Text

Application for Technical Specification Change to Adopt Risk-Informed Extended Completion Times - RITSTF Initiative 4B
	ML21039A648
Person / Time
Site:	Waterford
Issue date:	02/08/2021
From:	Gaston R; Entergy Operations
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	W3F1-2021-0003
	Download: ML21039A648 (162)
	v • d • e

~ Entergx Entergy Operations, Inc.

1340 Echelon Parkway Jackson, MS 39213 Tel 601-368-5138 Ron Gaston Director, Nuclear Licensing 10 CFR 50.90 W3F1-2021-0003 February , 2020 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

Subject:

Application for Technical Specification Change to Adopt Risk-Informed Extended Completion Times - RITSTF Initiative 4B Waterford Steam Electric Station, Unit 3 NRC Docket No. 50-382 Renewed Facility Operating License No. NPF-38 In accordance with the provisions of Section 50.90 of Title 10 of the Code of Federal Regulations (10 CFR), Entergy Operations, Inc. (Entergy) is submitting a request for an amendment to the technical specifications (TS) for Waterford Steam Electric Station Unit 3 (Waterford 3).

The proposed amendment would modify TS requirements to permit the use of Risk Informed Completion Times with the implementation of Nuclear Energy Institute (NEI) 06-09, Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS)

Guidelines. The attachments contain the following information.

o Attachment 1 provides a description and assessment of the proposed changes.

o Attachment 2 provides the existing TS pages marked up to show the proposed changes. Clean pages will be provided at a later date.

Additional information consistent with the implementation guidance contained in NEI 06-09 is contained in the Enclosures listed. Waterford 3 requests approval of the proposed license amendment by February 28, 2022 with the amendment being implemented within 120 days.

In accordance with 10 CFR 50.91(a)(1), Notice for Public Comment, the analysis concerning the issue of no significant hazards consideration using the standards in 10 CFR 50.92 is included in Attachment 1.

In accordance with 10 CFR 50.91(b)(1), Notice for Public Comment; State Consultation, a copy of this application is being provided to the designated Louisiana Official.

This letter contains no new commitments.

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Attachment 1 W3F1-2021-0003 Description and Assessment of the Proposed Change

W3F1-2021-0003 Page 1 of 9 Description and Assessment of the Proposed Change 1.0

SUMMARY

DESCRIPTION Entergy is requesting approval for proposed amendment to the Technical Specifications (TS),

Appendix A of Renewed Facility Operating License No. NPF-38 for Waterford Steam Electric Station, Unit 3 (Waterford 3).

The proposed amendment would modify the TS requirements related to completion times (CTs) for required actions (RAs) to provide the option to calculate a longer, risk-informed CT (RICT). A new program, the Risk-Informed Completion Time Program, is added to TS Section 6, Administrative Controls.

The methodology for using the RICT Program is described in NEI 06-09 Revision 0-A, "Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS)

Guidelines," (Reference 1) which was approved by the NRC on May 17, 2007. Adherence to NEI 06-09 is required by the RICT Program.

The NRC issued Technical Specifications Task Force (TSTF) Traveler TSTF-505, Revision 2, "Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b," (Reference 3) on July 2, 2018 and a final model safety evaluation (SE) approving TSTF-505, Revision 2 on November 21, 2019 (Reference 4). All references to TSTF-505 in this LAR are to Revision 2 of the document.

The proposed amendment is consistent with the methodologies presented in TSTF-505.

However, only those RAs described in Attachment 4 and Enclosure 1, as reflected in the proposed TS markup provided in Attachment 2, are proposed to be changed. This is because some of the modified RAs in TSTF-505 are not applicable to Waterford 3 and there are some plant-specific RAs not included in TSTF-505 that are included in the proposed amendment. This is consistent with the methodology described in NEI 06-09, Revision 0-A.

2.0 ASSESSMENT 2.1 Applicability of Published Safety Evaluation Entergy has reviewed TSTF-505, Revision 2 (Reference 3), and the model safety evaluation dated November 21, 2018 (Reference 4). This review included a review of the NRC staff's evaluation, as well as the supporting information provided to support TSTF-505 and the safety evaluation for NEI 06-09-A. As described in the subsequent paragraphs, Entergy has concluded that the technical basis is applicable to Waterford 3 and supports incorporation in the Waterford 3 TS.

The model safety evaluation for TSTF-505 discusses the applicable regulatory requirements and guidance, including the 10 CFR 50, Appendix A, General Design Criteria (GDC). Waterford 3's Updated Final Safety Analysis Report (UFSAR), Section 3.1, "Conformance with NRC General Design Criteria, provides an assessment against the GDC published in 1971. A review has determined that the Waterford 3 plant-specific requirements continue to meet Appendix A

W3F1-2021-0003 Page 2 of 9 GDC as related to the proposed changes. Therefore, the proposed changes are applicable to Waterford 3.

