduction of those pre-prepared precipitates in prototypical strainer head loss testing.

x Application of an aluminum solubility correlation to determine the maximum precipitation temperature.

x Time-based determination of acceptable head losses.

As discussed in the Response to 3.a.1, PBN1 and PBN2 have determined the debris generated at all ISI welds upstream of the first isolation valve on the primary RCS piping inside containment. These debris loads are important inputs for chemical effects analysis. For PBN, chemical precipitate was first quantified in a hand calculation using bounding inputs. This analysis informed the strainer head loss testing.

For risk quantification, break-specific analyses were performed in NARWHAL for each of the postulated breaks using the amount of LOCA generated debris for that respective break location. Other plant-specific inputs such as pH, temperature, aluminum quantity, and spray times were selected to maximize the generated amount of precipitates. The calculated chemical debris amount was used to determine its impact on strainer head loss using a head loss lookup table derived from PBN strainer head loss testing data, as discussed in Section 4.7.1 of Enclosure 4.

For PBN strainer head loss testing, SAS was prepared according to the WCAP-16530-NP-A recipes and was verified to meet the settling criteria within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 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.

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02

2. Content guidance for chemical effects is provided in Enclosure 3 dated March 2008 to a letter from the NRC to NEI.

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 5). PBN responses to the GL supplemental content evaluation steps are summarized below. The numbering of the following subsections to the Response to 3.o.2 follow the numbering scheme provided in Section 3 and Figure 1 of the March 2008 guidance (Reference 5, pp. 8-23). Figure 3.o.2.22-1 (provided at the end of the Response to 3.o) highlights the PBN chemical effects evaluation process using the flow chart in Figure 1 of the March 2008 guidance (Reference 5, p. 8).

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.

Response to 3.o.2.1:

PBN is not crediting clean strainer area to perform a simplified chemical effects analysis. See Figure 3.o.2.22-1.

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 consideration of chemical effects.

However, break location 2, with chemical effects 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.

Response to 3.o.2.2:

Head loss testing was performed for PBN that determined the maximum expected head loss considering the effects of conventional debris and chemical debris. As discussed in the Response to 3.f.4, four strainer head loss tests were conducted for PBN with different debris combinations.

Chemical precipitate was added to these tests as described in the Response to 3.f.4. See the Response to 3.f.10 for additional chemical head loss information.

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02

3. Plant-Specific Materials and Buffers: Licensees should provide their assumptions (and basis for the assumptions) used to determine chemical effects loading: pH range, temperature profile, duration of containment spray, and materials expected to contribute to chemical effects.

Response to 3.o.2.3:

The chemical model requires a number of plant-specific inputs. Each input was chosen to maximize the calculated quantity and minimize the solubility (aluminum only) of the chemical precipitates.

PBN uses sodium hydroxide (NaOH) to buffer the post-LOCA containment sump pool to a final pH between 7.0 and 9.5. The injection spray, which delivers NaOH to the containment sump pool, has a maximum pH of 10.5. The pH values used for chemical release were conservatively high, and the pH value used for aluminum solubility was conservatively low. Different pH values for release and solubility were combined in a non-physical way, bounding the effects of all potential pH profile variations.

To determine chemical release rate from unsubmerged aluminum, the maximum spray pH of 10.5 was used. In the sump pool, acids generated through radiolysis may decrease the sump pool pH over the 30-day post-LOCA event. The net effect of these acids over 30 days is conservatively bounded in the base NARWHAL model by a decrease in pH from 9.5 to 8.25 at PBN Containment Sump Pool conditions. The pH values used in the base NARWHAL model are summarized in Table 3.o.2.3-1. Impact on chemical release by sump pH was shown in two parametric sensitivity cases using a lower pH range (8.25 decrease to 7) and a higher pH range (10 decrease to 8.75). These sensitivity cases showed insignificant effect on the risk quantification results.

Table 3.o.2.3-1: PBN pH Values for NARWHAL Base Case Design Input pH Injection Spray pH Used to Determine Chemical Release 10.5 Rates Sump and Recirculation Spray pH Used to Determine 9.5 Chemical Release Rates Sump pH Used to Determine Aluminum Solubility 8.25 The maximum sump pool and containment temperature profiles were used to maximize chemical release rates.

The total amount of concrete assumed to be exposed and submerged in the containment sump pool was 10,000 ft². The quantity of chemical precipitates was negligibly impacted by this large assumed surface area of exposed concrete.

Therefore, exposed concrete is not a significant impact to chemical product E3-173

Enclosure 3 Updated Final Responses to Generic Letter 2004-02 generation in the PBN post-LOCA containment sump pool and is not tracked for this purpose.

The containment sump pool was assumed to be well mixed. This assumption conservatively maximizes aluminum release by not considering the concentration gradient that would form around submerged source materials at low pool velocity conditions.

At PBN1, the total amount of unsubmerged aluminum metal exposed to containment sprays is 306.25 ft² (including contingency). The total amount of submerged aluminum metal exposed to the sump fluid at PBN1 is 16.67 ft² (including contingency). The mass of these unsubmerged and submerged aluminum metals is in excess of the total aluminum released into the containment sump pool, and therefore, no limit was set on the quantity released from these sources. The aluminum coatings on the pressurizer and reactor vessel, which could get exposed during LOCA, were accounted for separately.

At PBN2, the total amount of unsubmerged aluminum exposed to containment sprays is 298.61 ft² (including contingency). The total amount of submerged aluminum exposed to the containment sump fluid at PBN2 is 29.17 ft² (including contingency). The mass of these unsubmerged and submerged aluminum metals is in excess of the total aluminum released into the containment sump pool, and therefore, no limit was set on the quantity released from these sources.

The aluminum coatings on the pressurizer and reactor vessel, which could get exposed during a LOCA, were accounted for separately.

The NARWHAL software analysis accounts for spray mode/duration and the change in water volume with respect to time. The quantities of E-Glass, calcium silicate, mineral wool, and aluminum metal coatings are specific to each analyzed break.

4. Approach to Determine Chemical Source Term (Decision Point): Licensees should identify the vendor who performed plant-specific chemical effects testing.

Response to 3.o.2.4:

PBN is using the separate chemical effects approach to determine the chemical source term. Alden Research Laboratory, Inc. performed the head loss testing in their test lab in Holden, MA.

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 E3-174

Enclosure 3 Updated Final Responses to Generic Letter 2004-02 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).

Response to 3.o.2.5:

PBN used the WCAP-16530-NP-A chemical effects base model to determine the chemical source term. The application of an aluminum solubility correlation to determine a maximum precipitate formation temperature is discussed in the Response to 3.o.2.8 and Response to 3.o.2.9.i.

6. AECL Model:

i. Since the NRC is not currently aware of the complete details of the testing approach, the NRC staff expects licensees using it to provide a detailed discussion of the chemical effects evaluation process along with head loss test results.

Response to 3.o.2.6.i:

This question is not applicable because PBN is not using the AECL model.

See Figure 3.o.2.22-1.

ii. Licensees should provide the chemical identities and amounts of predicted plant-specific precipitates.

x Response to 3.o.2.6.ii:

This question is not applicable because PBN is not using the AECL model.

See Figure 3.o.2.22-1.

7. WCAP Base Model:

i. Licensees proceeding from block 7 to diamond 10 in the Figure 1 flow chart

[in Enclosure 3 dated March 2008 to a letter from the NRC to NEI (Reference 5, p. 8)] 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.

Response to 3.o.2.7.i:

The PBN chemical model quantifies chemical precipitates using the WCAP-16530-NP-A (Reference 22) methodology with the following two deviations:

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02

1. The application of an aluminum solubility correlation to determine a maximum precipitate formation temperature is discussed in the Response to 3.o.2.9.i.

2. The use of the NARWHAL software in place of the WCAP base model spreadsheet. The NARWHAL software follows the WCAP-16530-NP-A methodology for chemical product generation.

An aluminum solubility correlation was used to determine a maximum precipitate formation temperature, which effectively delays the onset of aluminum precipitation. Therefore, to allow for time-based head loss acceptance criteria, the NARWHAL WCAP-16530-NP-A algorithm includes the requirement in the SE to double the aluminum release rate from aluminum metal over the initial 15 days. Additionally, precipitation is forced to occur at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following initiation of the accident even if the solubility limit is not exceeded.

NARWHAL also allows for separate accounting of thick aluminum (not mass limited), thin aluminum (mass limited), and aluminum coatings as a type of debris generated by individual breaks.

ii. Licensees should list the type (e.g., AlOOH) and amount of predicted plant-specific precipitates.

Response to 3.o.2.7.ii:

As stated previously, chemical precipitate quantities were determined in the NARWHAL risk quantification calculation and in a bounding hand calculation.

The NARWHAL calculation performs comprehensive evaluation of GSI-191 phenomena in a self-consistent and time-dependent manner. The chemical debris quantities determined in NARWHAL were used to quantify chemical debris head loss in NARWHAL for each break.

Table 3.o.2.7.ii-1 provides the released aluminum mass, formed precipitates, and maximum aluminum precipitation temperatures that were calculated for multiple cases by the bounding hand calculation. These results were used as inputs for the PBN strainer head loss testing. Note that, per the WCAP-16530-NP-A Safety Evaluation, both aluminum precipitates are acceptable surrogates for aluminum precipitate in head loss testing, and AlOOH, when predicted to form, is converted to the stoichiometric equivalent amount of SAS (based on aluminum) for head loss testing.

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02 Table 3.o.2.7.ii-1: Summary of Precipitate Quantities and Precipitation Temperatures from Bounding Hand Calculation Total Al Al Precipitation Case NaAlSi3O8 AlOOH Released Temperature Case A: PBN1, sump pH of 9.5 - 8.25, 7.9 kg 103.1°F 77.0 kg -

Max Sump Case B: PBN1, sump pH of 8.25 - 7.0, 6.6 kg 143.4°F 63.8 kg -

Max Sump Case C: PBN2, sump pH of 9.5 - 8.25, 8.5 kg 104.0°F 82.7 kg -

Max Sump Case D: PBN2, sump pH of 8.25 - 7.0, 6.8 kg 144.1°F 66.3 kg -

Max Sump Case E: PBN1, sump pH of 9.5 - 8.25, 6.9 kg 101.0°F 67.5 kg -

Pressurizer Compartment, Max Sump Case F: PBN1, sump pH of 8.25 - 7.0, 5.8 kg 148.9°F 56.4 kg -

Pressurizer Compartment, Max Sump Case G: PBN2, sump pH of 9.5 - 8.25, 7.4 kg 102.1°F 71.6 kg -

Pressurizer Compartment, Max Sump Case H: PBN2, sump pH of 8.25 - 7.0, 6.0 kg 148.9°F 58.0 kg -

Pressurizer Compartment, Max Sump Case I: PBN1, sump pH of 8.25 - 7.0, 6.5 kg 146.7°F 63.0 kg -

Min Sump Case J: PBN2, sump pH of 8.25 - 7.0, 6.8 kg 147.3°F 65.6 kg -

Min Sump Case K: PBN1, sump pH of 8.25 - 7.0, 5.8 kg 152.7°F 56.4 kg -

Pressurizer Compartment, Min Sump Case L: PBN2, sump pH of 8.25 - 7.0, 6.0 kg 152.6°F 57.9 kg -

Pressurizer Compartment, Min Sump Case O: PBN1, sump pH of 9.5 - 8.25, 12.1 kg 110.7°F 68.9 kg 11.1 kg Reactor Cavity, Max Sump Case P: PBN1, sump pH of 8.25 - 7.0, 11.1 kg 160.2°F 65.9 kg 9.5 kg Reactor Cavity, Max Sump Case Q: PBN2, sump pH of 9.5 - 8.25, 12.5 kg 111.1°F 121.1 kg -

Reactor Cavity, Max Sump Case R: PBN2, sump pH of 8.25 - 7.0, 11.3 kg 160.1°F 109.4 kg -

Reactor Cavity, Max Sump Case S: PBN1, sump pH of 8.25 - 7.0, 11.1 kg 164.0°F 64.5 kg 9.8 kg Reactor Cavity, Min Sump Case T: PBN2, sump pH of 8.25 - 7.0, 11.3 kg 163.9°F 109.4 kg -

Reactor Cavity, Min Sump

8. WCAP Refinements: State whether refinements to WCAP-16530-NP-A were utilized in the chemical effects analysis.

Response to 3.o.2.8:

Refinement to the model for aluminum solubility is discussed in the Response to 3.o.2.9.i. No other refinements to the WCAP-16530-NP-A methodology were used.

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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.

Response to 3.o.2.9.i:

The base WCAP-16530-NP-A model assumes that aluminum precipitates form immediately upon the release of aluminum into solution. However, as justified in the Response to 3.o.2.7.i, the PBN chemical model includes the following application of an aluminum solubility correlation to determine formation temperature and timing.

The aluminum solubility limit was determined using Equation 3.o.2.9-1, developed by Argonne National Laboratory (ANL).

 

 (Equation 3.o.2.9-1)

 

Nomenclature:

= pH change due to radiolysis acids

 = solution temperature, °F The aluminum solubility limit equation was used to determine the temperature and timing of aluminum precipitation. When precipitation was predicted by this equation, the full amount of aluminum released was assumed to precipitate. The aluminum solubility limit equation was not used to reduce the predicted quantity of precipitate by crediting the amount remaining in solution.

Additionally, precipitation is forced to occur at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following initiation of the accident even if the solubility limit is not exceeded.

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.

Response to 3.o.2.9.ii:

Silicon and phosphate inhibition of aluminum release were not credited. See the Response to 3.o.2.9.i.

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02 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.

Response to 3.o.2.9.iii:

Reductions in precipitate quantity due to residual solubility of aluminum after precipitation occurs was not credited. See the Response to 3.o.2.9.i.

iv. Licensees should list the type (e.g., AlOOH) and amount of predicted plant-specific precipitates.

Response to 3.o.2.9.iv:

The type and amount of plant-specific precipitates are provided in 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.

Response to 3.o.2.10:

As discussed in the Response to 3.o.2.12, PBN 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.

Response to 3.o.2.11.i:

The direct chemical injection method was not used in head loss testing for PBN.

See Figure 3.o.2.22-1.

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.

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02 Response to 3.o.2.11.ii:

The direct chemical injection method was not used in head loss testing for PBN.

See Figure 3.o.2.22-1.

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).

Response to 3.o.2.11.iii:

The direct chemical injection method was not used in head loss testing for PBN.

See Figure 3.o.2.22-1.

12. Pre-Mix in Tank: Licensees should discuss any exceptions taken to the procedure recommended for surrogate precipitate formation in WCAP-16530-NP-A.

Response to 3.o.2.12:

The WCAP-16530-NP-A precipitate formation methodology for SAS was followed with no exceptions.

13. Technical Approach to Debris Transport (Decision Point): State whether near-field settlement is credited or not.

Response to 3.o.2.13:

PBN strainer head loss testing accounted for chemical effects. As stated in the Response to 3.f.12 of this enclosure, no near-field settling was credited during testing. The mixing section of the test tank was equipped with multiple mixing nozzles to provide sufficient turbulence to keep the introduced debris in suspension without disturbing the debris bed formed on the test strainer.

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 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of head loss testing.

Response to 3.o.2.14.i:

PBN is not crediting near-field settlement of chemical precipitate in chemical head loss testing. See Figure 3.o.2.22-1.

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.

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02 Response to 3.o.2.14.ii:

PBN is not crediting near-field settlement of chemical precipitate in chemical head loss testing. See Figure 3.o.2.22-1.

15. Head Loss Testing Without Near Field Settlement Credit:

i. Licensees should provide an estimate of the amount of debris and precipitate that remains on the tank/flume floor at the conclusion of the test and justify why the settlement is acceptable.

Response to 3.o.2.15.i:

Measures taken during the test, as described in the Response to 3.f.12, to keep debris suspended and transportable to the test strainer, prevented notable settling of debris or precipitate.

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 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />).

Response to 3.o.2.15.ii:

The maximum allowable clear volume at the top of a 10 mL sample of SAS precipitates after one hour of settling was 4 mL. The precipitates were continuously mixed and used within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of the execution of a successful settling test.

16. Test Termination Criteria: Licensees should provide the test termination criteria.

Response to 3.o.2.16:

A test was terminated once the head loss had stabilized with the rate of head loss change less than 1% for two consecutive 30 minute periods. The debris bed in this state was characterized using flow sweeps before termination. Figure 3.o.2.16-1 shows an example of the test sequence. Similar figures for the other tests can be found in the Response to 3.f.4.

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02 Figure 3.o.2.16-1: FDL2 Test Chemical Debris Timeline

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.

Response to 3.o.2.17.i:

See the Response to 3.f.4 for the pressure drop curves of the four tests.

ii. Licensees should explain any extrapolation methods used for data analysis.

Response to 3.o.2.17.ii:

Extrapolation methods were not used because the chemical head loss had stabilized before a test was terminated. As shown in Figure 3.o.2.16-1, the measured head loss varied very little over time before the first flow sweep. See the Response to 3.o.2.16.

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02

18. Integral Generation (Alion): Licensees should explain why the test parameters (e.g., temperature, pH) provide for a conservative chemical effects test.

Response to 3.o.2.18:

PBN is using the separate chemical effects approach to determine the chemical source term. This section is not applicable to the PBN chemical effects analysis.

See Figure 3.o.2.22-1.

19. Tank Scaling / Bed Formation:

i. Explain how scaling factors for the test facilities are representative or conservative relative to plant-specific values.

Response to 3.o.2.19.i:

PBN is using the separate chemical effects approach to determine the chemical source term. This section is not applicable to the PBN chemical effects analysis. See Figure 3.o.2.22-1.

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.

Response to 3.o.2.19.ii:

PBN is using the separate chemical effects approach to determine the chemical source term. This section is not applicable to the PBN chemical effects analysis. See Figure 3.o.2.22-1.

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.

Response to 3.o.2.20:

PBN is using the separate chemical effects approach to determine the chemical source term. This section is not applicable to the PBN chemical effects analysis.

See Figure 3.o.2.22-1.

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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.

Response to 3.o.2.21:

PBN is using the separate chemical effects approach to determine the chemical source term. This section is not applicable to the PBN chemical effects analysis.

See Figure 3.o.2.22-1.

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.

Response to 3.o.2.22:

PBN is using the separate chemical effects approach to determine the chemical source term. This section is not applicable to the PBN chemical effects analysis.

See Figure 3.o.2.22-1.

Figure 3.o.2.22-1: Chemical Effects Evaluation Process for PBN (Reference 5

p. 8)

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02

p. Licensing Basis The objective of the licensing basis 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:

The PBN FSAR was updated in 2007 to reflect the containment sump recirculation strainer perforation size for the replacement strainers. In this submittal, PBN uses a risk-informed approach to respond to GL 2004-02, which replaces the current deterministic methodology in the current PBN licensing basis. Adopting this change requires exemptions from certain requirements of 10 CFR 50.46(a)(1), and the request for the exemptions is provided in Enclosure 1. Additionally, an amendment to the PBN Unit 1 and Unit 2 operating licenses is required to incorporate the revised methodology per the requirements of 10 CFR 50.59. The request for this license amendment is in Enclosure 2 of this submittal.

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02

4. NRC Audit Report Responses The NRC performed an audit in January 2019 on the 2017 PBN GL 2004-02 submittal and issued an audit report (Reference 6). Responses to the questions identified in the audit report are provided in the table below.

No. Question from Audit Report Appendix C Response The NRC staff requested that NextEra/FPL justify This question has been that evaluating partial breaks at eight orientations addressed during on-site is adequate to capture debris generation audit. No further response is 1 amounts for all breaks. This issue is discussed in required.

the section on Turkey Point, Question 1. See that item for a discussion of this issue. No further information is needed at this time.