2.2 Changes to the Technical Specifications The proposed amendment would modify the Waterford 3 TSs in the following manner to incorporate the RICT Program.

Administrative Controls Section 6.5.19, which describes the RICT Program, is added to TSs and reads as follows. This is consistent with TSTF-505 and NEI 06-09, Revision 0-A and amended for the adjustments made to revision 2 of TSTF 505 following NRC review:

Risk Informed Completion Time Program This program provides controls to calculate a Risk Informed Completion Time (RICT) and must be implemented in accordance with NEI 06-09, Revision 0-A, Risk Managed Technical Specifications (RMTS) Guidelines. The program shall include the following:

a. The RICT may not exceed 30 days.

b. A RICT may only be used in MODE 1 and 2.

c. When a RICT is being used, any plant change within the scope of the Configuration Risk Management Program must be considered for the effect on the RICT.

1. For planned changes, the revised RICT must be determined prior to implementation of the change in configuration.

2. For emergent conditions, the revised RICT must be determined within the time limits of the Required Action Completion Time (i.e., not the RICT) or 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after the plant configuration change, whichever is less.

3. Revising the RICT is not required if the plant configuration change would lower plant risk and would result in a longer RICT.

d. Use of a RICT is not permitted for entry into a configuration which represents a loss of specified safety function or inoperability of all required trains of a system required to be OPERABLE.

2.3 LAR Contents and Regulatory Commitments In accordance with Section 4.0, Limitations and Conditions, of the safety evaluation for NEI 06-09, the following is provided:

1. Enclosure 1 identifies each of the TS Required Actions to which the RICT Program will apply, with comparison of the TS functions to the functions modeled in the probabilistic risk assessment (PRA) of the structures, systems and components (SSCs) subject to those actions.

2. Enclosure 2 provides a discussion of the results of peer reviews and self-assessments conducted for the plant-specific PRA models which support the RICT Program, as required by Subsection 4.2 of Regulatory Guide (RG) 1.200, (Reference 5).

W3F1-2021-0003 Page 3 of 9

3. Enclosure 3 is not applicable since each PRA model used for the RICT Program is addressed using a standard endorsed by the Nuclear Regulatory Commission.

4. Enclosure 4 provides appropriate justification for excluding sources of risk not addressed by the PRA models.

5. Enclosure 5 provides the plant-specific baseline core damage frequency (CDF) and large early release frequency (LERF) to confirm that the potential risk increases allowed under the RICT Program are acceptable.

6. Enclosure 6 is not applicable since the RICT Program is not being applied to shutdown modes.

7. Enclosure 7 provides a discussion of the licensee's programs and procedures that assure the PRA models that support the RICT Program are maintained consistent with the as-built, as-operated plant.

8. Enclosure 8 provides a description of how the baseline PRA model, which calculates average annual risk, is evaluated and modified for use in the Configuration Risk Management Program (CRMP) to assess real-time configuration risk, and describes the scope of, and quality controls applied to, the CRMP.

9. Enclosure 9 provides a discussion of how the key assumptions and sources of uncertainty in the PRA models were identified, and how their impact on the RICT Program was assessed and dispositioned.

10. Enclosure 10 provides a description of the implementing programs and procedures regarding the plant staff responsibilities for the RICT Program implementation, including risk management action (RMA) implementation.

11. Enclosure 11 provides a description of the implementation and monitoring program as described in NEI 06-09, Section 2.3.2, Step 7.

12. Enclosure 12 provides a description of the process to identify and provide RMAs.

13. Enclosure 13 provides a cross reference between the Waterford 3 custom TS and NUREG-1432, "Standard Technical Specifications Combustion Engineering Plants," Revision 4 (Reference 2), upon which TSTF 505 is based.

There are no regulatory commitments contained in this letter.

2.3 Optional Changes and Variations Entergy is proposing the following variations from the TS changes described in TSTF-505 Revision 2, or the applicable parts of the NRC's model safety evaluation dated

W3F1-2021-0003 Page 4 of 9 November 21, 2018. These options are noted as acceptable variations in TSTF-505 and the model safety evaluation.

The TSTF-505 markups applicable to Waterford 3 are based on NUREG-1432, Standard Technical Specifications - Combustion Engineering Plants (Reference 2).

However, Waterford 3 has custom TSs which use different format, and has some variation in numbering and titles than the Standard Technical Specifications (STS). Many of the differences between Waterford 3 TS and the Standard TS are administrative and do not affect the applicability of TSTF-505 to the Waterford 3 TS.

Individual LCO Required Actions (RA) modified by the proposed amendment to be included in the RICT program are identified in Enclosure 1. Notes regarding TSTF-505 refer to TSTF-505-A, Rev. 2. In some cases, Conditions or Actions were added to conform to TSTF-505. The designations of the Conditions and Actions refers to the new designation, not the previous designation. In some cases, TSTF-505 may include additional Actions for which Waterford 3 is not requesting approval. Only the Actions proposed to be modified are discussed.

Site-specific LCOs which are included in this request are consistent with the methods described in TSTF-505. In instances where site-specific required actions are included, each will refer to one of the inclusion criteria listed in TSTF-505 Revision 2, Item 18 of Section 2.3 as described below. Identification numbers have been added for clarity.

From TSTF-505:

It is understood that in order to be within the review performed for the South Texas Project lead plant submittal, the traveler will only modify Required Actions that specify that either:

1. a system be restored to OPERABLE status, or

2. require an instrument channel to be placed in trip, or

3. require isolating an inoperable isolation valve.

The above criteria will be referred to as Scope Criteria 1, 2 or 3.

An additional enclosure (Enclosure 13) is included which provides a cross-reference between the NUREG-1432 Required Actions included in TSTF-505 and the Waterford 3 Required Actions included in this license amendment request. The enclosure includes a summary description of the referenced Required Actions, which is provided for information purposes only and is not intended to be a verbatim description of the Required Actions.

The cross-reference in Enclosure 13 includes the following:

1. Waterford 3 Actions that have identical numbers to the corresponding NUREG-1432 Required Actions and are not deviations from TSTF-505, except for administrative deviations (if any) such as formatting. These deviations are administrative with no impact on the NRC's model safety evaluation dated November 21, 2018.

2. Waterford 3 Actions that have different numbering than the NUREG-1432 Required Actions and are an administrative deviation from TSTF-505 with no impact on the NRC's model safety evaluation dated November 21, 2018.

W3F1-2021-0003 Page 5 of 9

3. NUREG-1432 Required Actions that are not contained in the Waterford 3 TS. The corresponding TSTF-505 mark-ups for the Required Actions are not applicable to WF3. This is an administrative deviation from TSTF-505 with no impact on the NRC's model safety evaluation dated November 21, 2018.

4. There are several plant-specific LCOs and associated Required Actions for which Waterford 3 is proposing to apply the RICT Program that are variations from TSTF-505 as identified in Enclosure 13 with additional justification provided below.

The Waterford 3 TS are text based and have significant formatting differences than the grid/table based STS. The TS markups included in this submittal and the WF3 to STS mapping in Enclosure 13 are included to show a comparison to the STS.

Variations from the TSTF-505 that require a justification per Table 1 in Reference 3 are described below.

3.4.3.1 Pressurizer Heaters TSTF-505 notes that Pressurizer (PZR) Heaters for CEOG plants are generally not in the PRA model. Revision 2 of TSTF-505 also notes that inclusion of Pressurizer Heaters is a Condition Requiring Additional Technical Justification.

The Pressurizer Heaters are included in the Waterford 3 PRA model. However, the PRA model includes PZR heaters fail to de-energize which contributes to high pressurizer pressure. The TS function is for the heaters to operate, which is opposite of the PRA modeled function (success in PRA model is heaters shutting off).

The TSTF-505 discusses pressurizer heaters and notes:

Licensee must justify the ability to calculate a RICT for the condition, including how the system is modeled in the PRA, whether all functions of the system are modeled, and, if a surrogate is used, why that modeling is conservative.

Inoperability of the heaters would result in a decrease in pressure and limit operators ability to maintain pressure in the RCS and Pressurizer. A surrogate event of a stuck open Atmospheric Dump Valve (ADV) will be used as a PRA surrogate for this TS function for RICT evaluations. This represents a conservative and bounding PRA treatment for the loss of the heaters. A stuck open ADV would (like failed heaters) limit the ability of the system to maintain RCS pressure as desired. This failure also has secondary side impacts which ensures that the PRA related risk results would be bounding for the loss of Pressurizer Heater function for RICT considerations.

3.6.1.3 Containment Air Locks Revision 2 of TSTF-505 (Reference 3) notes Containment Air Lock as a Condition Requiring Additional Technical Justification.

The containment airlocks are not explicitly modeled in the PRA. This type of failure will be conservatively analyzed as an early containment failure. All actions included in the RICT program have at least one of the two airlock doors closed (operable). Actions are only applied when one of the two doors is inoperable.