The NRC staff requested that NextEra/FPL verify The debris generation that the cassettes that contain mineral wool are analyses have been re-at least as robust as those tested and found to performed for this submittal have the destruction pressure of 114 psi during and a 5.4D ZOI for all air jet impact testing for boiling-water reactors. mineral wool cassettes is The NRC staff has accepted that cassettes with now used. See the welded (including spot welded) seams can be response to 3.b.3 for more credited with a relatively high destruction detail and the justification pressure, but cassettes with riveted seams have for use of this ZOI size.

a much lower destruction pressure. NextEra/FPL indicated that the cassette information does not specify construction. NextEra/FPL viewed file photos from previous outages. NextEra/FPL stated that Point Beach, Unit 1, cassettes are primarily riveted, and Point Beach, Unit 2, are welded, with riveted repairs for cutouts. The NRC 3

staffs SE on NEI 04-07, accepted 114 psi as the destruction pressure or 2D ZOI (regardless of construction). NextEra/FPL indicated that based on the SONGS submittal, which was reviewed and approved previously, the NRC accepted 114 psi for all Transco and Darmet cassettes, regardless of construction. NextEra/FPL indicated that Point Beach is using a 4D ZOI.

NextEra/FPL believes that the lower destruction pressure only applies to Diamond Power and Mirror type reflective metal insulations. Transco is used at Point Beach. The NRC staff reviewed the SONGS audit and found that the SONGS cassettes are welded and that the report stated that welded cassettes are more robust than riveted. The NRC staff also reviewed air jet E3-186

Enclosure 3 Updated Final Responses to Generic Letter 2004-02 No. Question from Audit Report Appendix C Response impact testing and found corroborating evidence.

The NRC staff concluded that cassettes that are supplied as riveted should be considered to fail at a much lower pressure than welded cassettes.

The NEI 04-07 guidance as accepted by the NRC staffs SE is for RMI, not cassettes with other materials enclosed. Concern with RMI debris is generally much less significant than other insulation types. Lacking additional information, the NRC staff considers only cassettes that are purchased in a welded configuration to be of robust construction such that a reduced ZOI is justified. In cases where the cassettes were purchased as welded, but repaired or modified in the field using rivets, NextEra/FPL should confirm that changes have not significantly affected the structural integrity of the cassette if the smaller ZOI is applied.

Additional information is required for this issue.

The NRC staff requested that NextEra/FPL This question has been confirm that the Cal-Sil installed at Point Beach is addressed during on-site not manufactured using a molding process but is audit. No further response is the rigid pressed type. NextEra/FPL provided required.

information regarding the Cal-Sil. NextEra/FPL 6 also provided documentation on its electronic portal during and after the audit, which the NRC staff reviewed. The documentation supports NextEra/FPLs claim that the Cal-Sil installed at Point Beach is not of the molded type. No further information is needed at this time.

The NRC staff requested that NextEra/FPL The Alternate Break provide justification for not performing transport Methodology (i.e., R-I and cases with two trains in service. For the R-I R-II breaks) is no longer analysis, it is likely that a single pump running is used. In the NARWHAL conservative from a transport perspective model, recirculation because debris is split if two pumps are running. transport of fine debris 7 For R-I, the assumption should be examined to already uses a transport ensure it is valid. For R-II cases that requires fraction of 100%. Erosion of both pumps, additional total debris may transport small and large pieces into due to higher velocities and turbulence in the fine debris is accounted for pool. NextEra/FPL stated that the recirculation in the model. Recirculation transport cases only modeled a single pump transport of small and large running and that the R-II analysis assumed two pieces of debris is not E3-187

Enclosure 3 Updated Final Responses to Generic Letter 2004-02 No. Question from Audit Report Appendix C Response pumps were in service for debris distribution. considered since small and NextEra/FPL stated that the sump level large pieces of debris increases 12 inches within the first recirculation transport by tumbling along (in first 45 minutes after swapover due to CS the floor and the PBN injection). This would reduce the fluid velocities in strainers are installed the pool. NextEra/FPL stated that the strainers several inches off the floor.

are not co-located and that flowpaths are It was observed during separate. NextEra/FPL assumed that 100 testing that small pieces of percent of fines transport and 100 percent of debris pile up in front of the smalls are available to transport. The NRC staff strainer and had little direct reviewed the computational fluid dynamics plots contribution to head loss.

and was able to determine that it is unlikely for Additionally, some additional small debris to transport even with two introduced fine debris and pumps running due to the separation of the chemical debris could be containment areas. Also, the debris interceptors trapped by the small pieces, in Point Beach, Unit 1, appear to be effective for which may result in less eliminating transport of small fiber pieces. Only conservative head loss smalls that blow or wash to the strainer area will testing results. Therefore, in transport. Flow from the other train will not the NARWHAL model, significantly affect the flow in the near strainer recirculation transport of areas. Point Beach, Unit 2, has no debris small and large pieces of interceptors, but a similar argument applies. The debris was not considered.

NRC staff will consider this information in its The recirculation transport evaluation of NextEra/FPLs transport analysis. fraction for fine debris is No further information is needed at this time. already the highest for either two train or single train operation. See Section 4.6 in Enclosure 4 of this submittal.

The NRC staff requested that NextEra/FPL clarify These figures are now why Figure 3.f.7-3 includes a data point with a removed as the Alternate fiber value greater than 200 lbs. This value is not Break Methodology is no shown in Table 3.f.7-4 which lists the bounding longer used.

fiber break as 97 lbs. The NRC staff asked what the values in Tables 3.f.7-3 and 4, and Figure 3.f.7-3 represent. Figure 3.f.7-4 and Table 3.f.7-5 14 also indicate that fine fiber amounts greater than those in the table may be produced and transported. NextEra/FPL explained that the figures are total debris transported to a single strainer. NextEra/FPL also explained that the tables are strainer loadings for R-II breaks assuming half of the debris goes to each strainer.

The values in the tables should be less than half E3-188

Enclosure 3 Updated Final Responses to Generic Letter 2004-02 No. Question from Audit Report Appendix C Response of the figures because some of the debris penetrates the strainer. NextEra/FPL indicated that there are data points in the figures that are outside are R-II break debris amounts although most of R-II breaks are bounded by the test. The NRC staff reviewed the R-II debris amounts to ensure that the tests bounded the values considering judgement on substitutions. The NRC staff will consider the clarifications when evaluating NextEra/FPLs evaluation of R-II breaks. No further information is needed at this time.

The NRC staff requested that NextEra/FPL In the risk-informed provide a justification as to why additional analyses that informed this chemical additions were not made after the final submittal, any breaks that addition in Full Debris Load Test 1 that resulted have chemical debris loads in increased headloss. The NRC staff noted that exceeding the tested reactor cavity breaks are not bounded for quantity were assumed to chemicals. The NRC staff requested an overall result in a strainer failure.

discussion of the chemical effects methodology This failure contributed to and assumptions used including how aluminum the overall risk solubility was credited. NextEra/FPL stated that it quantification. This is performed 20 different chemical analysis cases discussed in the response and that the reactor cavity break had the to 3.f.5.

maximum chemical amounts. Because the fiber amount for the reactor cavity break is small NextEra/FPL does not predict a filtering bed for that case. Therefore, for high fiber cases, 15/24 NextEra/FPL did not add the full chemical amounts. Additional information is required for this issue. The licensee should explain its rationale for the determination that further chemical additions were not made to bound the total potential chemical source term.

In response to Audit Plan question 24, NextEra/FPL stated that two pH cases were evaluated. The cases were for low pH and high pH conditions. NextEra/FPL evaluated the change in temperature (see E1-200 of submittal showing table that compares the conditions).

NextEra/FPL used only the limiting test results to calculate NPSH margin, which is less than 0.14 feet at > 212 °F (100 °C). This is before E3-189

Enclosure 3 Updated Final Responses to Generic Letter 2004-02 No. Question from Audit Report Appendix C Response chemicals which precipitate at approximately 160

°F (71 °C). As the pool cools, NPSH margins increase. When chemicals precipitate, the margin is much greater. The NRC staff agreed that the amount of chemicals used in the high fiber cases was bounding for those breaks and that only the reactor cavity breaks have more chemicals than the tested amounts, but that those breaks result in very little fiber. The NRC staff will use the clarifications from the audit when evaluating the acceptability of the treatment of chemical effects.

No further information is needed at this time.

The NRC staff requested that NextEra/FPL A new approach is used in explain why it is reasonable to use thin-bed test this submittal to evaluate results to determine the head loss for Point strainer head loss. In the Beach, Unit 2, R-II when the test contained less NARWHAL model, debris fiber and Cal-Sil than are predicted to transport to head loss was determined the strainer for the limiting breaks. The NRC staff from a lookup table derived also noted that the test results did not indicate from the testing data of all that a thin bed formed, when conversely, a thin four head loss tests. See 16 bed formed during the confirmatory test that had more detailed discussion in lesser fiber amounts. NextEra/FPL indicated that the response to 3.f.10 and the justification is provided in the paragraph in Section 4.7.1 of below Table 3.f.7-5 of its submittal. The NRC Enclosure 4.

staff will rereview the information in the submittal and determined the acceptability of NextEra/FPLs evaluation of the R-II breaks. This issue is like Question 14 discussed above. No further information is needed at this time.

The NRC staff requested that NextEra/FPL justify As discussed in the that the R-I breaks have acceptable performance response to 3.f.7, flashing for flashing and deaeration. The methodology in and deaeration (i.e., void the submittal is intended for R-II breaks. The fraction) are analyzed in NRC staff accepted method to evaluate flashing NARWHAL for all breaks, and deaeration for R-I breaks is to show and any strainer failures significant margin by maximizing sump due to flashing and 18 temperature and minimizing containment deaeration contribute to the pressure and comparing thermal hydraulic overall risk quantification.

conditions, or showing significant margin using design basis and/or realistic containment analyses. NextEra/FPL stated that it will provide margin to flashing on a single plot by combining data from existing analyses. The NRC staff was unable to locate this information. Additional E3-190

Enclosure 3 Updated Final Responses to Generic Letter 2004-02 No. Question from Audit Report Appendix C Response information is required for the NRC staff to make a conclusion on this issue.

The NRC staff requested that NextEra/FPL As discussed in the describe how deaeration between 0-2 percent is response to 3.g.3, the RG accounted for in the NPSH analysis as 1.82 guidance is used to recommended in RG 1.82. If is it is not accounted adjust the NPSH required for NextEra/FPL should justify that the pumps will based on the void fraction at 19 operate as required with anticipated void the pump suction.

fractions for the duration of the recovery period.

This question is the same as Turkey Point Question 17 and St. Lucie Question 9 discussed above. Additional information is required for this issue.

The NRC staff requested that NextEra/FPL See the response to discuss the statement on page E1-135 regarding Question 16.

R-II breaks being bounded by tested debris loads because it appears to be inconsistent with the headloss section (Cal-Sil not bounded). The responses to Questions 14 and 16 are relevant to this issue and are discussed above. NextEra/FPL 20 indicated that the test, in general, is bounding of all debris loads. The NRC staff noted that Cal-Sil was not specifically bounded. NextEra/FPL indicated that it believes that other particulate in the test makes up for this. The NRC staff will consider this information in its evaluation of R-II breaks. No further information is needed at this time.

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02

5. References

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

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

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

4. NRC Staff Review Guidance (ML080230462). NRC Staff Review Guidance Regarding Generic Letter 2004-02 Closure in the Area of Coatings Evaluation.

March 2008.

5. NRC Staff Review Guidance (ML080380214). NRC Staff Review Guidance Regarding Generic Letter 2004-02 Closure in the Area of Plant-Specific Chemical Effects Evaluation. March 2008.

6. NRC Letter (ML19217A003). Audit Report Regarding Generic Letter 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation during Design Basis Accidents at Presurized-Water Reactors" Closure Methodology (EPID 2017-LRC-0000). December 2, 2019 .

7. NextEra Letter NRC 2017-0045 (ML17363A253). Point Beach Nuclear Plant, Units 1 and 2, Dockets 50-266 and 50-301, Renewed License Nos. DPR-24 and DPR-27, Updated Final Response to NRG Generic Letter 2004-02. December 29, 2017.

8. NEI Guidance Report NEI 04-07 Volume 2. 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. Revision 0.

9. NRC Correspondence ML19228A011. U.S. Nuclear Regulatory Commission Staff Review Guidance for In-Vessel Downstream Effects Supporting Review of Generic Letter 2004-02 Responses. September 4, 2019.

10. WCAP-17788-P, Volume 1. Comprehensive Analysis and Test Program for GSI-191 Closure (PA-SEE-1090). Revision 1.

11. NEI Guidance Report NEI 04-07 Volume 1. Pressurized Water Reactor Sump Performance Evaluation Methodology 'Volume 1 - Pressurized Water Reactor Sump Performance Evaluation Methodology'. December 2004. Revision 0.

12. ENERCON Report. BADGER Orientation and Size Increments used for Debris Generation Calculation (ML16217A084). Revision 1, August 3, 2016.

13. NMC letter NRC 2005-0109 (ML052500302). Nuclear Managerment Company Response to Generic Letter 2004-02, "Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors" for Point Beach Nuclear Plant. September 1, 2005.

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02

14. NRC Revised Guidance Regarding Coatings Zone of Influence for Review of Final Licensee Responses to Generic Letter 2004-02 (ML100960495). Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors. April 6, 2010.

15. Report NEDO-32686-A. Utility Resolution Guide for ECCS Suction Strainer Blockage, Volume 3, Technical Support Documentation. October 1998.

16. ML072740400. U. S. NRC Audit Report "North Anna Power Station Corrective Actions for Generic Letter 2004-02". November 15, 2007.

17. ML080650562. Virginia Electric and Power Company, Surry Power Station Units 1 and 2, Supplemental Response to NRC Generic Letter 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation during Design Basis Accidents at Pressurized Water Reactors. February 29, 2008.

18. NUREG/CR-6772. GSI-191: Separate Effects Characterization of Debris Transport in Water. August 2002.

19. NUREG/CR-6808. Knowledge Base for the Effect of Debris on Pressurized Water Reactor Emergency Core Cooling Sump Performance. February 2003.

20. Nuclear Energy Institute. ZOI Fibrous Debris Preparation: Processing, Storage, and Handling. January 2012. Revision 1.

21. NUREG/CR-6224. Parametric Study of the Potential for BWR ECCS Strainer Blockage Due to LOCA Generated Debris. October 1995.

22. WCAP-16530-NP-A. Evaluation of Post-Accident Chemical Effects in Containment Sump Fluids to Support GSI-191. March 2008.

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

24. NEI Guidance Report NEI 09-10. Guidelines for Effective Prevention and Management of System Gas Accumulation. April 2013. Revision 1a-A.

25. ML070950240. San Onofre Nuclear Generating Station Unit 2 and Unit 3 GSI-191 Generic Letter 2004-02 Corrective Actions Audit Report.

26. ML082050433. Indian Point Energy Center Corrective Actions for Generic Letter 2004-02.

27. FPL letter NRC2008-0013 (ML080630613). Supplemental Response to Generic Letter 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors. February 29, 2008.

28. Nuclear Industry Procedure IP-ENG-001. Standard Design Process. Revision C, September 8, 2016.

29. ASME B&PV Code.Section III, Division 1, Subsections NB, NC, and Appendices.

1998 Edition, through 1999 Addenda.

30. Point Beach Nuclear Plant, Units 1 and 2 (ML043360295). Issuance of Amendments Re: Leak-Before-Break Evaluation for Primary Loop Piping. June 6, 2005.

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02

31. Point Beach Nuclear Plant, Units 1 and 2 (ML003767681). Point Beach Nuclear Plant, Units 1 and 2 - Review of LeakBefore-Break Evaluation for the Accumulator Line Piping as Provided by 10 CFR Part 50, Appendix A, GDC 4 (TAC Nos.

MA7834 and MA7835). November 7, 2000.

32. Point Beach Nuclear Plant, Units 1 and 2 (ML043580008). Supplement to Safety Evaluation on Leak-Before-Break Regarding Correction of Leak Detection Capability. February 7, 2005.

33. Point Beach Nuclear Plant, Units 1 and 2 (ML003777863). Point Beach Nuclear Plant, Units 1 and 2 - Review of Leak-Before-Break Evaluation For The Pressurizer Surge Line Piping as Provided by 10 CFR Part 50, Appendix A, GDC 4 (TAC NOS.

MA7805 and MA7806). December 15, 2000.

34. Point Beach Nuclear Plant, Units 1 and 2 (ML003777964). Point Beach Nuclear Plant, Units 1 and 2 - Review of Leak-Before-Break Evaluation for the Residual Heat Removal System Piping Piping as Provided by 10 CFR Part 50, Appendix A, GDC 4 (TAC NOS. MA7836 and MA7837 . December 18, 2005.

35. PWROG-16073-P. TSTF-567 Implementation Guidance, Evaluation of In-Vessel Debris Effects, Submittal Template for Final Response to Generic Letter 2004-02 and FSAR Changes. February 2020. Revision 0.

36. Westinghouse WCAP-17788-P, Volume 1. Comprehensive Analysis and Test Program for GSI-191 Closure (PA-SEE-1090). Revision 1, December 2019.

37. Westinghouse WCAP-17788-P, Volume 4. Comprehensive Analysis and Test Program for GSI-191 Closure (PA_SEE-1090) - Thermal-Hydraulic Analysis of Large Hot Leg Break with Simulation of Core Inlet Blockage. Revision 1, December 2019.

38. Westinghouse WCAP-16793-NP. Evaluation of Long-Term Cooling Considering Particulate, Fibrous and Chemical Debris in the Recirculating Fluid. October 2011.

Revision 2.

39. NRC Final Safety Evaluation of WCAP-16793-NP. Final Safety Evaluation by the Office of Nuclear Reactor Regulation - Topical Report WCAP-16793-NP, Revision 2 - Evaluation of Long-Term Cooling Considering Particulate, Fibrous and Chemical Debris in the Recirculating Fluid"-PWROG-Project No. 694. April, 2013.

40. WCAP-17788-NP, Volume 3. Comprehensive Analysis and Test Program for GSI-191 Closure (PA-SEE-1090)-Cold Leg Break (CLB) Evaluation Method for GSI-191 Long-Term Cooling. December 2014.

41. NEI Guidance Report on Fibrous Debris Preparation (ML120481057). Generic Procedure - ZOI Fibrous Debris Preparation: Processing, Storage, and Handling.

January 24, 2012. Revision 1.

42. NUREG/CR-6224 (ML083290498). Parametric Study of the Potential for BWR ECCS Strainer Blockage Due to LOCA Generated Debris. October 1995.

43. NUREG-1829 Volume 1. Estimating Loss-of-Coolant Accident (LOCA) Frequencies Through the Elicitation Process. April 2008.

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Enclosure 3 Updated Final Responses to Generic Letter 2004-02

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

45. Letter from John Butler (NEI) to Stewart Bailey (NRC) (ML120730654 and ML120730660). Defense-In-Depth Measures in Support of GSI-191 Resolution Options. March 5, 2012.

46. AISC Manual of Steel Construction. AISC Manual of Steel Construction:

Allowable Stress Design. July 1989. 9th Edition.

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Point Beach Nuclear Operating Corporation Licensing Submittal for a Risk-Informed Resolution of Generic Letter 2004-02 Enclosure 4 Overview of Risk-Informed Approach Table of Contents 1.0 Introduction............................................................................................................ 2 2.0 Overview of Risk Quantification Methodology ....................................................... 3 3.0 Systematic Risk Assessment of Debris ................................................................. 4 4.0 Quantification of Risk Attributable to Debris Effects ............................................ 10 5.0 Risk Quantification Results.................................................................................. 35 6.0 Uncertainty Quantification ................................................................................... 39 7.0 Technical Adequacy of PBN PRA Results........................................................... 47 8.0 Defense-in-Depth and Safety Margin .................................................................. 60 9.0 Monitoring Program ............................................................................................. 61 10.0 Quality Assurance ............................................................................................... 61 11.0 Periodic Update of Risk-Informed Analysis ......................................................... 61 12.0 Reporting and Corrective Actions ........................................................................ 62 13.0 License Application ............................................................................................. 62 14.0 References .......................................................................................................... 62

1. Risk Informed GSI-191 Design Basis...65

2. Available Debris Margins66

3. Application of Debris Margins for Operability Evaluation..67

4. Application of Debris Margins for Future Plant Modifications...69 E4-1

Enclosure 4 Overview of Risk-Informed Approach 1.0 Introduction Generic Safety Issue (GSI)-191 was raised by the United States Nuclear Regulatory Commission (NRC) to ensure that post-accident debris blockage will not impede or prevent the operation of the emergency core cooling system (ECCS) and 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 that would require recirculation (Reference 17). In 2004, the NRC issued Generic Letter (GL) 2004-02, which required all PWR licensees to address the GSI-191 concerns.

In 2010, due to the ongoing challenges of resolving GSI-191, the NRC commissioners issued a staff requirements memorandum (SRM) directing the NRC staff to consider new and innovative resolution approaches (Reference 1). One of the options in the SRM was to address the debris effects 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 respond to GL 2004-02. In 2012, the NRC staff issued SRM-SECY-12-0093 (Reference 2) providing recommendations for closure options.

In a letter to NRC on December 29, 2017 (Reference 22), NextEra Energy Point Beach, LLC (NextEra) provided an updated response to GL 2004-02, using the Alternate Evaluation Methodology defined in NEI 04-07 Section 6 to address the effects of LOCA-generated debris on ECCS and CSS recirculation functions for PBN. Following NRC review and audit in 2019, NextEra subsequently decided to use Option 2, the full risk-informed resolution path, for PBN as communicated to the NRC in a public meeting on May 18, 2021.

This enclosure provides a summary of the risk-informed evaluations performed for PBN Units 1 and 2. The results of this evaluation show with high confidence that the risk associated with strainer and reactor core failures caused by LOCA-generated debris is very low, as defined by Regulatory Guide (RG) 1.174 Region III (Reference 4).

Additionally, the analysis includes significant safety margins and does not affect any of the existing defense-in-depth measures that are in place to protect the public (see of this submittal).

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Enclosure 4 Overview of Risk-Informed Approach 2.0 Overview of Risk Quantification Methodology The risk quantification for PBN is carried out in the following steps.

1. Identify accident scenarios and high likelihood equipment configurations that need to be considered for risk quantification.

2. Evaluate the break scenarios identified in the previous step against all failure criteria to determine the conditional failure probabilities (CFPs) for the strainer and reactor core failures caused by LOCA-generated debris. This evaluation is performed for all the equipment configurations identified in the previous step. A break specific evaluation is performed for strainer failures using the NARWHAL software. The other failure criteria are analyzed using bounding analyses.

3. Calculate changes in core damage frequency ('CDF) and large early release frequency ('LERF) using the CFPs from the previous step. For the breaks on the primary loop, this calculation is performed outside of the PBN probabilistic risk assessment (PRA) model.

For the secondary side breaks inside containment (SSBIs), a bounding analysis is performed using the PBN PRA model by assuming all SSBIs resulting in a consequential LOCA would fail the strainers due to the effects of debris.

4. Compare the baseline CDF and LERF, and CDF and LERF calculated in the previous step to the acceptance guidelines defined in RG 1.174 (Reference 4).

5. Perform an uncertainty analysis for the risk quantification results.

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Enclosure 4 Overview of Risk-Informed Approach 3.0 Systematic Risk Assessment of Debris As described in RG 1.174 (Reference 4), the systematic risk assessment should consider all hazards, initiating events, and plant operating modes. A screening process was used to eliminate scenarios that are not relevant, not affected by debris, or have an insignificant contribution.

3.1 Hazards, and Initiating Events The scenarios that need to be considered for debris effects are breaks that require recirculation through the containment sump strainers. If recirculation is not required, there is no potential for debris-related failures of the strainers, pumps, downstream components, or reactor core.

A systematic process was used to determine the hazards, initiating events, and operating modes to be addressed in the PBN analysis. The process was based on the identification of hazards and initiating events with the potential to (1) generate debris inside containment, (2) require sump recirculation for mitigation of the event, and (3) result in debris transport to the containment sump. Hazards or initiating events that do not meet these three criteria were excluded from the analysis.

The initiating events contained in the PBN Internal Events (IE) PRA model are defined in the PBN Initiating Events Notebook. Among internal plant hazards, the following initiating events do not have the potential to generate debris inside containment and were screened from the analysis:

x Steam generator tube rupture (SGTR) x Interfacing systems LOCAs (ISLOCAs) that discharge outside containment x Anticipated transients including inadvertent safety injection (SI), inadvertent or stuck-open power operated relief valves (PORVs) that discharge to the pressurizer relief tank (PRT), and loss of offsite power x Secondary side breaks outside containment x Initiating events due to loss of component cooling water, loss of service water, and loss of AC or DC power The initiating events contained in the PBN IE PRA model that have the potential to generate debris inside containment are LOCAs (small, medium, and large) and SSBIs.

Although the excessive LOCA (i.e. reactor vessel rupture) initiating event has the potential to generate a significant quantity of debris, this event is assumed to be a catastrophic failure that exceeds the capacity of the ECCS and results directly in core damage.

Therefore, the effects of debris do not need to be addressed for this initiating event.

Internal flood hazards, both inside and outside containment, have been screened from consideration as initiating events. Floods originating within containment are screened from the Internal Flooding PRA on the basis that the containment structure is designed for flooding and such events would be handled as LOCAs. Floods originating outside containment do not have the potential to generate debris inside containment.

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Enclosure 4 Overview of Risk-Informed Approach In addition to the internal initiating events discussed above, internal fires, seismic events, and other external events must also be considered.

Consistent with the guidance in NUREG/CR-6850 (Reference 18), internal fire hazards are not assumed to result in pipe breaks. However, fire induced LOCAs can occur, including spurious opening of a pressurizer PORV, spurious reactor head vent, continuous letdown, spurious ISLOCA, or reactor coolant pump (RCP) seal LOCA due to loss of seal cooling. Of these, only an RCP seal LOCA has the potential to generate debris inside containment. A spurious opening of a pressurizer PORV or spurious reactor head vent is discharged to the PRT, which has negligible sources of debris near the rupture disk. Spurious ISLOCAs or continuous letdown would discharge outside containment.

Therefore, these scenarios are screened out of the analysis. The RCP seal LOCAs are equivalent to a small or medium pipe break and are treated similarly in the risk quantification.

Seismic events can result in direct or indirect LOCAs that generate and transport debris similar to a random pipe break LOCA. A direct seismic-induced LOCA occurs when the reactor coolant system (RCS) pressure boundary fails due to seismic forces. An indirect seismic-induced LOCA occurs when a support or structure fails due to seismic forces, which subsequently causes an RCS pressure boundary failure. As PBN does not have a seismic PRA model, its Individual Plant Examination for External Events (IPEEE) was reviewed to assess the risk impact of seismic-induced LOCAs.

Per the IPEEE, seismic-induced medium and large LOCAs were screened out of the seismic model at a High Confidence of Low Probability of Failure of 0.3g. Further, seismic-induced small LOCAs were demonstrated to not be a significant risk contributor.

Accordingly, seismic-induced large LOCAs with even smaller conditional probabilities of occurrence would also be an insignificant contributor to risk. For the risk quantification associated with debris effects, this finding is further compounded by the fact that many large LOCAs would not be capable of producing enough debris to challenge sump strainer operability. Therefore, seismic-induced LOCAs are judged to have a negligible risk contribution for ECCS strainer performance and are screened from the analysis.

An evaluation of external hazards conducted for PBN concluded that the only other potentially significant external hazards applicable to PBN are:

x Meteor impacts x Aircraft impacts x Turbine missiles x Heavy load drops x High winds None of these external hazards has the potential to generate debris inside containment and are screened from the analysis.

In summary, the following events were addressed qualitatively or quantitatively in the risk quantification for the debris effects:

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Enclosure 4 Overview of Risk-Informed Approach

1. Small, medium, and large LOCAs due to:

x Pipe breaks x Failure of non-piping components x Water hammer

2. SSBIs that result in a consequential LOCA (e.g., due to failure to terminate SI, loss of auxiliary feedwater, or a stuck open PORV) and require sump recirculation 3.2 Initiating Event Frequencies Initiating event frequencies were determined for pipe LOCAs on the primary side of the system, as summarized in Section 3.2.1. LOCAs due to failure of non-piping components were screened out as discussed in Section 3.2.2. Water hammer induced LOCAs were screened out as discussed in Section 3.2.3. For SSBIs, a bounding analysis is performed, as summarized in Section 3.2.4.

3.2.1 Small, Medium, and Large Pipe Break LOCAs For the PBN risk quantification, small, medium, and large pipe break LOCAs were postulated at each in-service inspection (ISI) weld on the RCS within the unisolable portion of the Class 1 pressure boundary. The use of Class 1 ISI welds for break locations is both systematic and thorough because there are multiple welds on every RCS pipe, and the welds cover the range of possible break locations. In addition, a weld is generally closer to equipment that has a large quantity of insulation compared to a break in the middle of a span of straight pipe (e.g., a break on the hot leg weld at the base of the steam generator (SG) will typically generate more debris than a break halfway between the SG and reactor vessel). Also, welds are almost universally recognized as likely failure locations because they can have relatively high residual stress, are preferentially attacked by many degradation mechanisms, and are most likely to have preexisting fabrication defects (Reference 6, p. xviii). The PBN Debris Generation Calculations evaluated partial breaks of 0.5 inches up to the double ended guillotine break (DEGB) for each weld.

Note that the breaks on piping past the first RCS isolation valve are not risk significant 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 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.

The PBN IE PRA model uses the following definitions for small, medium, and large LOCAs:

x Small Break LOCA (SBLOCA): 0.375 to 2-inch breaks x Medium Break LOCA (MBLOCA): 2 to 6-inch breaks x Large Break LOCA (LBLOCA): Greater than 6-inch breaks E4-6

Enclosure 4 Overview of Risk-Informed Approach Table 3-1 shows the exceedance frequencies on a per calendar year (yr-1) basis at the break size bounds for the above break size categories. The exceedance frequencies were calculated from the LOCA frequencies in the PBN IE PRA model. The PBN IE model uses LOCA frequencies from the NRCs regularly updated NUREG-6928 dataset (References 19), which derived the LOCA frequencies from the NUREG-1829 40-year geometric mean data (Reference 6, Table 7.19) by fitting the data to a gamma distribution before Bayesian updating it to reflect the US fleets 797 critical years of operating history (Reference 19). Note that the LOCA frequency of the 31 inch break size in Table 3-1 is the 40-year exceedance frequency based on geometric mean aggregation in NUREG-1829 (Reference 6, Table 7.19). This value would conservatively bound the result quantified using the method in NUREG-6928 as it does not incorporate the Bayesian update as discussed above.

Table 3-1: Exceedance LOCA Frequencies for Risk Quantification Base Case Break Size Exceedance Frequency (yr-1) 0.375 5.20E-04 2 1.46E-04 6 5.52E-06 31 7.50E-08 The exceedance frequencies in the above table are used to determine initiating event frequency for each of the three PRA size categories: SBLOCA, MBLOCA and LBLOCA, as follows. For the size category of LBLOCA, the exceedance frequency for the 31 inch breaks is set to 0, which conservatively maximizes the frequency for LBLOCAs.

x FSBLOCA = F0.375 - F2 = 5.20E 1.46E-04 = 3.74E-04 yr-1 x FMBLOCA = F2 - F6 = 1.46E 5.52E-06 = 1.40E-04 yr-1 x FLBLOCA = F6 - F31 = 5.52E 0 = 5.52E-06 yr-1 1F 3.2.2 Non-Pipe LOCAs Non-pipe LOCAs are not explicitly evaluated. Non-pipe components whose failure could result in a LOCA include nozzles, component bodies, pressurizer heater sleeves, manways, control rod drive mechanism penetrations, safety relief valves, RCP seals, reactor vessel, pressurizer vessel, steam generator vessels, welded caps on retired lines, and other components. It is judged 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 located inside the reactor cavity, all of the 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 RCPs, that could be used to estimate debris generated from non-weld locations in those areas. The modeled welds E4-7

Enclosure 4 Overview of Risk-Informed Approach 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 the reactor vessel (e.g., control rod drive penetrations, manways, and instrument lines connected to the reactor vessel) 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 failure of these components is expected to be bounded by the reactor vessel nozzle breaks. Note that the analysis of the reactor nozzle breaks did not credit shadowing by the reactor and the zone of influence (ZOI) for these breaks include a line-of-sight cone projecting out of the closest primary shield penetration to the radius of the ZOI sphere.

3.2.3 Water Hammer Induced LOCAs The approach used to demonstrate that the risk of water hammer induced LOCA is acceptably low is to verify that water hammer is not likely to cause a pipe rupture in the break locations that can produce and transport problematic quantities of debris.

The portions of the RCS that are subject to a LOCA are designed to the requirements of the USAS B31.1 Code for Pressure Piping, which includes consideration of appropriate transients. The RCS pressure boundary 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 was given to loadings under normal operating conditions, as well as abnormal loadings, such as pipe rupture and seismic loadings. A search of PBNs corrective action program was performed, and no issues were found related to water hammer induced LOCAs.

In addition to the robust design of the RCS, the PBN safety systems, including the safety injection (SI), residual heat removal (RHR) and containment spray (CS) systems, are protected from the potential for gas accumulation, which was addressed under the responses to GL 2008-01 (References 8, 9 and 10). PBN reviewed the design drawings of these systems to identify potential locations for gas accumulation and performed laser scanning and manual slope measurements. Ultrasonic testing was conducted at the high risk locations. For the locations where gas voids were detected, the size of the voids were analyzed to demonstrate that they would not result in excessive pressure transient at system actuation.

PBN also implemented modifications and procedure updates as necessary. These changes included adding vent valves at high points of the SI, RHR and CS systems, revising procedures to incorporate the use of new vent valves for filling and venting operations and during monthly surveillances. Additionally, PBN implemented a long-term gas accumulation management program, which included supporting operating, administrative and surveillance procedures, as well as preventive maintenance callups to perform monthly ultrasonic testing examinations of high points in the inaccessible piping for both units. PBNs responses to GL 2008-01 were accepted by the NRC and deemed E4-8

Enclosure 4 Overview of Risk-Informed Approach effective in precluding gas accumulation in the SI, RHR and CS system and in preventing a water hammer event that could challenge system integrity or operation (Reference 11).

Based on the fact that the RCS piping is designed to B31.1 standards, the implementation of an approved gas accumulation management program, and the lack of historical data for water hammer induced LOCA events, the relevance of water hammer events in the analysis of LOCA-generated debris is deemed insignificant. It is therefore reasonable to conclude that LOCA frequencies are not impacted by water hammer considerations in these systems.

3.2.4 Secondary Side Breaks Similar to a LOCA, insulation and coatings debris could be generated by SSBIs (e.g., a large break in a main steam or feedwater line). SSBIs, which result in a consequential LOCA upon failure to terminate safety injection or a stuck open PORV, may require sump recirculation. Under this scenario, the debris generated by the SSBIs could be transported to the ECCS strainers.

The PBN IE PRA model was used to perform a bounding evaluation of the risk contribution from the SSBIs. The evaluation assumed that all SSBIs resulting in a consequential LOCA would fail the strainers due to the effects of debris. This is a conservative assumption because SSBIs would generate less debris than equivalent primary side breaks due to the lower pressure on the secondary side.

In order to determine the bounding risk due to strainer failures for the SSBI scenarios, the existing sump strainer failure events relevant to SSBIs in the PBN IE PRA model were set to TRUE and the CDF and LERF contributions for the scenarios were quantified. CDF and LERF were also quantified assuming the sump functions perfectly (i.e., by setting the strainer failure events to FALSE) to establish the baseline values for CDF and LERF. The differences in CDF and LERF between these two runs were taken to be the bounding SSBI risk impact (CDF and LERF) due to debris effects. The results of the analysis are shown in Section 5.3 of this enclosure.

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Enclosure 4 Overview of Risk-Informed Approach 4.0 Quantification of Risk Attributable to Debris Effects Based on the qualitative screening and assessment of relevant hazards and initiating events described above, the following events are included in the quantitative risk assessment:

x Primary loop pipe breaks resulting in small, medium, and large break LOCAs.

These include breaks ranging from 1/2 partial breaks to DEGBs on every Class 1 ISI weld within the first isolation valve.

x SSBIs that result in a consequential LOCA and require sump recirculation.

As stated in Section 3.2.4, a bounding analysis is performed to quantify the risk contribution from the SSBIs. The results of this analysis are shown in Section 5.3.

The remainder of this section summarizes the quantification of risk increase due to strainer and reactor core failures caused by debris generated by LOCAs on the primary loop piping. Note that the evaluation was performed for LOCAs 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 reasonable because the RCS pressure and temperature, which are the key inputs affecting the sizes of break ZOIs, would either be approximately the same or significantly lower for Modes 2 through 6. Also, the flow rate required to cool the core would be significantly reduced for low power or shutdown modes.

4.1 Failure Mode Identification Various failure criteria are considered in the risk quantification for the primary loop breaks, as listed below.

1. Debris accumulation in an upstream flow path choke point exceeds blockage limits and reduces the available sump volume.

2. Strainer head loss exceeds the net positive suction head (NPSH) margin for the pumps required during post-LOCA recirculation operation.

3. Strainer head loss exceeds the strainer structural margin.

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

5. Gas void fraction at pump suction due to degasification exceeds the acceptance limit.

6. Flashing of sump fluid downstream of the sump strainer.

7. Air entrainment due to vortexing.

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

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

10. Boric acid concentration in the reactor core exceeds the solubility limit resulting in boric acid precipitation (BAP).

Those failure modes related to strainer head loss due to accumulation of debris on the strainers are analyzed for all postulated breaks in a time dependent manner in E4-10

Enclosure 4 Overview of Risk-Informed Approach NARWHAL, as listed below. Note that the last failure mode listed is used to address the scenario where the debris loads for some of the breaks are not bounded by the strainer head loss testing. Refer to Section 4.7.6 for details.

1. Strainer head loss exceeds the NPSH margin for the RHR pumps.

2. Strainer head loss exceeds the strainer structural margin.

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

4. Gas void fraction at the RHR pump suction exceeds the acceptance limit.

5. Flashing occurs downstream of the strainer.

6. Debris accumulation on the strainer exceeds one or more tested debris limits.

Other failure modes, as shown below, are addressed in a bounding manner outside of NARWHAL. Refer to sections in Enclosure 3 for details.

1. Upstream blockage, see the Response to 3.I in Enclosure 3.

2. Air entrainment due to vortexing, see the Response to 3.f.3 in Enclosure 3.

3. Ex-vessel downstream effects, see the Response to 3.m in Enclosure 3.

4. In-vessel downstream effects and BAP, see the Response to 3.n in Enclosure 3.

Figure 4-1 shows the relationship between various testing and analyses for the PBN risk quantification and uncertainty analysis. Note that the overall evaluation and testing for PBN were based on models that have been used in the past and accepted by the NRC for responding to GL 2004-02.

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Enclosure 4 Overview of Risk-Informed Approach Figure 4-1: Calculations and Analyses Used in Risk and Uncertainty Quantification E4-12

Enclosure 4 Overview of Risk-Informed Approach 4.2 Sub-Model Development The failure modes related to strainer head loss are analyzed using the NARWHAL software. The PBN NARWHAL model includes the following sub-models:

x Post-accident conditions o Plant configuration o Plant states o Random equipment failures o Water volume and level o Pump flow rates o Containment pressure, and containment and sump temperature o Containment spray and sump pH x Debris sources o ZOI-generated insulation, and qualified coatings debris o Unqualified coatings debris o Actively delaminating qualified (ADQ) epoxy coatings o Latent debris o Miscellaneous debris x Chemical effects o Chemical release into sump pool o Aluminum solubility o Precipitate debris quantity x Debris transport o Blowdown o Washdown o Pool fill-up o Recirculation o Erosion x Strainer head loss o Debris groups o Clean strainer head loss (CSHL) o Conventional debris head loss o Chemical debris head loss o Head loss correction for temperature and flow rate x Strainer air intrusion x 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 These sub-models are described in more detail below.

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Enclosure 4 Overview of Risk-Informed Approach 4.3 Scenario Development 4.3.1 Plant Configuration The plant configuration determines the flow paths during different phases of accident mitigation. At PBN, each unit has two separate sump recirculation strainer assemblies that individually and specifically feed either the A or B train RHR pump. Each RHR pump supplies suction to an associated SI or CS pump. In addition to these pumps, each train of ECCS contains a safety injection accumulator (SIA).

The figure below shows the equipment and piping connection diagram created in NARWHAL to model both PBN units. The diagram shows only two loops are used in the model, consistent with the PBN design.

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Enclosure 4 Overview of Risk-Informed Approach Figure 4-2: PBN System Diagram Created in NARWHAL E4-15

Enclosure 4 Overview of Risk-Informed Approach 4.3.2 Plant States This section describes operations of the PBN ECCS and CS systems during the injection and recirculation phases modeled in NARWHAL. The equipment lineups during different modes of operation are in accordance with the PBN emergency operating procedures (EOPs) and are applied to every break.