W3F1-2021-0003 Page 6 of 9 3.7.1.6 Main Feedwater Isolation Valves The TSTF-505 has a comment stating that Conditions A and B do not specify a restoration action and Condition C is a default Condition, thus the LCO conditions were excluded. However, the corresponding Waterford 3 TS 3.7.1.6, Action for one or more MFIVs inoperable does contain a restoration action with a Completion Time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

Therefore, WF3 proposes to apply a RICT to the existing Waterford 3 TS 3.7.1.6, Action.

This is acceptable because the TSTF states that there may also be plant-specific TS to which changes of the type presented in the TSTF may be applied.

3.5.2 (ECCS)

TS 3.5.2 (ECCS) has four actions (a, b, c, and d). Actions a and b will be included in the RICT program. Action a is for conditions with one low pressure safety injection train inoperabe. Action b covers one or more ECCS subsystems inoperable, but for conditions other than a and 100% ECCS flow equivalent to a single subsystem available. Both conditions can be evaluated with the PRA model and included in RICT program consistent with the intent of TSTF-505.

Actions c and d will not be included in the RICT program. Action c is for both Low Pressure Safety Injection (LPSI) trains inoperable and represents a loss of function condition. Action d represents a condition with inadvertent actuation and injection into the Reactor Coolant System. This condition is not a standard LCO condition and not included in the RICT program.

3.6.2.1 (Containment Spray) and 3.6.2.2 (Containment Cooling System)

NUREG 1432 (Reference 2) includes a single technical specification for Containment Spray and Cooling Systems. Waterford 3 has individual TS for each system. The cooling function of the systems is redundant. However, iodine removal is a function of the Containment Spray system but not the Containment Cooling System. The Waterford 3 RICT program will include a completion time for a single Containment Spray train inoperable. The program will also include a completion time for one of the two trains of Containment Cooling being inoperable. Having both Containment Spray trains inoperable is a loss of iodine removal function and will not be included in the RICT program.

W3F1-2021-0003 Page 7 of 9

3.0 REGULATORY ANALYSIS

3.1 No Significant Hazards Consideration Determination Entergy has evaluated the proposed changes to the TS using the criteria in 10 CFR 50.92 and has determined that the proposed changes do not involve a significant hazards consideration.

Waterford 3, requests adoption of an approved change to the plant-specific technical specifications (TS), to modify the TS requirements related to Completion Times for Required Actions to provide the option to calculate a longer, risk-informed Completion Time. The allowance is described in a new program in Chapter 6, "Administrative Controls," entitled the "Risk-Informed Completion Time Program."

As required by 10 CFR 50.91(a), an analysis of the issue of no significant hazards consideration is presented below:

1. Do the proposed changes involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed changes permit the extension of Completion Times provided the associated risk is assessed and managed in accordance with the NRC approved Risk- Informed Completion Time Program. The proposed changes do not involve a significant increase in the probability of an accident previously evaluated because the changes involve no change to the plant or its modes of operation. The proposed changes do not increase the consequences of an accident because the design-basis mitigation function of the affected systems is not changed and the consequences of an accident during the extended Completion Time are no different from those during the existing Completion Time.

Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Do the proposed changes create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed changes do not change the design, configuration, or method of operation of the plant. The proposed changes do not involve a physical alteration of the plant (no new or different kind of equipment will be installed).

W3F1-2021-0003 Page 8 of 9 Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Do the proposed changes involve a significant reduction in a margin of safety?

Response: No.

The proposed changes permit the extension of Completion Times provided that risk is assessed and managed in accordance with the NRC approved Risk-Informed Completion Time Program. The proposed changes implement a risk-informed configuration management program to assure that adequate margins of safety are maintained. Application of these new specifications and the configuration management program considers cumulative effects of multiple systems or components being out of service and does so more effectively than the current TS.

Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

Based on the above, Entergy concludes that the proposed changes present no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

3.2 Conclusions 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 amendment will not be inimical to the common defense and security or to the health and safety of the public.

4.0 ENVIRONMENTAL EVALUATION Entergy has reviewed the environmental evaluation included in the model safety evaluation published on March 15, 2012 (77 FR 15399) as part of the Notice of Availability. Entergy has concluded that the NRC staff findings presented in that evaluation are applicable to Waterford 3.

The proposed changes would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or

W3F1-2021-0003 Page 9 of 9 would change an inspection or surveillance requirement. However, the proposed changes do not involve (i) a significant hazards consideration, (ii) a significant change in the types or 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 changes meet 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 changes.