In the injection mode, two SI pumps, two RHR pumps, and two CS pumps take suction from the refueling water storage tank (RWST). The SI pumps deliver to the RCS through the cold leg connections, and the RHR pumps inject into the upper plenum of the reactor vessel.

Recirculation mode starts when the RWST level reaches the low level setpoint. The RHR pumps are switched over to recirculation to continue providing upper plenum injection.

The SI pumps are secured before switchover to recirculation.

The switchover of CS from injection mode to recirculation mode occurs when the RWST water level reaches the low-low level setpoint. Upon receiving the RWST low-low level alarm, the operator initiates manual actions to realign one of the CS pumps from RWST to the RHR pump discharge. The CS pump is operated for two hours in the recirculation mode before being secured.

After the CS pump is secured, one of the SI pumps is restarted to take suction from the RHR pump discharge and supplies flow to the cold leg, through the downcomer, to the reactor core inlet. This flow helps mitigate the BAP concerns. The RHR pumps continue to inject into the upper plenum of the reactor vessel.

A plant state in NARWHAL represents a unique plant mode of operation and is defined by the activity state of the valves and pumps in the connection diagram. In the PBN model, 4 plant states are defined and are applied for all breaks analyzed, as shown in the table below.

Table 4-1: Definition of Plant States Plant State Name Initiating Time or Variable Injection Automatically start at time 0 Start of RHR Recirculation RWST low level alarm Start of CS Recirculation RWST low-low level alarm After Termination of CS Recirculation 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after start of CS recirculation E4-16

Enclosure 4 Overview of Risk-Informed Approach Table 4-2 shows the pump operating status in each of the plant states defined.

Table 4-2: Plant State Component Activity State Name Component Activity Description RHR Pumps Active Suction from RWST SI Pumps Active Suction from RWST Injection CS Pumps Active Suction from RWST SIAs Active Injection into CL Start of RHR RHR Pumps Active Suction through sump strainer Recirculation SI Pumps Inactive Not operating (RWST Low CS Pumps Active Suction from RWST Level) SIAs Inactive Not operating Start of CS RHR Pumps Active Suction through sump strainer Recirculation SI Pumps Inactive Not Operating (RWST Low- CS Pumps Active Suction from RHR pump discharge Low Level) SIAs Inactive Not Operating After RHR Pumps Active Suction through sump strainer Termination of SI Pumps Active Suction from RHR pump discharge CS CS Pumps Inactive Not Operating Recirculation SIAs Inactive Not Operating 4.3.3 Random Equipment Failures Random equipment failures are defined as failure to start or failure to run due to issues unrelated to the debris effects. The PBN PRA model was used to identify the likelihood of different equipment configurations that could occur in response to a LOCA. All potential equipment failure combinations of the RHR, SI and CS pumps were screened to identify unique equipment failure scenarios that need to be modeled in detail in NARWHAL.

The table below shows the equipment failure combinations considered in the screening.

Each number in the table indicates the number of failed pumps. Note that the PRA success criteria require at least one RHR pump and one SI pump being operable. As a result, the scenarios with failures of both RHR pumps and/or both SI pumps are not shown in the table.

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Enclosure 4 Overview of Risk-Informed Approach Table 4-3: Screening of Equipment Failure Scenarios Case ID RHR SI CS Note A 0 0 0 No pump failure - Unique B 0 0 1 Same as Case A since only 1 SI pump and 1 CS C 0 1 0 pump are required during recirculation.

D 0 1 1 I 0 0 2 Failure of both CS pumps Same as Case I since only 1 SI pump is required J 0 1 2 during recirculation E 1 0 0 Same as Case H since only 1 SI pump and 1 CS F 1 0 1 pump are required during recirculation G 1 1 0 H 1 1 1 Single Train Failure - Unique K 1 0 2 Failure of both CS pumps and 1 RHR pump Same as Case K since only 1 SI pump is required L 1 1 2 during recirculation For Case I, both CS pumps are assumed to fail at the start of recirculation. The CFPs for this case are expected to be bounded by (i.e., lower than) Case A. Without CS operating in recirculation mode, less chemical precipitate will form, and therefore failures due to chemical effects will decrease. Case I is also similar to Case A with both RHR pumps and strainers in operation. As a result, the equipment lineup of Case I was not modeled in NARWHAL. Instead, the CFPs of Case A were applied to Case I.

Similarly, Case K is judged to be bounded by Case H, and the CFPs of Case H were applied for Case K.

In summary, the following two equipment failure scenarios were modeled in detail in NARWHAL to evaluate strainer failures:

x No Equipment Failures (Case A) x Single Train Failure (Case H)

These two scenarios are bold faced in Table 4-4 below. The resulting CFPs of each modeled scenario were applied to two other scenarios. The functional failure probabilities of the three scenarios in each group (separated by the double line in the table) can therefore be combined, as shown in the last column of the table.

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Enclosure 4 Overview of Risk-Informed Approach Table 4-4: Combined Equipment Failure Scenarios Combined Functional Failure Equipment Configuration Functional Failure Probabilities Probabilities All pumps available 9.78E-01 1 CS pump failure 7.86E-03 9.86E-01 2 CS trains failure 5.43E-05 1 RHR pump failure 1.35E-02 1 RHR pump + 1 CS pump failure 2.29E-04 1.37E-02 1 RHR pump + 2 CS pump failures 1.22E-06 In the PBN NARWHAL models, the random pump failures were assumed to occur at the start of recirculation. This is conservative because it results in an earlier switchover to recirculation when compared to failure at the beginning of the event. Additionally, allowing the CS pumps to operate during the injection phase ensures that the debris blown into the upper containment can be washed down into the sump pool, resulting in more transportable debris in the pool.

4.3.4 Sump Pool Volume and Level The quantity of water in the recirculation pool is based on the total water in containment minus any transitory and geometric hold-up volumes. These values have been defined for the minimum and maximum pool volumes. In the NARWHAL base case model, only the values required for calculating the minimum pool volumes were used.

In NARWHAL, sump pool volumes and levels are calculated as a function of time based on the source water volumes and hold-ups. The initial water sources available for the sump pool include the RWST, RCS, spray additive tank (SAT), and SIAs. See the Response to 3.g.12 in Enclosure 3 for a summary of the water sources considered.

x The RWST inventory is delivered to the sump as a function of the pump flow rates during injection phase. The RWST water quantities at two intermediate levels are used in the NARWHAL model to determine timing for switchover to sump recirculation for the RHR and CS pumps. The start of RHR recirculation is initiated by the low level setpoint of the RWST while the start of CS recirculation by the low-low level setpoint of the RWST.

x The RCS contribution to or holdup from the sump pool depends on break size and elevation.

x The sodium hydroxide (NaOH) solution in the SAT is delivered to the sump pool through an eductor into CS.

x The SIAs delivers their inventory into RCS cold leg during the injection phase, which could spill into the pool.

The total quantity of water hold-up from the sump pool depends on the break size, CS activation, and break elevation. See the Response to 3.g.8 for a summary of the hold-up E4-19

Enclosure 4 Overview of Risk-Informed Approach volumes accounted for. Hold-up volumes are filled by the appropriate flow rates over time.

For example, the hold-up in the refueling cavity is filled by CS until it is full, after which any water entering this compartment spills to the sump.

For each time step, the pool volume is calculated first, and pool water level is then determined using a piecewise function that relates pool water level to volume.

4.3.5 Flow Rates Table 4-5 shows the RHR pump flow rates used for all breaks during the four plant states defined in the NARWHAL base model. For the injection phase, the design flow rates of the SI, RHR and CS pumps were used. For the recirculation phase, since the NARWHAL model only evaluates strainer failures, higher RHR pump flow rates were used to maximize the strainer flow rate. The CS and SI pumps piggyback off a RHR pump during recirculation. Therefore, the CS and SI pump flow rates do not impact strainer failures analyzed in NARWHAL. See the Responses to 3.g.1 and 3.g.2 for details.

The pump flow rates used in the in-vessel downstream effects analysis are discussed in the Response to 3.n of Enclosure 3.

Table 4-5: RHR Pump Flow Rates for PBN States RHR Pump (gpm)

Injection 1560 Start of RHR Recirculation 2100 Start of CS Recirculation 2100 After Termination of CS Recirculation 2100 4.3.6 Pressure and Temperature The sump temperature profile, containment temperature profile, and containment pressure profile are used in NARWHAL to determine time-dependent thermal hydraulic properties. The maximum post-LOCA containment temperature and sump pool temperature profiles are used in NARWHAL.

The containment pressure was assumed to be the saturation pressure at a pool temperature above 212°F, and 14.7 psia at pool temperatures at or below 212°F.

Containment accident pressure was not credited for pump NPSH or strainer degasification evaluations. For flashing evaluation, 2 psi of containment accident pressure was credited for the first 200 minutes after initiation of the accident to mitigate flashing failure. See the justification for this approach in the Response to 3.f.14 in .

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Enclosure 4 Overview of Risk-Informed Approach 4.3.7 Sump and Spray pH The sump and spray pH values were used for the sump pool chemical effects evaluation.

A conservative combination of pH values was used, as detailed in the Response to 3.o.2.3 of Enclosure 3.

4.4 Debris Sources The NARWHAL model considered all types of debris generated inside the break ZOI, including low density fiberglass (LDFG), Nukon (for PBN2 only), Temp-Mat, Mineral Wool, Cal-Sil, Asbestos Cal-Sil, and qualified epoxy and inorganic zinc (IOZ) coatings. For these debris types, the debris loads vary with break size and location, and are obtained from the BADGER software runs in the PBN debris generation calculations. The Responses to 3.a, 3.b and 3.c in Enclosure 3 summarize the debris generation evaluation. The Response to 3.b.4 in Enclosure 3 shows the generated debris loads of bounding breaks for these debris types.

The NARWHAL model also accounted for unqualified epoxy, IOZ, and alkyd coatings, actively delaminating qualified (ADQ) epoxy coatings, latent debris, aluminum coatings and miscellaneous debris. These materials may fail in the post-accident environment. The debris loads for these debris types, except for aluminum coatings, were assumed to be independent of break size or location. The debris loads for these debris types are shown in the Responses to 3.b.5, 3.d.3, and 3.h.1 in Enclosure 3. For aluminum coatings, see discussion in Section 4.5 of this enclosure.

Debris from all sources was treated as being generated at the beginning of the event. The strainer surface area reduction due to blockage by the miscellaneous debris was applied prior to other debris transporting to the strainer. Unqualified coatings were assumed not to transport to inactive cavities during pool fill-up and were 100% available to transport at the start of recirculation.

4.5 Chemical Effects Sump pool chemical effects refer to formation of chemical precipitate in the post-LOCA pool and its impact on strainer performance. The evaluation of chemical effects includes the following:

x Quantifying release of aluminum into the sump pool. For PBN, the methodology is based on WCAP-16530-NP-A (Reference 20) and its associated NRC safety evaluation.

x Determining if aluminum precipitates would form. Precipitation occurs when the concentration of released aluminum in the pool exceeds the aluminum solubility limit at the corresponding sump temperature. The Argon National Lab (ANL) equation was used to calculate aluminum solubility.

x Determining the types and quantities of precipitate that forms from the released aluminum.

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Enclosure 4 Overview of Risk-Informed Approach Chemical debris was first quantified in a hand calculation using bounding inputs, and the results of the hand calculation informed the PBN strainer head loss testing.

In NARWHAL, the evaluation outlined above is performed for each postulated break in a time-dependent manner using break-specific debris loads. In the model, precipitation is forced to occur at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (or 1,440 minutes) following initiation of the accident even if the solubility limit is not exceeded. The calculated quantity of precipitates is treated as a separate type of debris and is transported to the strainers as a function of pool turnover.

The maximum release of aluminum from debris (e.g., E-Glass, mineral wool) depends on material composition. The maximum amount of aluminum that can leach from each of these debris types was taken from the WCAP-16530-NP-A methodology. Surfaces inside containment that may release aluminum include aluminum metal, aluminum coatings and concrete, which are either submerged in the containment pool or exposed to containment sprays. All breaks of each unit are analyzed with the bounding surface areas of aluminum metal and concrete found within the two units.

Aluminum coatings on the pressurizer and reactor vessel may become exposed following a LOCA in the pressurizer compartment and reactor cavity. Quantities of aluminum in the exposed coatings were determined and input into NARWHAL. When evaluating aluminum release into the sump pool, the aluminum in the exposed coatings was treated the same as aluminum metal. To do this, the total mass of aluminum in the exposed coatings was specified in NARWHAL. The maximum amount of aluminum coatings on the pressurizer of either of the units contains 4.5 lbm of aluminum.. This quantity is conservatively assigned to all breaks inside the pressurizer compartment for both units.

The maximum amount of aluminum coatings on the reactor vessel (excluding the reactor head) contains 15.6 lbm of aluminum for both PBN units. It was assumed that all reactor cavity breaks greater than 20 inches can destroy all of the insulation on the reactor vessel (except for the reactor head) and expose the aluminum coatings. As a result, the full amount of 15.6 lbm of aluminum was assigned for all reactor cavity breaks greater than 20 inches. For breaks smaller than or equal to 20 inches, the amount of aluminum from coatings was scaled down from the maximum amount based on the generated RMI debris load. This approach is conservative because the largest break inside the reactor cavity generates more than double the amount of RMI debris of the worst 23 and 26 inch breaks.

It is therefore conservative to assume that all reactor cavity breaks greater than 20 inches can expose the maximum amount of aluminum coatings on the reactor vessel (except for the reactor head).

4.6 Debris Transport As described in the Response to 3.e in Enclosure 3, debris transport analysis considered the blowdown, washdown, pool fill-up, and recirculation phases, as well as debris erosion.

The blowdown transport fractions are a function of break location and size of debris. The only debris transported during the blowdown phase would be debris generated inside a break ZOI. Fine debris was transported with the blowdown flow, with no credit for retention E4-22

Enclosure 4 Overview of Risk-Informed Approach on structures. The transport of small and large pieces of fiberglass and small pieces of calcium silicate 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. In the PBN NARWHAL models, containment sprays were assumed to be operating for all breaks such that the fine debris blown into the upper containment is washed down to the sump pool. Some credit was taken for small and large pieces of fiberglass being retained in upper containment.

Pool fill-up transport refers to a relatively small fraction of debris being transported to inactive cavities as the sump pool is filled. These transport fractions were applied to all debris that was in the sump 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 was analyzed using computational fluid dynamics (CFD) modeling of flow inside the sump pool. Several simulations were run to determine the recirculation transport fractions for debris of various types and sizes, different break locations, and different operating ECCS train. The recirculation transport fraction is 100% for all types of fine debris and particulate debris, and all analyzed break cases.

Erosion fractions were applied to the small and large pieces of fiberglass debris, and small pieces of Cal-Sil debris to account for formation of fine debris due to erosion. Specific erosion fractions used are shown in the Response to 3.e.6 of Enclosure 3.

The debris transport fractions of each phase described above (see values shown in the Response to 3.e.6 of Enclosure 3) were input into NARWHAL, which uses the logic tree approach to determine the overall transport fraction for each debris type and size category. Although debris transport could vary with break size, strainer flow rate, pool water level, actuation of CS, etc, the PBN NARWHAL model used debris transport fractions from the bounding conditions (e.g., the LBLOCA strainer flow rates were used to calculate recirculation transport fractions that were applied to breaks of all sizes).

Conservative timing was assumed for the transport phases discussed above.

x Blowdown was treated as an instantaneous process at the beginning of the event.

x Washdown was treated as occurring after the inactive cavities were filled during the pool fill-up phase, but before the start of recirculation.

x Although erosion is a time-dependent process, all of the erosion fines were treated as being generated at the start of recirculation.

x The recirculation transport of debris to the strainers was modeled as a time-dependent process, where debris arrives at the strainers as a function of the pool turnover time (i.e., as a function of the pool volume and strainer flow rates). The concentration of each transportable debris type in the recirculation pool was assumed to be homogeneous. When both strainers are in operation, the debris E4-23

Enclosure 4 Overview of Risk-Informed Approach accumulation on each strainer was assumed to be proportional to the flow split between the two strainers during each time step.

For PBN, failure time was not credited for unqualified coatings. The unqualified coatings were assumed to be in the lower containment and 100% available for recirculation transport.

Though transport fraction of large and small pieces of debris were input into the NARWHAL model for blowdown, washdown and pool fill up, their recirculation transport fraction in the sump pool were not considered. This is acceptable because it was observed during head loss testing that small and large pieces of debris transport by tumbling along the floor. Since the PBN strainers are a few inches off the floor, the transported small and large pieces pile up in front of the strainer and therefore have little direct contribution to strainer head loss.

In the NARWHAL model, debris arriving at the strainer during each time step was assumed to be 100% captured by the strainer, which conservatively maximizes debris load on the strainer. Note that in-vessel effects and fiber bypass were analyzed outside of NARWHAL.

4.7 Strainer Failure Evaluation The strainer failure modes discussed below involve strainer head loss due to accumulation of debris on the strainer. This section describes strainer head loss determination and how it is used to assess different failure criteria.

4.7.1 Strainer Head Loss In the NARWHAL model, total strainer head loss was calculated during each time step by combining the CSHL and appropriate debris head loss. The CSHL is discussed in the Response to 3.f.9 of Enclosure 3. In NARWHAL, the head loss contributions of conventional debris and chemical debris are applied separately, using lookup tables derived from PBN-specific strainer head loss testing data. Refer to the Responses to 3.f.4 and 3.f.10 of Enclosure 3.

4.7.2 Degasification and Flashing Degasification and flashing were evaluated during each time step using the total strainer head loss, containment pressure, sump temperature, and water level. Refer to the Response to 3.f.14 in Enclosure 3 for details.

4.7.3 NPSH Margin In NARWHAL, pump NPSH margin is defined to be the difference between NPSH available and NPSH required where the calculated NPSH available does not account for strainer head loss. Therefore, an NPSH failure is recorded for the pump if the total strainer E4-24

Enclosure 4 Overview of Risk-Informed Approach head loss exceeds the calculated pump NPSH margin. Since only the RHR pumps take direct suction through the sump strainer, pump NPSH analysis and failure involve only the RHR pumps.

Pump NPSH available is calculated for each time step based on the sump pool water level, containment pressure and sump temperature. No containment accident pressure was credited for NPSH evaluation. For each time step, the NPSH required from the vendor curve was increased to account for the impact on pump performance due to void fraction at pump suction. This adjustment used the methodology recommended in RG 1.82 (Reference 23).

4.7.4 Strainer Structural Margin For each time step, the strainer head loss was compared against the strainer structural margin. If the structural margin is exceeded, a strainer failure was recorded.

4.7.5 Strainer Partial Submergence If the strainer is partially submerged, the strainer is assumed to fail if the total strainer head loss exceeds half of the submerged strainer height. No failures due to strainer partial submergence occurred for PBN because the strainer is fully submerged during recirculation for all break sizes (see the Response to 3.g.1 in Enclosure 3).

4.7.6 Debris Limits If a break has one or more debris loads exceeding the maximum tested quantities, the break was assumed to cause a debris limit failure. Debris limits were defined for different debris types or debris groups based on the maximum debris loads used during head loss testing. See the Response to 3.f.5 in Enclosure 3.

4.8 Sub-Model Integration in NARWHAL 4.8.1 NARWHAL Software For PBN, the NARWHAL software is used to holistically analyze strainer failures caused by LOCA-generated debris in a time dependent manner. 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 debris effects 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 of strainer failures caused by LOCA-generated debris.

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Enclosure 4 Overview of Risk-Informed Approach 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 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.

A single NARWHAL simulation relies on a series of marching algorithms (Figure 4-3).

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.

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Enclosure 4 Overview of Risk-Informed Approach Figure 4 NARWHALs 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 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, and maximum head loss for each strainer. 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.

In addition to calculating which breaks pass and fail the acceptance criteria, NARWHAL can also be used to post-process the results to calculate the CFP values.

The following flow diagrams show an overview of the analytical models that determine how the water is transported, how the conventional debris is generated and transported, how chemical precipitates form and transport, and how the strainer failure criteria are analyzed. NARWHAL holistically analyzes each of these models in a time-dependent manner.

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Enclosure 4 Overview of Risk-Informed Approach In the PBN NARWHAL model, each break evaluated in NARWHAL was run for a duration of 30 days (43,200 minutes). The first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (1,440 minutes) were evaluated with a time step size of 1 minute. After the first day, the time step is increased to 60 minutes.

This is a reasonable application since the majority of the transient occurs within the first few hours. Long term effects can be effectively analyzed with a coarser time-step.

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Enclosure 4 Overview of Risk-Informed Approach Figure 4-4: Water Transport and Accumulation E4-29

Enclosure 4 Overview of Risk-Informed Approach Figure 4-5: Debris Generation and Transport Note: Boxes No. 8 and 9 in this flow diagram are for fiber penetration and in-vessel effects, which are analyzed outside of NARWHAL.

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Enclosure 4 Overview of Risk-Informed Approach Figure 4-6: Chemical Product Formation and Transport Note: Boxes No. 6 and 7 in this flow diagram are for chemical debris penetration through the strainer and deposition inside the core. This is part of the analysis for in-vessel effects, which are analyzed outside of NARWHAL.

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Enclosure 4 Overview of Risk-Informed Approach Figure 4-7: Comparison of Head Loss to Strainer Failure Criteria E4-32

Enclosure 4 Overview of Risk-Informed Approach 4.8.2 Calculation of Conditional Failure Probability The CFP was quantified for each PRA size category and each equipment lineup using the following steps:

1. Each postulated break was analyzed against the strainer failure criteria, as described in Section 4.1. For PBN, the CFPs were computed for the three PRA size categories:

SBLOCA, MBLOCA and LBLOCA. Note that in-vessel effects were analyzed outside of NARWHAL.

2. Overall LOCA frequencies were allocated to individual welds and break sizes using the top-down methodology.

3. The size category of LBLOCAs (6 to 31-inch breaks) was further discretized into three size ranges. The break size at the upper end of each size range was included in the next size range (e.g., 15-inch breaks are included in Size Range 2, not Size Range 1). All breaks within each size range were assumed to have the same probability to occur.

Size Range 1 6 inches to 15 inches Size Range 2 15 inches to 25 inches Size Range 3 Greater than 25 inches The use of multiple size ranges was necessary in order to more accurately distribute frequency because the LOCA frequency varies significantly with break size. For example, a 6- inch break is roughly two orders of magnitude more likely to occur than a 31-inch break.

4. For a given equipment configuration, the CFPs were reported for each PRA size category and success criterion. To estimate the CDF, the CFPs for all success criteria within each PRA size category were summed, resulting in a single CFP value for each PRA size category.

4.8.3 Calculation of 'CDF and LERF Various equipment configurations were screened, and the unique ones were modeled in detail in NARWHAL (see Section 4.3.3). From the NARWHAL results, the CFPs were calculated for each equipment configuration. These CFP values, along with the event initiating frequencies and functional failure probabilities, were substituted into the following equation to estimate CDF.

 Equation 1 Where:

i= Each PRA size category j= Each equipment configuration IEF = Initiating event frequency for each PRA size category E4-33

Enclosure 4 Overview of Risk-Informed Approach CFP = CFP of each PRA size category and each equipment configuration FFP = Functional failure probability for each equipment configuration The value of LERF was calculated by multiplying the CDF by the conditional large early release probability (CLERP). The CLERP for large LOCA accident sequences was calculated to be 1.92E-03 for both PBN units.

4.9 In-Vessel Downstream Effect As discussed in the Response to 3.n of Enclosure 3, in-vessel downstream effects were addressed in a bounding manner and therefore were not included in the PBN NARWHAL model. The evaluation showed that in-vessel downstream effects would not result in any reactor core failures even under the worst-case conditions and therefore would not contribute to the overall risk.

4.10 Ex-Vessel Downstream Effects As discussed in the Response to 3.m of Enclosure 3, ex-vessel downstream effects were analyzed outside of NARWHAL model. The evaluation showed that ex-vessel downstream effects would not cause any failures of downstream components or systems and therefore would not contribute to the overall risk.

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Enclosure 4 Overview of Risk-Informed Approach 5.0 Risk Quantification Results This section summarizes the results and conclusions of the base cases for the risk quantification.

5.1 PBN NARWHAL CFP Calculation NARWHAL runs were performed to determine the CFPs for the PBN Units 1 and 2 base cases for the LOCAs on the primary side of the system, as summarized in Table 5-1 and Table 5-2. NARWHAL outputs showed no failures of SBLOCAs or MBLOCAs for any equipment lineup. For both units, the smallest break that resulted in a failure is a 11.188-inch DEGB due to exceeding the mineral wool debris limit.

For the single train failure case, most of the failures are due to debris loads exceeding one or more debris limits. Some of the breaks larger than or equal to 20 inches fail the coatings particulate debris limit. Some breaks fail the mineral wool debris limit since the head loss testing was performed before the increase in the size of break ZOI for mineral wool. Some of the reactor cavity breaks fail the SAS chemical debris limit due to the conservative treatment of the aluminum coatings on the reactor vessel. Lastly, some breaks do not exceed any of the debris limits but their combinations of debris loads were not bounded by any of the head loss tests. These breaks were assumed to fail due to lack of head loss testing data.

For the scenario with no pump failures, fewer breaks fail the acceptance criteria compared with the single train case because debris is split between two strainers. Most of the failures result from exceeding the mineral wool debris limit.

Table 5-1: PBN Unit 1 Base Case CFPs Equipment Lineup SBLOCA MBLOCA LBLOCA All pumps available 1 CS pump failure 0 0 2.983E-03 2 CS trains failure Single train failure 1 RHR pump failure 0 0 7.622E-02 1 RHR pump + 2 CS pump failures E4-35

Enclosure 4 Overview of Risk-Informed Approach Table 5-2: PBN Unit 2 Base Case CFPs Equipment Lineup SBLOCA MBLOCA LBLOCA All pumps available 1 CS pump failure 0 0 5.409E-03 2 CS trains failure Single train failure 1 RHR pump failure 0 0 9.621E-02 1 RHR pump + 2 CS pump failures 5.2 Baseline CDF and LERF The PBN PRA models include initial events, internal flooding, and internal fire. A separate other external hazards analysis was also included in the baseline. The baseline CDF and LERF values were taken from the quantification notebooks for each of the PRA models.

Since PBN does not have a completed seismic model, the baseline seismic CDF and LERF are addressed with a bounding estimate. Finally, an additional CDF and LERF contribution were accounted for due to legacy design code nonconformances associated with both units containment dome construction trusses, as required by NRC commitment.

These baseline CDF and LERF values are summarized in Table 5-3.

Table 5-3: Baseline CDF and LERF Values Unit 1 Unit 2 PRA Model -1 CDF (yr ) LERF (yr-1) -1 CDF (yr ) LERF (yr-1)

Internal Events 2.36E-06 9.81E-08 2.30E-06 9.63E-08 Internal Flooding 4.33E-07 3.33E-08 4.68E-07 3.21E-08 Internal Fire 5.77E-05 9.01E-07 6.84E-05 1.04E-06 Seismic 6.24E-06 2.62E-07 6.24E-06 2.62E-07 Other External Hazards <1.00E-06 <1.00E-07 <1.00E-06 <1.00E-07 Construction Truss 1.23E-06 5.92E-07 1.24E-06 5.95E-07 Total 6.90E-05 1.99E-06 7.96E-05 2.13E-06 5.3 'CDF and 'LERF Using the initiating event frequencies in Table 3-1, equipment functional failure probabilities in Table 4-4, and the CFP values shown in Table 5-1 and Table 5-2, the risk increase due to failures caused by LOCA-generated debris on the primary side of the system was calculated. The results are summarized in the table below.

Table 5-4: PBN Units 1 and 2 Primary Loop LOCAs Risk Quantification Results

'CDF (yr-1) 'LERF (yr-1)

Unit 1 2.201E-08 5.128E-11 Unit 2 3.673E-08 8.558E-11 E4-36

Enclosure 4 Overview of Risk-Informed Approach A bounding evaluation was performed for the risk contribution of the SSBIs using the PBN IE PRA model. The analysis conservatively assumed that all SSBIs that result in a consequential LOCA would fail the strainers due to the effects of debris. Table 5-5 shows the risk quantification results of the SSBIs.

Table 5-5: PBN Units 1 and 2 Risk Quantification Results for SSBIs

'CDF (yr-1) 'LERF (yr-1)

Unit 1 6.31E-07 1.43E-09 Unit 2 6.36E-07 1.44E-09 5.4 Comparison with RG 1.174 Guide The calculated risk increase due to failures caused by LOCA-generated debris (CDF and LERF) shown in Table 5-4 and Table 5-5, along with the baseline CDF and LERF from Table 5-3, is compared with the guidelines in RG 1.174, as shown in the figures below.

Figure 5-1: Acceptance Guidelines for Core Damage Frequency E4-37

Enclosure 4 Overview of Risk-Informed Approach Figure 5-2: Acceptance Guidelines for Large Early Release Frequency It is concluded that the risk increases of each unit due to failures caused by LOCA-generated debris are very small and fall well within Region III of the RG 1.174 guideline.

Note that, for each unit, the risk contributions of the primary side breaks and SSBIs are not aggregated when comparing with the RG 1.174 guideline. Using aggregated risk metrics does not provide a realistic picture of the true risk associated with the effects of debris on strainer performance because each of the two contributing hazards was evaluated in a bounding manner with different levels of conservatism.

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Enclosure 4 Overview of Risk-Informed Approach 6.0 Uncertainty Quantification Uncertainty quantification is a key requirement in RG 1.174 for a risk-informed evaluation (Reference 4). As defined in RG 1.174 and explained in more detail in NUREG-1855 (Reference 12) and two corresponding EPRI reports (References 13 and 14), there are three types of uncertainty that should be addressed:

x Parametric uncertainty x Model uncertainty x Completeness uncertainty Parametric uncertainty is the uncertainty in the value for a specific parameter.

Conservative or bounding values have been used for most parameters in the PBN risk quantification, and no uncertainty quantification was performed for these parameters. Due to the wide range of plant-specific post-LOCA conditions related to debris effects, understanding which parameters are uncertain along with the level of parametric uncertainty is important for understanding the overall uncertainty in the risk quantification results.

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 12). For example, the use of a spherical ZOI to model the debris quantity generated by a DEGB is a consensus model that has been widely adopted and accepted by the NRC (References 15 and 16). In general, PBN used models that have been widely accepted for deterministic evaluations. By using these consensus approaches, the effort to address model uncertainty is minimized.

Completeness uncertainty refers to 1) the uncertainty associated with scenarios or phenomena that are excluded from the risk evaluation, and 2) the uncertainty associated with unknown phenomena. Although it may not be practical to quantify the uncertainty associated with factors that are not explicitly evaluated, their potential impact can be qualitatively assessed. Uncertainties associated with unknown phenomena, on the other hand, cannot be directly evaluated (either quantitatively or qualitatively). Uncertainties associated with unknown phenomena are the reason that it is important to maintain defense-in-depth and safety margins (see Enclosure 5).

6.1 Parametric Uncertainty Parametric uncertainty is the uncertainty due to variabilities in specific input parameters.

If a bounding value is selected for an input parameter, it is essentially equivalent to a consensus model where the uncertainty does not need to be quantified. This is consistent with the guidance in Section C.4 of Draft RG 1.229 (Reference 3). The term consensus input is used to refer to conservative or bounding inputs that are consistent with general industry guidance, which has been accepted by the NRC in similar evaluations.

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Enclosure 4 Overview of Risk-Informed Approach The purpose of the parametric uncertainty quantification is to determine the overall uncertainties associated with the input parameters. For PBN, the parametric uncertainty was defined as the maximum change in the base case CDF when all the non-consensus inputs are changed to bounding values. Although it does not show the shape of the CDF probability distribution, the proposed method determines the maximum (or worst-case)

CDF, which helps understand the level of confidence that the risk is not significantly higher than the base case result.

Table 6-1 summarizes the key inputs for the risk quantification with an indication of which direction is more limiting in terms of strainer failures for PBN. The table also shows the inputs used in the base case and the parametric uncertainty cases. Note that the in-vessel downstream effects were analyzed outside of NARWHAL using a conservative methodology and combinations of inputs. Therefore, no uncertainty cases were run for reactor core failures.

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Enclosure 4 Overview of Risk-Informed Approach Table 6-1: Bounding (Worst Case) Direction for Important Input Parameters Bounding Parameter Base Case Input Parametric Uncertainty Case Input Direction ZOI Debris Quantities Maximum Consensus (Maximum) Same as Base Case Unqualified and ADQ Epoxy Maximum Consensus (Maximum) Same as Base Case Coatings Debris Quantity Latent and Miscellaneous Maximum Consensus (Maximum) Same as Base Case Debris Quantity Debris Transport Fractions Maximum Consensus (Maximum) Same as Base Case Minimum or Pool Volume/Level Minimum Minimum and Maximum Maximum Containment Pressure Minimum Consensus (Minimum) Same as Base Case Minimum or Same as Base Case (Maximum is Conservative Pool Temperature Design Basis (Maximum)

Maximum based on Parametric Sensitivity Results)

RHR, CS and SI Pump Flow Maximum Design Flow Rates Maximum Flow Rates from Hydraulic Analysis Rates during Injection RHR Pump Flow Rate Maximum from Hydraulic Maximum Increased by 10% from value in base case during Recirculation Analysis CS Pump Flow Rate during Value from Hydraulic N/A Recirculation Analysis Same as Base Case because CS and SI pump SI Pump Flow Rate during Value from Hydraulic flow rates do not impact strainer failure N/A Recirculation Analysis Function of Water ECCS/CS Switchover Time Minimum Function of Water Volume and Flow Rates Volume and Flow Rates Secure CS Time N/A Per EOP Requirement Same as Base Case Minimum or Maximum for release, pH Same as Base Case Maximum Minimum for solubility Head Loss Maximum Consensus (Maximum) Same as Base Case Structural Margin Minimum Design (Minimum) Same as Base Case NPSH Margin Minimum Consensus (Minimum) Same as Base Case Pump Void Fraction Limit Minimum Consensus (Minimum) Same as Base Case LOCA Frequency Value Maximum Nominal (mean) Maximum (95th percentile)

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As shown in Table 6-1, most of the parameters used to analyze strainer failures are consensus inputs based on the level of conservatism and consideration of competing effects. The list below highlights the inputs that were varied in the parametric uncertainty cases.

x Either a minimum or maximum pool volume could be more conservative with respect to strainer failures. Therefore, two parametric uncertainty cases were run using different pool volume inputs.

x For sump pool temperature, either a minimum or maximum sump pool temperature could end up being more conservative. However, the parametric sensitivity analysis showed that a maximum temperature is more conservative. Therefore, the design basis (maximum) temperature profiles are used for the uncertainty cases.

x During recirculation phase, the maximum flow rate for the RHR pumps based on hydraulic analysis was used in the base case. This flow rate was further increased by 10% in the parametric uncertainty cases.

x For the parametric uncertainty case, the LOCA frequencies used for calculating the CFPs and 'CDF are based on the 95th percentile of NUREG-1829 40-year geometric mean aggregation values (Reference 6). Note that the frequencies for 2-inch and 6-inch breaks were calculated using log-linear interpolation.

In summary, for each unit, two parametric uncertainty cases were performed with the only difference between the two cases being the pool volume inputs. The results of these parametric uncertainty cases for minimum and maximum pool volume are shown in Table 6-2 and Table 6-3 for Units 1 and 2, respectively.

Table 6-2: Results of Parametric Uncertainty Evaluation for Unit 1 Case Description CDF (yr-1) 1 Minimum water volume 7.306E-08 2 Maximum water volume 7.306E-08 Table 6-3: Results of Parametric Uncertainty Evaluation for Unit 2 Case Description CDF (yr-1) 1 Minimum water volume 1.250E-07 2 Maximum water volume 1.256E-07 As shown in the tables above, the parametric uncertainty cases have higher CDF values than the base cases (2.201E-08 yr-1 for Unit 1 and 3.673E-08 yr-1 for Unit 2). The increase is due mainly to the use of 95th percentile geometric LOCA frequency. Varying the pool volume inputs had little impact on the results.

Note that parametric uncertainty was quantified in a very conservative manner by using the worst-case combinations of inputs. Although the scenario is hypothetically possible, E4-42

Enclosure 4 Overview of Risk-Informed Approach the probability of all of the worst-case conditions occurring simultaneously is extremely low. The results of this evaluation show that the parametric uncertainty is low.

6.2 Model Uncertainty To meet the guidance in NUREG-1855 (Reference 12), model uncertainty was addressed for any models, for which no consensus exists. Most of the models used in the PBN risk quantification are consensus models that have been widely used by the industry and accepted by the NRC. However, the following models are not consensus models and were therefore included in the model uncertainty analysis:

x Break model x LOCA frequencies x LBLOCA size range discretization x Time step size To address the uncertainty in these models, alternative models were evaluated. Given the similarity between PBN Units 1 and 2 in terms of plant configurations and the resulting CDF of the base cases, model uncertainty was analyzed for Unit 2 only due to its slightly higher base case CDF value than Unit 1. The conclusion is applicable for both units.

The table below summarizes the alternative models used for model uncertainty analysis.

Table 6-4: Alternative Models Used to Quantify Model Uncertainty Model Uncertainty Model Base Case Notes Case Break Model Continuum DEGB-only model LBLOCA size ranges model changed to: 6-10, 10-20 and >20 LOCA PBN IE PRA NUREG-1829 arithmetic Frequencies for the 2 Frequencies model LOCA mean LOCA frequencies and 6-inch break frequencies sizes calculated by log-linear interpolation LBLOCA Size 6-15, 15-25 Two options with biased Range and >25 allocation of frequencies Discretization to smaller break sizes and larger break sizes Time Step Size 1 minute for Multiple cases with time After 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the in NARWHAL the first 24 steps of 2, 3, 4, 5 and 15 time-step size model hours minutes for the first 24 remained the same as hours base case.

Table 6-5 shows the CDF result for each of the model uncertainty cases. Figure 6-1 illustrates the change in CDF for each of the model uncertainty cases in comparison to the NARWHAL base case.

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Enclosure 4 Overview of Risk-Informed Approach Table 6-5: Results of Unit 2 Model Uncertainty Quantification Change in CDF Model in Base Case Model Uncertainty Case CDF from Base Case Continuum Break Model DEGB-Only Model 2.372E-07 2.004E-07 Top-Down Allocation of Top-Down Allocation of NUREG-LOCA Frequencies from 1829 Arithmetic Mean LOCA 5.406E-07 5.039E-07 PBN PRA Model Frequencies LBLOCA Size Range Bias 1 (6-10, 10-15, and 15-31 in) 4.803E-08 1.130E-08 Discretization (6-15, 15-25, and 25-31 inches) Bias 2 (6-20, 20-27, and 27-31 in) 8.389E-08 4.716E-08 2 minutes 3.673E-08 0.000E+00 3 minutes 3.673E-08 0.000E+00 Time Step Size 4 minutes 3.673E-08 0.000E+00 5 minutes 3.682E-08 8.785E-11 15 minutes 3.651E-08 -2.206E-10 1.0E-06 Alternate Models Base Case

'CDF 1.0E-07 1.0E-08 NUREG-1829 Arithmetic Time Step 15 minutes DEGB-Only Model Time Step 2 minutes Time Step 3 minutes Time Step 4 minutes Time Step 5 minutes LBLOCA Size Range Bias 1 LBLOCA Size Range Bias 2 Mean LOCA Frequency (6-10, 10-15 & 15-31 inches) (6-20, 20-27 & 27-31 inches)

Model Uncertainty Case Figure 6-1: Comparison of Unit 2 Model Uncertainty Cases to Base Case E4-44

Enclosure 4 Overview of Risk-Informed Approach As shown in the above figure, the model uncertainty runs for NUREG-1829 arithmetic mean LOCA frequencies, DEGB-only model and biasing of the LBLOCA size range result in higher CDF results than the base case. However, for these models, the overall results are still well within RG 1.174 Region III. The model uncertainty on time step has very little impact on the results.