5.0 REFERENCES

1. NEI 06-09 Revision 0-A. Risk-Informed Technical Specifications Intitative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines. NEI 06-09. ADAMS Accession No. ML12286A322. November 2006.

2. NUREG 1432 Revision 4. Standard Technical Specifications - Combustion Engineering Plants. ADAMS Accession No. ML12102A169. April 2012.

3. TSTF-505 Revision 2. Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b. ADAMS Accession No. ML18183A493. July 2018.

4. Final Revised Model Safety Evaluation of Traveler TSTF-505, Revision 2. ADAMS Accession No.18267A259. November 2019.

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

(ADAMS Accession No. ML090410014) March 2009.

Attachment 2 W3F1-2021-0003 Proposed Technical Specification Changes (Mark-up)

TABLE 3.3-1 (Continued) INSERT 1 TABLE NOTATION ACTION 1 - With the number of channels OPERAS ne less than required by the Minimum Channels OPERABLE requir , restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months o be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and/or open the protective system trip breakers.

ACTION 2 - With the number of channels OPERABLE one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may continue provided the inoperable channel is placed in the bypassed or tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . If the inoperable channel is bypassed, the desirability of maintaining this channel in the bypassed condition shall be documented by the On-Site Safety Review Committee in accordance with plant administrative procedures. The channel shall be returned to OPERABLE status prior to STARTUP following the next COLD SHUTDOWN.

u WATERFORD- UNIT 3 .3/4 3-4a AMENDMENT NO. 445; 188

TABLE 3,3-3 (Continued) u TABLE NOTATION

{a) Trip function may be bypassed in this MODE when pressurizer pressure is less than 400 psia; bypass shall be automatically removed when pressurizer pressure is greater than or equal to 500 psia.

{b) An SIAS signal is first necessary to enable CSAS logic.

The provisions of Specification 3.0.4 are not applicable . INSERT 1 ACTION 12 - With the number of OPERAS annels one less than the Total Number of Channels, res he inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

ACTION 13 - With the number of channels OPERABLE one less than the Total Number of Channels, STARTUP and/or POWER OPERATION and/or operation in the other applicable MODE{S) may continue provided the inoperable channel is placed in the bypassed or tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . If the inoperable channel is bypassed, the desirability of maintaining this channel in the bypassed condition shall be documented by the On-Site Safety Review Committee in accordance with plant administrative procedures.

u The channel shall be returned to OPERABLE status no later than prior to entry into the applicable MODE(S) following the next COLD SHUTDOWN.

With a channel process measurement circuit that affects multiple functional units inoperable or in test, bypass or trip all associated functional units as listed below:

Process Measurement Circuit Functional Unit Bypassed/Tripped

1. Containment Pressure - High Containment Pressure - High {ESF)

Containment Pressure - High {RPS)

2. Steam Generator Pressure - Steam Generator Pressure - Low Low Steam Generator AP 1 and 2

{EFAS)

WATERFORD- UNIT 3 3/4 3-17 AMENDMENT NO. 109, 154, 188

\ '

~

TABLE 3.3-3 (Continued) u TABLE NOTATION ACTION 14 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE, STARTUP and/or POWER OPERATION and/or operation in the other applicable MODE(S) may continue provided the following conditions are satisfied:

a. Verify that one of the inoperable channels has been bypassed and place the other inoperable channel in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

b. All functional units affected by the bypassed/tripped channel shall also be placed in the bypassed/tripped condition as listed below.

Process Measurement Circuit Functional Unit Bypassedn-ripped

1. Containment Pressure Circuit Containment Pressure - High (ESF)

Containment Pressure - High (RPS)

2. Steam Generator Pressure - Steam Generator Pressure - Low Low Steam Generator t>.P 1 and 2 (EFAS) u STARTUP and/or POWER OPERATION and/or operation in the other applicable MODE(S) may continue until the performance of the next required CHANNEL FUNCTIONAL TEST. Subsequent STARTUP and/or POWER OPERATION and/or operation in the other applicable MODE(S) may continue if one channel is restored to OPERABLE status and the provisions of ACTION 13 a_re satisfied.

ACTION 15 - With the number of OPERABLE channels one less than *the Total Number of Channels, restore the inoperable channels to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN wi the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ACTION 16 - With the number of O hannels one less than the Total Number of Channe re th channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months o th Ive inoperable and take the ACTION r ,..,....,,,. 1.5.