6.3 Completeness Uncertainty Completeness uncertainty was qualitatively determined to be low. As described below, the PBN evaluation was rigorous and comprehensive, and the areas that were not explicitly evaluated have a low potential for any significant risk impact:

x Wide ranges of hazards, initiating events, and plant operating modes were considered as described in Sections 3.1, 4.3.1, and 4.3.2.

x A full spectrum of LOCAs were directly evaluated in the risk quantification as described in Section 4.0.

o The LOCA evaluation included pipe breaks on each ISI weld within the Class 1 pressure boundary inside the first isolation valve.

o Break sizes ranging from 1/2-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 Breaks outside of the first isolation valve were not quantitatively analyzed.

Even if these breaks result in any strainer failures, the risk contribution would be negligibly small due to the low likelihood of a simultaneous isolation valve failure or malfunction (e.g., failure to close, spuriously opening, or developing a large leak).

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, and low likelihood equipment configurations were addressed using a bounding approach (see Section 4.3.3).

x The risks associated with water hammer induced LOCAs, SSBIs, fire induced LOCAs, seismic induced LOCAs, and external hazards were evaluated and shown to be low as described in Sections 3.1 and 3.2.

x All known debris failure mechanisms associated with the GSI-191 phenomena and GL 2004-02 concerns were evaluated in a bounding or reasonably conservative manner.

Although there is also some uncertainty associated with unknown phenomena, this uncertainty is judged to be small. The nuclear industry has been actively addressing the GL 2004-02 concerns, most of which were originally identified in 1995 as part of GSI-191, 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 NRC and industry, as well as regulators and utilities around the world over the last 40 years to resolve issues related to debris and E4-45

Enclosure 4 Overview of Risk-Informed Approach strainer performance. These tests have investigated nearly every aspect of debris effects including:

x Insulation and coatings destruction from break jets x Unqualified coatings failure x Blowdown and washdown debris transport x Containment pool settling, tumbling, and lift-over-curb debris transport x Debris erosion x Chemical release, solubility, and precipitation x Strainer head loss and vortexing x Fiber penetration, in-vessel downstream effects and BAP x Ex-vessel component wear Based on the extensive research that has been performed, it is unlikely that there are unidentified phenomena that would significantly increase the risk of failures caused by LOCA-generated debris.

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Enclosure 4 Overview of Risk-Informed Approach 7.0 Technical Adequacy of PBN PRA Results This section provides justification for the technical adequacy of the PBN IE PRA model for use in the risk assessment. Note that the internal flood, internal fire, and other external hazards PRA models and analyses are only used to determine the total plant baseline CDF and LERF values. No specific GSI-191 evaluations were performed using these models. Therefore, the justification of technical adequacy described in this section is limited to the IE PRA model RG 1.200 (Reference 5, Section 4.2) requires the following information to demonstrate the technical adequacy of the PRA when used in a risk-informed application:

1. How the PRA model represents the as-built, as-operated plant.

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

3. Documentation that the parts of the PRA used in the application are performed consistently with the PRA standard as endorsed by RG 1.200 (Reference 5).

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

5. Identification of the key assumptions and approximations relevant to the results used in the decision-making process, as well as the peer reviewers assessment of those assumptions.

6. A discussion of the resolution of the peer review findings and observations that are applicable to the parts of the PRA required for the application.

7. Identification of the parts of the PRA used in the application that were assessed to have capability categories lower than required for the application (Reference 7, Section 1-3).

The following sections demonstrate how the PBN IE PRA model meets the technical adequacy requirements of RG 1.200 for the use in the risk quantification.

7.1 PRA Model of As-Built, As-Operated Plant The PBN Version 6.04 IE PRA model represents the as-built, as-operated plant configuration and is based on plant data as of December 2017. The PBN PRA uses the CAFTA software with fault tree linking methodology to perform model quantification. The PBN PRA models are controlled in accordance with guidance documents for PRA generation, maintenance, and updates.

The IE PRA model was developed to and is maintained in accordance with the ASME/ANS PRA standard (Reference 7) and RG 1.200 requirements (Reference 5).

Compliance with the standard includes periodic PRA model updates, reliability data updates, and peer reviews in order to ensure the model reflects the as-built, as-operated plant.

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Enclosure 4 Overview of Risk-Informed Approach 7.2 Plant Changes Not Incorporated in PRA The PBN PRA model maintenance and update process is governed by a guidance document, which ensures that all plant changes are systematically reviewed to identify PRA impact, determine significance, and schedule timely implementation. The guidance document also ensures that model updates occur on a periodic basis to ensure that the PRA model continually represents the as-built, as operated plant.

The PBN Model Change Database Open Items Report documents all currently unincorporated model change requests at PBN. This document was reviewed for model modifications with the potential to impact the risk quantification. This analysis does not directly use PBN IE model of record (MOR) to quantify the risk associated with GSI-191, but rather uses limited and specific fault trees for specific accident sequences within the model (i.e. RHR fault trees, Containment Event Tree, and main steam/feedwater line breaks). No unaddressed plant changes were identified with the potential to impact the results presented in this document.

7.3 Parts of the PBN PRA Used for Risk Assessment The IE PRA model was the primary model used in the PBN risk assessment. The IE model was not directly used to calculate CDF associated with the effects of debris following a LOCA. However, the model was used to calculate LERF using the CLERP, the RHR and CS equipment configuration probabilities, and the bounding CDF and LERF associated with the effects of debris following SSBIs. In addition, the IE, internal flooding, and internal fire PRA models were referenced to determine the baseline CDF and LERF values.

7.4 Summary of the Risk Assessment Methodology The PBN risk assessment methodology involves using the total CDF and LERF from all modeled internal and external hazards, calculating FPIE (Full Power Internal Events)

CDF based on LOCA frequency, calculating FPIE LERF using the CLERP and CDF results, and performing a bounding analysis of FPIE SSBIs to determine the risk impact of strainer clogging. No permanent PRA model modifications were required for this evaluation. The risk assessment methodology is described in more detail in Sections 2.0 and 4.8.

7.5 Key Assumptions and Sources of Uncertainty in the PRA The uncertainty associated with PRA modeling is addressed by making assumptions.

NUREG-1855 (Reference 12) defines assumptions as follows:

x An assumption is a decision or judgment that is made in the development of the PRA model. An assumption is either related to a source of model uncertainty or related to scope or level of detail.

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Enclosure 4 Overview of Risk-Informed Approach x An assumption related to a model uncertainty is made with the knowledge that a different reasonable alternative assumption exists. A reasonable alternative assumption is one that has broad acceptance in the technical community and for which the technical basis for consideration is at least as sound as that of the assumption being made.

x By contrast, an assumption that is related to scope or level of detail is one that is made for modeling convenience. Such assumptions result in defining the boundary conditions for the PRA model.

x An assumption is labeled key when it may influence (i.e., have the potential to change) the decision being made. Therefore, a key assumption is identified in the context of an application.

The PRA assumptions and modeling uncertainties that could potentially impact the risk-informed GSI-191 application were identified based on a review of the relevant PRA notebooks.

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Enclosure 4 Overview of Risk-Informed Approach Table 7-1: Evaluation of Relevant Assumptions in PBN IE PRA Model Key Relevant Assumption Impact on GSI-191 Assumption?

Subcriticality is not explicitly modeled for large LOCA This assumption does not play into any of the No sequences. This is due to the rapid depressurization that modeled fault trees used. This conservative causes the nuclear reaction to quickly shutdown due to assumption would therefore not result in any changes voiding in the core region. Additionally, given successful to the delta risk and does not impact the risk-reflood, subcriticality is assured by the boron concentration informed GSI-191 application.

of the injected emergency cooling water. Therefore, reactor trip is not modeled as a requirement for reactor shutdown. This is a conservative assumption, given the low probability of reactor trip failures.

The lubricating and seal cooling system for the RHR pumps The reliability of the lubricating and seal cooling No is considered part of the pump package failure rate and is system is unaffected by the functionality of the reactor not modeled separately. The CCW system cooling sump strainers. Given this fact and the low importance requirements for the RHR pump heat exchanger are of the CCW system relative to the parent component modeled. CCW is only required for pump operability during failure rates, this modeling assumption does not high and low head recirculation function. These boundary impact the risk-informed GSI-191 application.

conditions are typically insignificant relative to the parent component failure rates.

Flow diversion through the RHR to CS system pump suction The probability of such flow diversion is exceedingly No was not modeled in the fault tree. This modeling assumption low via this pathway (1.2E-15 for a given train) due to was made due to low probability of occurrence requiring spurious opening of the RHR to CS MOV (1(2)-RH-871A(B), 4E-08) and large internal leakage on the CS train suction check valve (1(2)-SI-858A(B),

3E-08). This modeling assumption is bounded by the current quantification of equipment configuration probabilities and does not impact the risk-informed GSI-191 application.

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Enclosure 4 Overview of Risk-Informed Approach Key Relevant Assumption Impact on GSI-191 Assumption?

A flow diversion created by transferring open or For a significant flow diversion to occur, the opposing No misalignment of the RHR heat exchanger cross-connect trains pump, discharge check valves and suction valves, RH-713A, RH-713B, RH-716C, and RH-716D was check valves would need to fail open spuriously or be not modeled in the fault tree. left misaligned after testing and maintenance. Given that the cross-connect valves are verified shut monthly via 1(2)-TS-ECCS-001, this event is both unlikely to be present and unlikely to cause a significant impact on the RHR pump trains. This modeling assumption is bounded by the current quantification of equipment configuration probabilities and does not impact the risk-informed GSI-191 application.

The HEP for operator failure to close containment isolation Given that the large LOCA event tree is not directly No (CI) valves following the failure of the valves to used in quantification of CDF or LERF and this automatically close is modeled with the Large LOCA conservatism does not impact the RHR or CS fault initiating event tree. trees, this modeling assumption does not impact the risk-informed GSI-191 application.

During the recirculation phase of RHR system operation, it Component cooling water was excluded from No is assumed that component cooling to the RHR pumps calculation of RHR FFPs due to being a common and heat exchangers is required for successful system dependency for both trains of RHR and CS for a given operation. This assumption is based on the RHR unit. Based on this exclusion, this modeling recirculation mission time, which can be almost 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, assumption does not impact the risk-informed GSI-which is longer than the pump seals can survive without 191 application.

component cooling. If component cooling is not provided to the pump seals, seal failure is assumed to occur. The heat exchangers are required to remove decay heat from the containment sump water to provide core cooling.

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Enclosure 4 Overview of Risk-Informed Approach Key Relevant Assumption Impact on GSI-191 Assumption?

The RHR min flow line is not required for successful This assumption is a realistic assumption that when a No system operation during recirculation. train of RHR is in recirculation mode, it will have adequate flow through the pump to provide adequate cooling and the min flow line will not divert sufficient flow from the main path to significantly impact RHR operation. This modeling assumption is bounded by the current quantification of equipment configuration probabilities and does not impact the risk-informed GSI-191 application.

Support systems for valves RH-624 and RH-625 are not Failure of instrument air or AC power to the 1(2)-RH- No modeled in the fault trees. 624 and 1(2)-RH-625 will cause the valve to fail in the open position; thereby allowing RHR to be successfully aligned. This modeling assumption does not impact the risk-informed GSI-191 application.

The flow diversion through SI-857A, Train A RHR to Train The valves required for this flow diversion event to No A SI Pump Crossover, and SI-857B, Train B RHR to Train occur are normally closed and checked monthly as B SI Pump Crossover, is not modeled in the fault tree. part of surveillance 1(2)-TS-ECCS-001. Based on low probability of occurrence, this modeling assumption is bounded by the current quantification of equipment configuration probabilities and does not impact the risk-informed GSI-191 application.

The misalignment of the valves, RH-709A/B, RHR Pump The discussed valves are checked monthly as part of No Discharge Valve, RH-715A/B RHR Heat Exchanger Inlet surveillance 1(2)-TS-ECCS-001. Based on low Valve, and RH-716A/B RHR Heat Exchanger Outlet Valve, probability of occurrence, this modeling assumption is after test or maintenance is not included in the fault tree. bounded by the current quantification of equipment configuration probabilities and does not impact the risk-informed GSI-191 application.

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Enclosure 4 Overview of Risk-Informed Approach Key Relevant Assumption Impact on GSI-191 Assumption?

Plugging of the flow orifices (7/8 ID) in the SI injection Given the 3/4 OD tubing present in the RHR heat No paths during recirculation mode as the water is drawn from exchangers tubes upstream of the orifices, no particles the containment sump is not included in the SI fault tree. In capable of plugging the 7/8 ID flow orifices in the SI addition, SI cold leg injection flow orifice plugging is not injection paths would be able to cross the heat assumed during the injection phase of the SI system due exchanger tubes during recirculation. Further, an to the purity of the pumped water from the RWST. evaluation of all effects downstream of the ECCS strainers has concluded that orifice blockages caused by debris laden fluid are not a concern with respect to GSI-191. Based on this conclusion, this modeling assumption does not impact the risk-informed GSI-191 application.

RHR heat exchanger tube plugging is not considered in the Given that the water pumped through the RHR heat No fault tree. exchanger tubes is from clean sources (RWST or RCS), this represents a realistic assumption. While additional debris sources have the potential to be introduced by LOCAs, an evaluation of all effects downstream of the ECCS strainers has concluded that tube blockages caused by debris laden fluid are not a concern with respect to GSI-191. Based on this conclusion, this modeling assumption does not impact the risk-informed GSI-191 application.

The failure mode of the RHR heat exchanger drain valves Given that the drain valves are normally closed, have No in the open position is not considered to result in a failure of a plug installed, and are manually operated, these the heat exchangers. valves are not modeled as a failure mode for the RHR heat exchangers. Further, these lines at <10% of the size of the RHR pump discharge line do not represent a flow diversion pathway. Based on this assessment, this modeling assumption does not impact the risk-informed GSI-191 application.

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Enclosure 4 Overview of Risk-Informed Approach 7.6 Assessment of PRA Model Technical Adequacy The PBN IE PRA model, data, and documentation has been subject to the following industry peer reviews:

x November 2010 full-scope peer review of the PBN IE PRA model, including Internal Flooding x October 2011 focused-scope peer review of the PBN Internal Events PRA model The PRA model was reviewed against the Capability Category (CC) II requirements from the ANS/ASME PRA standard (Reference 7), including clarifications imposed by RG 1.200 (Reference 5). The review was conducted by industry experts who are experienced in performing PRAs.

Following these peer reviews, PBN incorporated changes to their IE model to remedy these issues. The Appendix X finding closure process (References 24 and 25) was implemented to review and close out identified findings. These closures were reviewed and documented by Enercon Services, Inc. in July 2018. Of the 31 identified IE facts and observations (F&Os), 25 were closed and six remain open.

Table 7-2 summarizes the open F&Os and the resolutions with respect to use of the IE PRA model in the risk assessment.

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Enclosure 4 Overview of Risk-Informed Approach Table 7-2: Resolution of PBN Internal Events PRA Peer Review Findings F&O Review CC F&O Description F&O Impact on GSI-191 Evaluation Number Element IE-A1-01 IE-A1 Not Met Original Peer Review: PRA lacked systematic approach and The unmodeled Loss of 4kV bus initiating events (IE-1) IE-A5 CC-I documentation for treatment of special initiating events. Examples discussed in this finding would be excluded from IE-B2 Not Met given included loss of a 4kV bus, loss of HVAC. Discussion in the the analysis since it does not introduce debris IE-D2 Not Met PRA documentation needs more explanation for why not all special inside containment.

initiators were included.

This open F&O has no impact on the risk Utility Resolution: Additional explanation added to PRA 2.0, Rev. quantification.

6 to address this finding. A sensitivity run was done for NFPA 805 LAR and lnit 5b LAR that indicates CDF due to a failed 4,160 VAC Vital Switchgear bus initiator was between 1.9E-7 and 1.2E-9. LERF increase is between 3.9E-10 and 9.4E-12. Due to the low CDF/LERF values, these initiators are not considered significant. ltem #1 in model change database suggests addition of loss of 4,160 VAC bus as a special initiator be considered for adding to model.

F&O Closure Review: The assessment of initiating events is adequately documented in PRA 2.0 and includes a systematic assessment of special initiators including assessment of postulated loss of HVAC and loss of electrical buses. As a result of the assessment, these initiators are not modeled in detail. The sensitivity cases performed to support not modeling the loss of an AC bus show that there is only a very small impact to the baseline model if the loss of an AC bus is included as an initiating event.

Including loss of an AC bus as a modeled initiator is judged as a worthwhile model improvement to the baseline model and will improve capturing the risk of maintenance or testing on opposite train equipment.

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Enclosure 4 Overview of Risk-Informed Approach F&O Review CC F&O Description F&O Impact on GSI-191 Evaluation Number Element AS-B6- AS-B6 Not Met Original Peer Review: Electrical limitations, e.g., load As documented in the F&O Closure review, this 01 (IE-6) SY-A5 Met management failures may need to be considered in PRA model. F&O remains open due to an improvement SY-A21 Not Met being needed in the HRA documentation and/or SY-B6 Not Met Utility Resolution: Additional explanation added to PRA EDG system notebooks. Further it SY-B15 Met notebooks. All but one aspect was accepted by Oct 2011 peer demonstrated that the inclusion of the 3 of 4 review. That remaining aspect was addressed in Att. 1 of NPM EDGs failing to start probability would still result 2013-0099. Also see SY-A21-01 in negligible impact on the PRA model.

F&O Closure Review: The electrical limitations of concern were Based on this F&O being resolved except for addressed by imminent plant modifications. The Accident documentation updates which have no Sequence, EDG, 4kV System notebooks were updated to reflect quantitative impact on the PRA model, this open the impact of these modifications. The load management failures F&O has no impact on the GSI-191 evaluation.

were dispositioned as negligible without appropriate justification.

Procedures identify these actions as critical, but they are not present in the human failure analysis. The response to this finding aims to establish the failure to manage EDG loads as a negligible human failure event. In order to do this, the probability of 3 out of 4 EDGs failing to start should also be considered. The response only looks at the possibility of 3 of 4 EDGs failing to run and does not include failure to start.

AS-B7- AS-B7 Not Met Original Peer Review: Inadequate treatment of time-based As documented in the utility resolution, this 01 dependencies, e.g., recovery of offsite power, HVAC treatment, finding causes the current DC battery model to (IE-7) and battery depletion treatment only allow for limited recovery of offsite power due to not accounting for extra time afforded by Utility Resolution: The model has been improved in 3 areas to battery depletion. This results in conservative better reflect the impact of time phased events. CDF and LERF values while underestimating First, the Power Recovery Convolution has been revised. This the risk associated with batteries.

calculation determines the likelihood of the recovery of offsite power at the specific times that the MAAP and the RCP Seal LOCA The GSI-191 assessment does not directly analyses identified as being critical to the development of accident quantify CDF using the PRA model. Further, this sequences. The current Convolution analyses were developed conservative handling of LOOP recovery will specifically for SBO (no power available from any source) and are equally affect the availability of the RHR and CS therefore not applicable to a partial power situation such as LOOP. systems. Given that this assessment only Additionally, the modifications to the DC modeling resolve the bulk evaluates accident sequences where the sump of the cutsets in LOOP that give the appearance of being long term strainers are used, this open F&O has no impact SBO sequences. Second, the HVAC Notebook analyses have been on the GSI-191 evaluation.

revised. Additional consideration was given to the available E4-56

Enclosure 4 Overview of Risk-Informed Approach F&O Review CC F&O Description F&O Impact on GSI-191 Evaluation Number Element information and additional analyses were performed to quantitatively support the conclusions presented in the notebook.

Third, the modeling and tagging of battery depletion and the recovery rules for restoration of power to a DC bus have been revised. The previous model had a single tag to identify a depleted bus and the HEP dependencies are cued off of this tag. This resulted in the failure of a single DC bus effectively failing all DC power (a modeling error). This has been revised such that there is a unique tag for each DC bus and the cutsets in LOOP that looked like they should be in SBO (erroneous cutsets) have been modified to correctly reflect the loss of DC at a specific bus and not the loss of all DC power.

Focused PRA Peer Review: The model was improved in 3 areas to better reflect the impact of time phased dependencies as described above. HVAC notebook was updated, and model includes HVAC as appropriate. However, the Model is still conservative because LOOP recovery for nonSBO scenarios is still neglected and the basis for this is inadequate. Also, DC life is still assumed to be 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> when realistic battery life is much greater (DC notebook does not mention true battery life other than full load test takes 2 1/2 days). In the convolution analysis, credit is not even taken for the one battery hour. As a minimum greater detail is required to document these assumptions and their impact on the results (QU). Since the 5.00 model is being reviewed the QU results will not address additional model changes being incorporated NEXERA.

This F&O was not included in the F&O closure review scope.

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Enclosure 4 Overview of Risk-Informed Approach F&O Review CC F&O Description F&O Impact on GSI-191 Evaluation Number Element SY-A21- SY-A21 Not Met Original Peer Review: Excessive electrical loading concerns F&O kept open due to same basis as AS-B6-01 01 (documentation improvements needed to (IE-9) Utility Resolution: Additional explanation added to PRA provide basis of negligible risk impact).

notebooks. All but one aspect was accepted by Oct 2011 peer review. That remaining aspect was addressed in Att. 1 of NPM Based on this F&O being resolved except for 2013- 0099. Also see AS-86-01. documentation updates which have no quantitative impact on the PRA model, this open F&O Closure Review: The close-out of IE-9 is directly related to F&O has no impact on the GSI-191 evaluation.

the close-out of IE-6 (the same close-out argument was used for both IE-6 and IE-9 in NPM 2013- 099 for control of excess electrical loads being of a low probability. It is suggested that the "low probability" prediction should be strengthened to also include CCF of DG fail-to-start. Also, LOCA with consequential LOOP may be more probable than independent LOCA and LOOP.

HR-D1- HR-D1 Not Met Original Peer Review: Screening pre-initiators values were used Review of the HRA Notebook indicated that two 01 in the model. Use of screening values for all pre-initiators only unscreened events are HEP-AF--TY-1-190 (IE-12) meets CC I (1P53 suction manual valve 1AF-190 restoration error) and HEP-AF-TY-2-190 (2P53 suction Utility Resolution: Slightly larger screening values used in model manual valve 2AF-190 restoration error). Both (5E-4 vs. 1E-4) and cutsets evaluated. Two significant mis- pre-initiator events do not factor into the FFP positioning BEs were identified, and detailed analysis performed on calculations for RHR and CS; and have been them. removed from the model.

F&O Closure Review: This finding questioned the suitability of pre- Based on the technical issue detailed in this initiator screening based on HEP significance. At least one of the finding being remedied by model changes and screened human failure events showed a RAW >2, typically the HRA events themselves not being relevant accepted as the threshold for significance. The screening threshold to the quantification of RHR and CS was reset from 1E-4 to 5E-4 to identify significant pre-initiators that configuration probabilities, this open F&O has may have been inappropriately screened. Two initiators with a no impact on the GSI-191 evaluation.

Fussel-Vessely >0.005 or RAW >2 were discovered. While the response claims these two events were added to the model, they do not appear to be in the 5.02 CAFTA model.

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Enclosure 4 Overview of Risk-Informed Approach F&O Review CC F&O Description F&O Impact on GSI-191 Evaluation Number Element LE-C9- LE-C9 CC I Original Peer Review: No justification provided for equipment This modeling is a demonstrated conservatism 01 LE-C10 CC I survivability or human actions credited under adverse in the modeling of containment performance.

(IE-28) LE-C11 CC I environments. This is a consequence of the PBN IE model only LE-C12 CC I analyzing LERF releases and not being a full LE-D3 CC I Utility Resolution: No effort to meet requirements beyond CC 1 Level 2 model.

have been taken. Thus, this item is still open.

Based on the described technical issue being a conservatism with minimal impact on the LERF analysis, this open F&O has no impact on the GSI-191 evaluation.

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Enclosure 4 Overview of Risk-Informed Approach 7.7 Capability Categories for Parts of the PRA The PRA standard has over 265 individual IE supporting requirements (SRs); 11 of these SRs were determined to be not applicable to the peer review. Based on the F&O closure review, only eight SRs are not met, as summarized below.

x IE-A1 - IDENTIFY those initiating events that challenge normal plant operation and that require successful mitigation to prevent core damage using a structured, systematic process for identifying initiating events that accounts for plant-specific features. For example, such a systematic approach may employ master logic diagrams, heat balance fault trees, or failure modes and effects analysis (FMEA).

Existing lists of known initiators are also commonly employed as a starting point.

x IE-B2 - USE a structured, systematic process for grouping initiating events. For example, such a systematic approach may employ master logic diagrams, heat balance fault trees, or failure modes and effects analysis (FMEA).

x IE-D2 - DOCUMENT the processes used to select, group, and screen the initiating events and to model and quantify the initiating event frequencies, including the inputs, methods, and results.

x AS-B6 - If plant configurations and maintenance practices create dependencies among various system alignments, DEFINE and MODEL these configurations and alignments in a manner that reflects these dependencies, either in the accident sequence models or in the system models.

x SY-B6 - PERFORM engineering analyses to determine the need for support systems that are plant-specific and reflect the variability in the conditions present during the postulated accidents for which the system is required to function.

x SY-A21 -IDENTIFY system conditions that cause a loss of desired system function.

x AS-B7 - MODEL time-phased dependencies (i.e., those that change as the accident progresses, due to such factors as depletion of resources, recovery of resources, and changes in loads) in the accident sequences.

x HR-D1 - ESTIMATE the probabilities of human failure events using a systematic process.

Because approximately 97% of the SRs in the IE PRA model satisfy Capability Category II requirements and the specific non-conforming aspects of the model were confirmed to have no impact to the GSI-191 risk assessment, the PBN IE PRA meets the requirements of RG 1.200 and is adequate for this risk-informed application.

8.0 Defense-in-Depth and Safety Margin As described in RG 1.174 (Reference 4), sufficient defense-in-depth and safety margin must be maintained. Both of these aspects were evaluated in detail, as summarized in .

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Enclosure 4 Overview of Risk-Informed Approach 9.0 Monitoring Program PBN has implemented procedures and programs for monitoring, controlling, and assessing changes to the plant that have a potential impact on plant performance related to the effects of LOCA-generated debris. 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. Procedures have been implemented to ensure the containment building is free of loose debris, to verify the condition of the sump strainers, and to control unattended temporary materials in containment. Strict controls have been imposed on the types and quantities of materials that may be taken into containment. PBN has also implemented a coatings condition assessment monitoring program to ensure that the coatings debris limit will not be adversely impacted. Enclosure 5 of this submittal provides further details on these procedures and programs.

10.0 Quality Assurance Most of the PBN analyses and testing that informed the risk quantification were performed as safety related under either PBN or vendor quality assurance (QA) programs that are compliant with 10 CFR 50 Appendix B. The NARWHAL and BADGER software packages, which were used for the PBN analyses, were developed and are maintained by ENERCON as safety related items in accordance with ENERCONs 10 CFR 50 Appendix B QA program. The PRA evaluations were not performed as safety related but were performed under the vendors QA program.

11.0 Periodic Update of Risk-Informed Analysis Consistent with RG 1.200, reliability data, unavailability data, initiating event frequency data, human reliability data, and other similar PRA inputs are reviewed and updated to maintain the base PBN PRA model consistent with the as-built, as-operated plant. Per NextEra procedure, this review and update is to occur no later than every two refueling cycles and includes guidance for updates of risk-informed applications. The risk-informed analysis for responding to GL 2004-02 will be updated concurrently with these periodic PRA model updates. This satisfies the requirements of draft RG 1.229 (Reference 3) to update within 48 months following initial NRC approval or since the last update. This update will 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 will also include any new information on LOCA frequencies that may be developed. The intent of the update is to capture the effects of any plant changes, procedure changes, or new information on the risk-informed analysis and to confirm that the acceptance criteria are still maintained (Reference 3).

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Enclosure 4 Overview of Risk-Informed Approach 12.0 Reporting and Corrective Actions Nonconformances with existing evaluations, or problem identification, will be entered into the PBN corrective action program for evaluation and corrective actions, as appropriate.

Nonconforming conditions that make the containment sump inoperable will meet 10 CFR 50.73 reporting criteria for a condition prohibited by Technical Specifications (TS). PBN will also report to the NRC and take corrective actions in the event that the debris-related CDF and LERF exceed the acceptance criteria corresponding to the upper threshold for RG 1.174 Region III (i.e., 1 x 10-6 for CDF and 1 x 10-7 for LERF)

(Reference 4) in accordance with 10 CFR 50.72 and 10 CFR 50.73, as applicable.

13.0 License Application The specific requirements for the license application described in RG 1.174 (Reference

4) are addressed in Enclosures 1 and 2.

14.0 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 (ML103570354)

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 (ML12349A378)

3. Draft Regulatory Guide 1.229 (ML16062A016, ML17025A263), Revision 0, Risk-Informed Approach for Addressing the Effects of Debris on Post-Accident Long-Term Core Cooling

4. Regulatory Guide 1.174, Revision 3, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, January 2018

5. Regulatory Guide 1.200, Revision 2, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities, March 2009

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

7. 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

8. Generic Letter 2008-01, Managing Gas Accumulation in Emergency Core Cooling, Decay Heat Removal, and Containment Spray Systems, January 11, 2008 (ML072910759)

9. NextEra Letter NRC 2009-0015, Point Beach Nuclear Plant, Unit 1, Nine-Month Supplemental (Post-Outage) Response to NRC Generic Letter 2008-01, February 11, 2009 (ML090420473)

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Enclosure 4 Overview of Risk-Informed Approach

10. NextEra Letter NRC 2010-0026, Point Beach Nuclear Plant, Unit 2, Nine-Month Supplemental (Post-Outage) Response to NRC Generic Letter 2008-01, March 5, 2010 (ML100640194)

11. NRC Letter, Point Beach Nuclear Plant, Units 1 and 2 - Closeout of Generic Letter 2008-01 Managing Gas Accumulation in Emergency Core Cooling, Decay Heat Removal and Containment Spray Systems (TAC Nos. MD7864 and MD7865),

January 7, 2010 (ML100050172)

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

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

14. 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

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

16. 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

17. NRC Generic Letter 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation during Design Basis Accidents at Pressurized-Water Reactors, September 13, 2004

18. NUREG/CR-6850, Fire PRA Methodology for Nuclear Power Facilities, September 2005

19. NUREG/CR-6928, Industry-Average Performance for Components and Initiating Events at U.S. Commercial Nuclear Power Plants, Component Reliability Data Sheets 2015 Update, February 2017

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

21. NextEra Letter, Point Beach, Units 1 & 2, Resolution Option and Implementation Schedule for GSI-191 Closure, May 16, 2013 (ML13140A013)

22. NextEra Letter, Point Beach Nuclear Plant, Units 1 and 2, Updated Final Response to NRC Generic Letter 2004-02, December 29, 2017 (ML17363A253)

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

24. Final Revision of Appendix X to NEI 05-04/07-12/12-16, Close-Out of Facts and Observations (F&Os), February 21, 2017 (ML17086A431)

25. U.S. Nuclear Regulatory Commission Acceptance on Nuclear Energy Institute Appendix X to Guidance 05-04, 07-12, and 12-13, Close-Out of Facts and Observations (F&Os), May 3, 2017 (ML17079A427)

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Point Beach Nuclear Plant Licensing Submittal for a Risk-Informed Resolution of Generic Letter 2004-02 Enclosure 4 Overview of Risk-Informed Approach Attachment 1 Maintain GSI-191 Compliance (for Information Only)

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Attachment 1 Maintain GSI-191 Compliance (for Information Only)

1. Risk-Informed GSI-191 Design Basis With approval of the new license amendment request for the use of risk-informed methodology to resolve GL 2004-02, the new design basis for PBN will be that the risk increase associated with failures due to LOCA-generated debris is within RG 1.174 Region III (i.e., CDF less than 1E-06 yr-1 and LERF less than 1E-07 yr-1). Note that the CDF guideline is more limiting for PBN than the LERF guideline because the calculated LERF is approximately three orders of magnitude lower than the calculated CDF. Therefore, LERF would not be exceeded without also exceeding CDF.

As described in Section 4.3.3, PBN equipment lineups during post-LOCA recirculation operation can be represented by two scenarios based on whether there are one or two RHR pumps operating. The probability of any configurations with one RHR pump failing is less than 2.0% (Table 4-4). Using log-linear interpolation of the 40-year geometric mean LOCA frequencies from NUREG-1829 (Reference 6, Table 7.19) and a 98/2 split for the equipment configuration, failure of all breaks greater than or equal to 12 inches for cases where both RHR pumps are running and failure of all breaks greater than or equal to 8 inches for cases where only one RHR pump is running would result in a CDF value of 8.9E-07 yr-1 (0.98 x 8.54E-07 + 0.02 x 2.70E-06). Therefore, the risk quantification would remain in RG 1.174 Region III (i.e., a CDF less than 1E-06 yr-1) even if all breaks larger than 12 inches fail when both RHR pumps are available and all breaks larger than 8 inches fail when only 1 RHR pump is available, as long as none of the breaks smaller than these thresholds fail. This is illustrated in the figure below.

Figure 14-1: NUREG-1829 40-Year Geometric Mean LOCA Frequencies To maintain GSI-191 compliance, it is only necessary to define acceptable limits for breaks smaller than or equal to 12 inches when both RHR pumps operating and for breaks E4-65

Attachment 1 Maintain GSI-191 Compliance (for Information Only) smaller than or equal to 8 inches when only one RHR pump operating. As long as these limits are met, it ensures that an identified issue would not push the risk up into RG 1.174 Region II.

2. Available Debris Margins Both the strainer and in-vessel debris limits were addressed to ensure that breaks smaller than or equal to 12 inches would not fail when both RHR pumps operating and that breaks smaller than or equal to 8 inches would not failure when only one RHR pump operating.

The acceptable debris limits based on the tested and analyzed debris quantities are shown in Table 1-1.

Fine fiber (e.g., insulation fiber and mineral wool) impacts both strainer head loss and in-vessel effects. The maximum transportable fiber fine quantity in the sump pool that does not cause strainer failures results in a smaller margin than that for in-vessel effects and is therefore more limiting. As a result, the limits shown in Table 1-1 for insulation fiber fines and mineral wool are based on the strainer head loss debris limits. All other debris limits in Table 1-1 (e.g., Cal-Sil, coatings, latent debris and miscellaneous debris) only impact strainer head loss. Therefore, the limits for these debris categories are based on the debris loads on one strainer resulting from the strainer head loss testing and analysis.

These debris limits are applicable for both PBN units.

Table 1-1: PBN Sump Strainer Debris Limits Debris Types Debris Limit Unit Fiber Insulation Fines 42.70 ft3 Mineral Wool 41.94 ft3 Cal-Sil and Asbestos Cal-Sil 384.44 lbm Coatings Particulate 12.29 ft3 ADQ Epoxy Fine and Small Chips 1.52 ft3 Latent Debris 150 lbm Miscellaneous Debris 200 ft2 The available margin for a given debris type is the difference between its debris limit and largest transportable quantity for the range of break sizes considered. For each debris type, its maximum transported quantity to one RHR strainer is obtained from the risk quantification analysis for breaks less than or equal to 12 inches for two train operation and for breaks less than or equal to 8 inches for single train operation. Since some of the debris types are grouped in the risk quantification analysis, some processing was performed to separate the debris loads in accordance with the debris types shown in Table 1-1. Afterwards, the greater value between the two train and one train operations is used to determine the available debris margin.

Tables 1-2 and 1-3 summarize the available debris margins for PBN Units 1 and 2, respectively. Note that the volumes for the fiber debris (i.e., insulation fiber and mineral wool) are in LDFG equivalent volumes.

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Attachment 1 Maintain GSI-191 Compliance (for Information Only)

Table 1-2: PBN Unit 1 Debris Margins Current Plant Debris Debris Types Margins Unit Quantity Limit Fiber Insulation 12.26 42.70 30.44 ft3 Mineral Wool 40.79 41.94 1.15 ft3 Cal-Sil and Asbestos Cal-Sil 90.19 384.44 294.25 lbm Coatings Particulate 8.42 12.29 3.87 ft3 ADQ Epoxy Fine and Small Chips 1.145 1.52 0.37 ft3 Latent Debris 19 150 131 lbm Miscellaneous Debris 120 200 80 ft2 Table 1-3: PBN Unit 2 Debris Margins Current Plant Debris Debris Types Margins Unit Quantity Limit Fiber Insulation 10.88 42.70 31.82 ft3 Mineral Wool 32.02 41.94 9.92 ft3 Cal-Sil and Asbestos Cal-Sil 93.64 384.44 290.80 lbm Coatings Particulate 10.92 12.29 1.37 ft3 ADQ Epoxy Fine and Small Chips 1.378 1.52 0.14 ft3 Latent Debris 30 150 120 lbm Miscellaneous Debris 152 200 48 ft2

3. Application of Debris Margins for Operability Evaluation The values in Tables 1-2 and 1-3 are similar to the debris limits and margins for a deterministic design basis. These values can be used to perform a prompt operability determination following discovery of an unanalyzed debris source.

For example, if it was determined that insulation previously thought to be reflective metallic insulation (RMI) was actually fiberglass insulation, the quantity of fiberglass could be compared to the available margin to ensure that the total quantity does not exceed the acceptable limit for breaks smaller than or equal to 12 inches. A very simplistic assessment could be performed with a conservative assumption that the entire quantity of unanalyzed fiberglass fails as fine debris and transports to one strainer. Alternatively, a more refined assessment could be performed to determine the quantity of insulation within a bounding ZOI in the vicinity of the insulation (for a break up to 12 inches) and/or determine realistic transport fractions for the newly identified debris source.

If the quantity of additional debris does not exceed the available margin, the sump can be declared operable. During the next outage, the debris source could be removed, or the design basis calculation could be updated to reflect the reduction in available margin.

However, if the quantity of additional debris exceeds the available margin, the sump E4-67

Attachment 1 Maintain GSI-191 Compliance (for Information Only) would be declared inoperable. Appropriate Technical Specification conditions would be entered. This process is laid out in the figure below.

Figure 1-2: Illustration of operability evaluation for an unanalyzed debris source E4-68

Attachment 1 Maintain GSI-191 Compliance (for Information Only)

4. Application of Debris Margins for Future Plant Modifications Future plant modifications will also be assessed for its potential impact on GSI-191 compliance using the debris margins shown in the tables above. The process is illustrated in Figure 1-3.

Figure 1-3: Illustration of design modification process with respect to GSI-191 parameters E4-69

Point Beach Nuclear Plant Licensing Submittal for a Risk-Informed Resolution of Generic Letter 2004-02 Enclosure 5 Defense-in-Depth and Safety Margin Table of Contents 1.0 Introduction............................................................................................................ 3

2.0 Defense-in-Depth .................................................................................................. 3

2.1 Evaluation for RG 1.174 DID Philosophy ........................................................... 3

2.2 Detecting and Mitigating Adverse Conditions ..................................................... 5

2.3 Barriers for Release of Radioactivity ................................................................ 10

2.4 Emergency Plan Actions .................................................................................. 12

3.0 Safety Margin ...................................................................................................... 13

4.0 References .......................................................................................................... 20

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Enclosure 5 Defense-in-Depth and Safety Margin List of Acronyms BDBEE Beyond-Design-Basis External Event CDF Core Damage Frequency CET Core Exit Thermocouple CS Containment Spray CST Condensate Storage Tank DID Defense in Depth ECCS Emergency Core Cooling System ECA Emergency Contingency Action FLEX Diverse and Flexible Coping Strategies ISI In-Service Inspection LERF Large Early Release Frequency LOCA Loss-of-Coolant Accident MSLB Main Steam Line Break NPSH Net Positive Suction Head NRC Nuclear Regulatory Commission PRA Probabilistic Risk Assessment PWSCC Primary Water Stress Corrosion Cracking RCP Reactor Coolant Pumps RCS Reactor Cooling System RG Regulatory Guide RHR Residual Heat Removal RVLIS Reactor Vessel Level Indications RWST Refueling Water Storage Tank SAMG Severe Accident Management Guidelines SI Safety Injection PBN PBN Generating Station E5-2

Enclosure 5 Defense-in-Depth and Safety Margin 1.0 Introduction For the purpose of this Point Beach Nuclear Plant (PBN) risk-informed Generic Letter (GL) 2004-02 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 measures include mitigative design features and actions that address protection of the public from radiation in the event that a loss-of-coolant accident (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 radiation barriers at PBN.