INSERT 1 U WATERFORD - UNIT 3 3/4 3-18 AMENDMENTCN00-fr, 154

. (J . I ti '1389

TABLE 3.3-3 (Continued)

TABLE NOTATION ACTION 17 - With *the number of .OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may continue provided the inoperable channel is placed in the tripped (D.C Relay energized) condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , the remaining Emergency Diesel Generator is OPERABLE, and the inoperable channel is restored to OPERABLE status within the next 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . Otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . The surveillance requirements of Table 4.3-2 are waived for all channels while this action requirement is in effect.

ACTION 18 - With more than one channel inoperable, or if the inoperable channel cannot be placed in the trip (D.C. Relay energized) condition, declare the associated Emergency Diesel Generator inoperable and take the ACTION required by Specification 3.8.1.1. The surveillance requirements of Table 4.3-2 are waived for all channels while this action requirement is in effect.

ACTION 19 - With the number of channels OPERABLE one less than the Total Number of Channels, STARTUP and/or POWER OPERATION and/or operation in the other applicable MODE(S) may continue, provided the inoperable channel is placed in the bypassed or tripped condition within 1 hour:

a. If the inoperable channel is to remain in the bypassed condition, the desirability of maintaining this channel in the bypassed condition shall be documented by the On-Site Safety Review Committee in u accordance with plant administrative procedures. The channel shall be returned to OPERABLE status no later than prior to entry into the applicable MODE(S) following the next COLD SHUTDOWN.

b. If the inoperable channel is required to be placed in the tripped condition, within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months either restore the channel to OPERABLE status or place the ch I in the bypassed condition. If the tripped channel can not be ret d to OPERABLE status in 48 s, be in at least HOT STAND within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in SHUTDOWN within t e following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months or place channel in bypass.

With a channe measur at affects multiple functional unit le or i or trip all associated functional uni INSERT 1 WATERFORD* UNIT 3 3/4 3-18a AMENDMENT NO. 47,143,154,188 u

TABLE 3.3-3 (Continued)

TABLE NOTATION Process Measurement Circuit Functional Unit Bypassed/Tripped

1. Steam Generator Pressure Steam Generator Pressure - Low

- Low Steam Generator .6P 1 and 2 (EFAS)

2. Steam Generator Level Steam Generator Level - Low Steam Generator 6P (EFAS)

ACTION 20 With the number of OPERABLE channels one less than the Minimum Channels OPERABLE, STARTUP and/or POWER OPERATION and/or operation in the other applicable MODE(S) may continue provided the following conditions are satisfied:

a. Verify that one of the inoperable channels has been bypassed and place the other inoperable channel in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . With a channel process measurement circuit that affects multiple functional units inoperable or in test, bypass or trip all associated functional units as listed below:

Process Measurement Circuit Functional Unit Bypassed/Tripped

1. Steam Generator Pressure Steam Generator Pressure - Low

- Low Steam Generator .6P 1 and 2 (EFAS)

2. Steam Generator Level Steam Generator Level - Low Steam Generator .6P (EFAS)

b. Restore at least one of the inoperable channels to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COL UTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . Subsequent operatio

t e applicable MODE(S) may continue if one channel is rest to OPERABLE status and the provisions of ACTION 19 are s sfied.

INSERT 1 WATERFORD - UNIT 3 3/4 3-18b AMENDMENT NO. 47, 143, 154, 225

REACTOR COOLANT SYSTEM 3/4.4.3 PRESSURIZER LIMITING CONDITION FOR OPERATION 3.4.3.1 The pressurizer shall be OPERABLE with :

a. A steady-state water volume greater than or equal to 26% indicated level (350 cubic feet) but less than or equal to 62.5% indicated level (900 cubic feet), and,

b. At least two groups of pressurizer heaters powered from Class 1E buses each having a nominal capacity of 150 kW.

APPLICABILITY: MODES 1, 2, and 3.

ACTION :

a. ne group of the above required pressurizer heaters

, restore at least two groups to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or e in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in H HU OWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b. ith the pressur er otherwise inoperable, be in at least HOT INSERT 1 STANDBY with th reactor trip breakers open within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN "thin the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

4.4.3.1.1 The pressurizer water volume shall be determined to be within its limit in accordance with the Surveillance Frequency Control Program.

4.4.3.1.2 The capacity of each of the above required groups of pressurizer heaters shall be verified to be at least 150 kW in accordance with the Surveillance Frequency Control Program.

4.4.3.1.3 The emergency power supply for the pressurizer heaters shall be demonstrated OPERABLE in accordance with the Surveillance Frequency Control Program by:

a. Verifying the above pressurizer heaters are automatically shed from the emergency power sources upon the injection of an SIAS test signal.

b. Verifying that the above heaters can be manually placed and energized on the emergency power source from the control room.

WATERFORD - UNIT 3 3/44-9 Amendment No. ~ . 249

EMERGENCY CORE COOLING SYSTEMS 3/4.5.2 ECCS SUBSYSTEMS - MODES 1, 2, AND 3 u LIMITING CONDITION FOR OPERATION 3.5.2 Two independent emergency core cooling system (ECCS) subsystems shall be OPERABLE with each subsystem comprised of:

a. One OPERABLE high-pressure safety injection train,

b. One OPERABLE low-pressure safety injection train, and

c. An independent OPERABLE flow path capable of taking suction from the refueling water storage pool on a safety injection actuation signal and automatically transferring suction to the safety injection system sump on a recirculation actuation signal.

APPLICABILITY: MODES 1, 2, and 3*#.

ACTION:

a. With one ECCS subsystem inoperable due to one low pressure safety injection A

train inoperable, restore the inoperable train to OPERABLE status within 7 days u r be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce pressurizer pressure to less than 1750 psia and RCS average temperature to less than INSERT 1 00°F within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

With one or more ECCS subsystems inoperable due to conditions other than (a) and 100% of ECCS flow equivalent to a single OPERABLE ECCS subsystem available, restore the inoperable subsystem to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months r be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce pressurizer pressure to less than 1750 psia and RCS average temperature to less than 500°F within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .*

With pressurizer pressure greater than or equal to 1750 psia.

With RCS average temperature greater than or equal to 500°F.

u WATERFORD - UNIT 3 3/4 5-3 AMENDMENT NO. -a-4, 164 MAY 2 5 2000

CONTAINMENT SYSTEMS CONTAINMENT AIR LOCKS LIMITING CONDITION FOR OPERATION 3.6.1.3 Each containment air lock shall be OPERABLE with:

a. Both doors cl'osed*except when the -air lock is being *used for nonnal transit entry and exit through the containment, then at least one air lock door shall.

be closed, and

b. An overall air lock leakage rate in-accordance with the :containment Leakage Rate Testing Program.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

a. With one containment air lock door inoperable:

1. Maintain at least the OPERABLE air lock door closed and either restore the inoperable air lock door to OPERABLE status within 24 hou~ or lock the O~ERABLE air lock door closed.

"'-.,,.,)

L________ .,r----:::=::-;,~

I.-INSERT

____ 1 I- 2. Operation may then continue until perfonnance of the next

. required overall air lock leakage test provided that the OPERABLE air lock door is verified to be locked closed at least once per 31 days.

3. Otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the ~ollowing 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

4. The provisions of Specification 3.0.4 are not applicable.

b. With the containment air lock inoperable, except as the result of an inoperable air lock door, maintain at least one air lock door closed; restore the inoperable air lock to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD UTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

INSERT 1 V

WATERFORD - UNIT 3 3/4 6-9 Amendment No.124 APR 1 0 1997

CONTAINMENT SYSTEMS 3/4.6.3 CONTAINMENT ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3.6.3 Each containment isolation valve shall be OPERABLE.*

APPLICABILITY: MODES 1, 2, 3, and 4.

INSERT 1 ACTION :

With the isolation valve inoperable for penetration(s) with closed system(s)

a. Restore the inoperable valve to OPERABLE stat

b. Isolate each affected penetration within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months y use of at least one deactivated automatic valve secured in the isolation position and verif affected penetration flow path is isolated once per 31 days, or

c. Isolate each affected penetration within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months se of at least one closed manual valve or blind flange and verify the affected penetration flow path is isolated once per 31 days, or

d. Be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months Note: Isolation devices in a high radiation area may be verified by use of administrative means.

For all other penetrations, with one or more of the isolation valve(s) inoperable, maintain at least one isolation valve OPERABLE in each affected penetration that is open and either:

e.

f.

Restore the inoperable valve(s) to OPERABLE status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , o Isolate each affected penetration within 4 hour~ use of at least one deactivated automatic valve secured in the isolation position and verify the affected penetration flow path is isolated once p~e~r ~3~1 ~d~a:ys~,; o;r- - : - - ~ - ~ - - -

g. Isolate each affected penetration within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months se of at least one closed manual valve or blind flange and verify the affected penetration flow path is isolated once per 31 days, or

h. Be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

The provisions of Specification 3.0.4 do not apply.

Locked or sealed closed valves may be opened on an intermittent basis under administrative control.

SURVEILLANCE REQUIREMENTS 4.6.3.1 Each containment isolation valve shall be demonstrated OPERABLE prior to returning the valve to service after maintenance, repair or replacement work is performed on the valve or its associated actuator, control or power circuit by performance of a cycling test and verification of isolation time.