Safety margin is defined as elements of the analysis that increase the confidence that a declared success is a success. Therefore, the safety margins identified in this enclosure are a combination of built-in conservatisms in the analyses and testing that increase the confidence that scenarios that go to success remain in success and why some scenarios that are assumed to fail might actually succeed.

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 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 PBN 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.

2.1 Evaluation for RG 1.174 DID Philosophy PBN is proposing a licensing basis change to use a risk-informed approach to address the concerns of GL 2004-02 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:

x A reasonable balance is preserved among prevention of core damage, prevention of containment failure, and consequence mitigation. As summarized in Enclosures 2 and 3, PBN has performed various physical and procedural changes, for example, installation of new strainers with increased surface areas and a reduced opening size to reduce strainer head loss and debris penetration, installation of flow diverters to prevent debris-laden fluid to directly reach the sumps, implementation of the standard design change process that identifies potential impact to GL 2004-02 compliance by planned modifications, and comprehensive program controls to ensure the debris load limits are not exceeded. The new risk-E5-3

Enclosure 5 Defense-in-Depth and Safety Margin informed elements of the analysis showed a very small increase in risk of containment or reactor failures related to the debris effects concerned in GL 2004-02, as demonstrated by the very small changes in core damage frequency (CDF) and large early release frequency (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.

x 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, reactor coolant system (RCS) leakage detection program, or containment cleanliness inspection activities. Therefore, the licensing change will not cause any over-reliance on these activities.

x 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). The proposed licensing basis change for the use of a risk-informed methodology does not change the redundancy, independence, and diversity of the emergency core cooling system (ECCS) or containment spray (CS) system. These systems have been fully analyzed relative to their contribution to nuclear safety through plant-specific probabilistic risk assessment (PRA). The risk contribution related to GL 2004-02 due to the proposed licensing basis change has also been evaluated for the full spectrum of LOCA events. Additionally, as described in Enclosure 4, Section 6.0, the uncertainties in the risk quantification were examined. Based on the results of the uncertainty quantification and a consideration of the significant conservatisms, it was concluded with high confidence that the risk associated with the effects of debris at PBN is very small and is within Region III of RG 1.174 guideline (Reference 1).

x 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 GL 2004-02 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 strengthened by the physical and procedural changes made by PBN. Additionally, the new risk-informed approach does not introduce any new common-cause failures or reduce the current plant defenses against common-cause failures.

x 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.

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Enclosure 5 Defense-in-Depth and Safety Margin The post-LOCA recirculation function is provided by the ECCS located inside the auxiliary building. During the recirculation phase, coolant spilled from the break and water collected from the containment spray is cooled and returned to the reactor coolant system by the residual heat removal (RHR) pumps which are aligned to take suction on the containment recirculation sump. This water is pumped back to the core and/or the suction of the safety injection (SI) and CS pumps through the residual heat removal heat exchangers. 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 which filters exhaust during normal and accident conditions to limit offsite releases. The proposed licensing basis change does not alter the design and operating requirements for ECCS and other support equipment.

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 cooling units and the CS system can provide sufficient heat removal from the containment atmosphere following an accident to maintain the post-accident 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.

x Defenses against human errors are preserved. The use of the risk-informed methodology in the GL 2004-02 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.

x The intent of the plants 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 GL 2004-02. As discussed above, the risk-informed analysis shows that the risk increase due to GL 2004-02 related failures is very small and meets the RG 1.174 acceptance criteria (Reference 1). Therefore, the intent of the plants design criteria is maintained.

2.2 Detecting and Mitigating Adverse Conditions For the purposes of GL 2004-02 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 PBN DID measures for detecting and mitigating adverse conditions in order to support the PBN application for a risk-informed approach to resolve GL 2004-02.

Inadequate strainer flow refers to the condition where significant pump cavitation occurs due to inadequate RHR pump net positive suction head (NPSH) margin associated with the high head losses across the sump strainers and debris bed. Additionally, E5-5

Enclosure 5 Defense-in-Depth and Safety Margin accumulation of debris on the strainer and high head loss through the debris bed would increase the structural loading on the strainer assembly and challenge the structural integrity of the strainer. For PBN, 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 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 following the latest NRC guidance (Reference 2).

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.

x PBN has Emergency Operating Procedures (EOPs) and Emergency Contingency Actions (ECAs), which provide the operators with guidance on monitoring sump strainer blockage. If sump blockage is detected, the procedure provides actions that operators can take to mitigate the condition.

x For small to medium break LOCAs, depletion of the refueling water storage tank (RWST) can be delayed by following guidelines which provide 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 recirculation switchover level. Therefore, sump recirculation may not be required and, in that case, sump blockage would not be an issue.

x The Technical Specification minimum required RWST volume is 275,000 gallons.

The total volume of the RWST is 289,504 gallons. As a result, additional inventory is available from the RWST, compared with the Technical Specification minimum volume credited in the analyses.

Several measures are in place to control the debris sources inside the PBN containment building.

x 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.

x To meet the requirements of the PBN Technical Specification 3.5.2 (ECCS-Operating), PBN has implemented procedures which require that, several walk-downs be performed by all personnel to ensure the containment building is 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.

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Enclosure 5 Defense-in-Depth and Safety Margin x For the Technical Specification Surveillance Requirement 3.5.2.6, PBN implemented a procedure to verify by visual inspection that the containment sump inlets are not restricted by debris and that the suction inlet debris screens show no evidence structural distress or abnormal corrosion.

x PBN implemented a procedure 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.

x PBN implemented a coatings condition assessment monitoring program in accordance with RG 1.54 (Reference 3), as supplemented by the 10CFR50.65, ASTM D5163-05A (Reference 4) and EPRI 1003102 (Reference 5) guidelines. The program also requires that coatings surveillance personnel meet the qualification requirements of ANSI N45.2.6 (Reference 6). The program ensures that the coatings debris limit will not be adversely impacted.

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, SI 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 to detect this condition in a timely manner. PBN 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 RG 1.97 (Reference 7). PBN has the following methods for detection of sump strainer blockage conditions.

x PBN procedures monitor flow rate, discharge pressure and motor current of the RHR pumps for any signs of pump cavitation, as an indication for sump strainer blockage.

x PBN has core exit thermocouple (CET) and reactor vessel level indications (RVLIS) in the control room to allow monitoring for any potential reduction in core cooling flow due to sump blockage.

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.

x The PBN ECAs 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 E5-7

Enclosure 5 Defense-in-Depth and Safety Margin the debris. This is also an important measure to avoid permanent pump damage that could be caused by the loss of suction condition.

x The PBN ECAs contain steps to shut down pumps that piggyback off a potentially cavitating RHR pump during recirculation.

x The PBN ECAs contain steps to initiate makeup to the RWST from the Makeup System, Holdup Tanks, opposite unit RWST, boric acid storage tank (BAST), Fuel Transfer Canal, and Spent Fuel Pool. This would restore some capability for pumps to take suction from the RWST.

x In response to the Nuclear Regulatory Commission (NRC) Order EA-12-049 (Reference 8), Mitigation Strategies for Beyond-Design-Basis External Event (BDBEE), PBN developed 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, 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 PBN plant design has simultaneous cold leg and upper plenum injection capability once the RWST is depleted and the RHR pumps have been realigned to sump recirculation. The intent of the simultaneous injection is to supply a portion of the cooling flow by the SI pump to the reactor core inlet to flush boron precipitate out of the core and prevent flow blockages that may inhibit post-LOCA cooling. Boric acid precipitation is not expected to occur before 4.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. The requirement for initiating the SI pump recirculation for boric acid precipitation control is within 3.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> after the start of the LOCA event.

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 CET temperature indication and reactor water level, as monitored by the RVLIS. An additional method for detection of a core blockage condition includes monitoring of containment radiation levels.

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

If recirculation has been established with flow maintained into the RCS, core exit temperature should be stable or slowly lowering during accident recovery.

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Enclosure 5 Defense-in-Depth and Safety Margin 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, PBN's functional restoration procedure would attempt to establish injection flow of clean water from the RWST. CETs are monitored during ECA usage as well.

x Reactor vessel water level is monitored and a decreasing water level could indicate a lower core region flow blockage. PBN employs the RVLIS to provide instrumentation for the detection of inadequate core cooling.

x 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 for PBN to mitigate an inadequate reactor core flow condition. Upon identification of an inadequate RCS inventory or an inadequate core heat removal condition, the EOPs and ECAs direct the operators to take actions to restore cooling flow to the RCS including:

x Reestablish SI and RHR flow to the RCS x Reduce RCS pressure by performing rapid secondary depressurization x Restart reactor coolant pumps (RCPs) and open pressurizer power operated relief valves 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 steam generator (SG) is available, the operator can start the RCPs associated with the available idle SGs.

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.

PBN has also implemented procedures per the severe accident management guidelines (SAMG), which provide the operator with actions to protect fission product boundaries E5-9

Enclosure 5 Defense-in-Depth and Safety Margin and return the plant to a controlled stable condition when the EOPs 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 PBN 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 GL 2004-02 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.

x The PBN ISI Program provides rules for the examination and testing of ASME Class 1, 2 and 3 components and component supports. The ISI Program Plan addresses those examinations and tests required by ASME Section XI and PBN E5-10

Enclosure 5 Defense-in-Depth and Safety Margin augmented ISI commitments. The integrity of the Class 1 welds, piping, and components are maintained at a high level of reliability through the inspection program. The PBN ISI Program also ensure that inspections are performed in accordance with the schedule requirements of the ASME code.

x PBN developed a program plan to manage the risk of Primary Water Stress Corrosion Cracking (PWSCC) degradation in Alloy 600 components and Alloy 82/182 welds. The plan is in accordance with 10 CFR 50.55a, ASME Code Cases N-722-1 (Reference 9) and N-770-2 (Reference 10), and NEI 03-08 (Reference 11). The plan identifies all Alloy 600/82/182 locations, ranks the locations based on their risks of developing PWSCC, provides inspection requirements, and presents mitigation/replacement options. Periodic inspections of the Alloy 600 components and Alloy 82/182 welds are covered in the ISI Program.

x RCS overpressure protection is provided by means of pressure relieving devices, as required by Section III of the ASME Boiler and Pressure Vessel Code. The system is also protected from overpressure at low temperatures by the Low Temperature Overpressure Protection System.

x The leak detection program at PBN is capable of early identification of RCS leakage to allow time for appropriate operator actions 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 GL 2004-02.

x Some of the operator actions outlined in the PBN 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 the containment.

2.3.3 Reactor Containment Integrity The PBN containment building is 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, CS system, and containment cooling units 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 cooling units, in conjunction with the CS system, can remove sufficient thermal energy following an accident to maintain the containment pressure below design values. Therefore, the containment building remains 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.

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Enclosure 5 Defense-in-Depth and Safety Margin Additionally, some of the operator actions outlined in the PBN SAMG procedures can help maintain integrity of the containment when directed by the EOPs. 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 increased complexity. Implementation of the proposed change would not result in any changes to the response requirements for emergency response personnel during an accident.

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Enclosure 5 Defense-in-Depth and Safety Margin 3.0 Safety Margin The GL 2004-02 testing and analyses have various built-in conservatisms, as summarized in the table below.

Table  Description of Safety Margin

1. Conservatism Credited as Topic Realistic Conditions Impact on Evaluation Safety Margin

1. Scenario All secondary side break Most (if not all) secondary side Overall likelihood of failure is Frequency scenarios that require strainer breaks would be successfully over-predicted for secondary recirculation are assumed to fail mitigated due to the relatively side breaks due to the effects of debris low strainer flow rates and debris loads for these scenarios

2. Thermal- No credit was taken for The post-LOCA containment NPSH margin is under-Hydraulics containment accident pressure in pressure would be significantly predicted, and degasification NPSH or degasification higher than that used in the and flashing are over-evaluations and minimal credit risk quantification predicted taken for flashing evaluation

3. Thermal- Maximum sump temperature Sump temperature profiles Chemical release (precipitate Hydraulics profile used for all break sizes would be significantly lower for quantities), degasification, smaller break sizes and flashing, are over-predicted

4. Water Level Full volume of pressurizer was Treating the entire pressurizer Sump pool water level and treated as a hold-up volume for volume as a hold-up is only pump NPSH margins are large break LOCAs (LBLOCAs) applicable for the breaks on under-estimated for above the top of the hot leg the safety/ relief lines at the LBLOCAs above the top of nozzles top of the pressurizer. The the hot leg nozzles sizes of these pipes limit the breaks to SBLOCAs and MBLOCAs.

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Enclosure 5 Defense-in-Depth and Safety Margin

1. Conservatism Credited as Topic Realistic Conditions Impact on Evaluation Safety Margin

5. Water Level Initial RWST level assumed to be The RWST level is typically Sump water level is under-at the Technical Specification maintained above the TS estimated. Flashing, minimum level minimum level degasification, and NPSH failures are over-estimated

6. 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

7. Debris Unqualified epoxy coatings fail as Epoxy coatings are likely to fail Unqualified coatings debris Generation 100% fine particulate in a range of sizes (including transport and particulate both fine particulate and chips) debris quantity on strainers are over-predicted

8. Debris All unqualified coatings are in the Some of the unqualified Particulate debris quantity on Generation lower containment coatings are located in the strainers is over-predicted upper containment and may not reach the sump pool

9. Debris The ZOIs for the main loop piping Shadowing walls or equipment The debris loads for some of Generation were grouped by loop and reduces debris loads. the breaks were over-truncated collectively based on predicted.

concrete walls and collective line-of-sight. Additionally, shadowing by the reactor or structures was not considered for reactor nozzle breaks

10. Chemical Maximum pH for chemical Consistent time-dependent pH Aluminum precipitate quantity Effects release and minimum pH for profile resulting in lower is over-predicted and solubility release and/or increased precipitates would form later solubility than predicted E5-14

Enclosure 5 Defense-in-Depth and Safety Margin

1. Conservatism Credited as Topic Realistic Conditions Impact on Evaluation Safety Margin

11. 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 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (whichever breaks that do exceed the over-predicted comes first) solubility limit would still have some aluminum in solution

12. Chemical It was assumed that all of the Some of the aluminum may Aluminum precipitate quantity Effects submerged aluminum reacts with not be sprayed or may be and strainer head loss are the sump fluid and that all submerged in a portion of the over-predicted unsubmerged aluminum reacts pool that does not interact with with the containment spray the fluid that recirculates through the containment sump strainer

13. Chemical The maximum amount of Some of the aluminum Aluminum precipitate quantity Effects aluminum coatings on the coatings on the pressurizer and strainer head loss are pressurizer were assumed to be may not be exposed, and the over-predicted destroyed by all breaks within the failed coatings reacts with the pressurizer compartment, and sump fluid gradually the aluminum contained in the failed coatings is instantly released into the sump pool

14. Chemical All insulation debris is assumed In reality, a large fraction of the Aluminum release from Effects to be in the sump for the debris would be captured in insulation is over-predicted, chemical release calculation upper containment, and the resulting in an over-prediction release of chemicals would be of the aluminum precipitate significantly reduced for quantity breaks where containment sprays are not initiated E5-15

Enclosure 5 Defense-in-Depth and Safety Margin

1. Conservatism Credited as Topic Realistic Conditions Impact on Evaluation Safety Margin

15. 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

16. 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)

17. Debris A conservatively low pool fill-up Calculation showed over 70% The strainer debris loads are Transport transport fraction of 15% was of debris could be transported over-predicted.

used for debris transport into the into the inactive cavity during inactive cavity pool fill-up

18. Debris Small and large pieces of LDFG Based on 30-day erosion test The quantity of fines Transport debris have a high erosion results, the erosion fraction for generated and subsequently fraction (10%) small pieces of fiberglass transported to the strainers 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

19. Debris Unqualified coatings fail after Unqualified coatings would fail Unqualified coatings that fail Transport pool fill-up but before start of gradually and may not fail until in upper containment after sump recirculation. As a result, much later in the event sprays are secured may not they would not be transported to transport to the lower inactive cavities but available for containment or strainer recirculation transport.

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Enclosure 5 Defense-in-Depth and Safety Margin

1. Conservatism Credited as Topic Realistic Conditions Impact on Evaluation Safety Margin

20. Strainer/Pump When evaluating strainer failures, In many cases, one type of The breaks that fail the Failures a break is assumed to fail if the debris exceeds the tested strainer acceptance criteria quantity of any one debris type quantity while other types of are over-predicted for this break exceeds the test debris are significantly below quantity the tested quantity

21. Strainer/Pump Miscellaneous debris (e.g., tags It is likely that a large portion The strainer surface area is Failures or labels) all transports to the of 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

22. 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)

23. Strainer/Pump Strainer failure is assumed in all It is likely that the strainer Strainer structural failures are Failures cases where the head loss could withstand higher head over-predicted exceeds the structural margin of losses than predicted, and the strainer even if a structural failure occurs, it may not result in a complete loss of functionality E5-17

Enclosure 5 Defense-in-Depth and Safety Margin

1. Conservatism Credited as Topic Realistic Conditions Impact on Evaluation Safety Margin

24. 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

25. Strainer/Pump Compression of voids was not Volume of the bubble would be Gas void fractions at the Failures credited reduced at the suction due to pumps are over-predicted increase in static pressure

26. 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 from effect on NPSH required margin is under-predicted RG 1.82. when gas voids are present

27. Core Failures The fiber penetration testing and The penetration of fiberglass Fiber penetration (and correlation ignores effects of fiber fines would be reduced by the subsequent accumulation and particulate interactions and accumulation of particulate within the reactor core) is accumulation of small and large and fiberglass fines and small over-predicted pieces of fiberglass on the pieces on the strainer strainer

28. Core Failures During penetration testing, the Bridging, which is expected to Core failures due to the number of strainer disks was occur at the plant strainer, accumulation of fiber debris reduced to increase spacing would decrease penetration are over-predicted between adjacent disks to since some of the fiber debris prevent bridging cannot reach the perforated surfaces.

29. Core Failures Fiber limits associated with core It is likely that significantly Core failures due to blockage and boron precipitation larger quantities of debris accumulation of fiber debris are based on bounding tests and could accumulate inside the inside the reactor core are analyses from WCAP-17788 reactor core without full over-predicted blockage E5-18

Enclosure 5 Defense-in-Depth and Safety Margin

1. Conservatism Credited as Topic Realistic Conditions Impact on Evaluation Safety Margin

30. Core Failures All breaks were evaluated for in- Debris accumulation in the Core failures due to vessel effects based on the hot core is significantly reduced for accumulation of fiber debris leg break (HLB) debris limits cold leg breaks (CLBs) and are over-predicted for CLBs these breaks are less likely to fail the acceptance criteria

31. Core Failures All RHR pump flow into the upper Only the portion of flow that Core failures due to plenum is assumed to reach the makes up for the core boil-off accumulation of fiber debris reactor core, without crediting the reaches the core are over-predicted reduction due to spill out of the broken hot leg E5-19

Enclosure 5 Defense-in-Depth and Safety Margin 4.0 References

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

2. NRC (ML19228A011). U.S. Nuclear Regulatory Commission Staff Review Guidance for In-Vessel Downstream Effects Supporting Review of Generic Letter 2004-02 Responses. September 2019.

3. Regulatory Guide 1.54. Guidance on Monitoring and Responding to Reactor Coolant System Leakage. May 2008. Revision 1.

4. ASTM D5163-05A. Standard guide for Establishing Procedures to Monitor the Performance of Coating Service Level 1 Coating Systems in an Operating Nuclear Power Plant.

5. EPRI Document 1003102. Guideline on Nuclear Safety Related Coatings. Revision 1.

6. ANSI N45.2.6. Qualification of Inspection, Examination and Testing Personnel for Nuclear Power Plants.

7. Regulatory Guide 1.97. Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions during and Following an Accident.

December 1980. Revision 2.

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

9. ASME Section XI Code Case N-722-1. Additional Examinations for PWR Pressure Retaining Welds in Class 1 Components Fabricated With Alloy 600/82/182 Materials.

10. ASME Section XI Code Case N-770-2. Alternative Requirements and Acceptance Standards for Class 1 PWR Piping and Vessel Nozzle Butt Welds Fabricated With UNS N06082 or UNS W86182 Weld filler Material With or Without Application of Listed Mitigation Activities.

11. NEI 03-08. Guideline for the Management of Material Issues. Revision 2.



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