WATERFORD - UNIT 3 3/4 6-19 AMENDMENT NO. 86, 190, 217

CONTAINMENT SYSTEMS CONTAINMENT VENTILATION SYSTEM LIMITING CONDITION FOR OPERATION 3.6.1 .7 Each containment purge supply and exhaust isolation valve (CAP 103, CAP 104, CAP 203, and CAP 204) shall be OPERABLE and may be open at no greater than the 52° open position allowed by the mechanical stop for less than 90 hours0.00104 days 0.025 hours 1.488095e-4 weeks 3.4245e-5 months per 365 days.

APPLICABILITY: MODES 1, 2, 3, and 4 .

ACTION :

a. With a containment purge supply and/or exhaust isolation valve(s) open for greater than or equal to 90 hours0.00104 days 0.025 hours 1.488095e-4 weeks 3.4245e-5 months per 365 days at any open position, close the open valve(s) or isolate the penetration(s) within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , otherwise be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b. With a containment purge supply and/or exhaust isolation valve(s) INSERT 1 having a measured leakage rate exceeding the Ii

ance Requirement 4.6.1 .7.2, store e valve(s) to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , erwise be in at least HOT ST AND BY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.6.1.7.1 The cumulative time that the purge supply or exhaust isolation valves are open during the past 365 days shall be determined in accordance with the Surveillance Frequency Control Program.

4 .6.1.7.2 Each containment purge supply and exhaust isolation valve with resilient material seals shall be demonstrated OPERABLE in accordance with the Containment Leakage Rate Testing Program.

4.6.1.7.3 Each containment purge supply and exhaust isolation valve shall be demonstrated OPERABLE in accordance with the Surveillance Frequency Control Program by verifying that the mechanical stops limit the valve opening to a position < 52° open .

WATERFORD - UNIT 3 3/4 6-15 Amendment No. 124, 213 , 249

CONTAINMENT SYSTEMS 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS CONTAINMENT SPRAY SYSTEM LIMITING CONDITION FOR OPERATION 3.6.2 .1 Two independent containment spray systems shall be OPERABLE with each spray system capable of taking suction from the RWSP on a containment spray actuation signal and automatically transferring suction to the safety injection system sump on a recirculation actuation signal. Each spray system flow path from the safety injection system sump shall be via an OPERABLE shutdown cooling heat exchanger.

APPLICABILITY: MODES 1, 2, 3, and 4* .

INSERT 1 ACTION:

a. With one containment spray system inopera restore the inoperable spray system to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; restore the inoperable spray system to OPERABLE status within the next 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b. With two containment spray systems inoperable, restore at least one spray system to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and be in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.6.2.1 Each containment spray system shall be demonstrated OPERABLE:

a. In accordance with the Surveillance Frequency Control Program by verifying that the water level in the containment spray header riser is> 149.5 feet MSL elevation.

b. In accordance with the Surveillance Frequency Control Program by verifying that each valve (manual, power-operated , or automatic) in the flow path that is not locked, sealed, or otherwise secured in position , is correctly positioned to take suction from the RWSP .

c. By verifying, that on recirculation flow , each pump develops a total head of greater than or equal to 219 psid when tested pursuant to the INSERVICE TESTING PROGRAM .

With Reactor Coolant System pressure> 400 psia.

WATERFORD - UNIT 3 3/4 6-16 AMENDMENT NO. 89 , 163, ~ . 249,250

CONTAINMENT SYSTEMS CONTAINMENT COOLING SYSTEM LIMITING CONDITION FOR OPERATION 3.6.2.2 Two independent trains of containment cooling shall be OPERABLE with one fan cooler to each train.

APPLICABILITY: MODES 1, 2, 3, and 4 .

ACTION:

INSERT 1 With one train of contain r-...e.C,5,,11:[}0 ing inoperable, restore the inoperable train to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; restore the inoperable containment cooling train to OPERABLE status within the next 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4 .6.2.2 Each train of containment cooling shall be demonstrated OPERABLE:

a. In accordance with the Surveillance Frequency Control Program by:

1. Starting each operational fan not already running from the control room and verifying that each operational fan operates for at least 15 minutes.

2. Verifying a cooling water flow rate of greater than or equal to 625 gpm to each cooler.

b. In accordance with the Surveillance Frequency Control Program by:

1. Verifying that each fan starts automatically on an SIAS test signal.

2. Verifying a cooling water flow rate of greater than or equal to 1200 gpm to each cooler.

3. Verifying that each cooling water control valve actuates to its full open position on a SIAS test signal.

WATERFORD - UNIT 3 3/4 6-18 Amendment No. 39, 131,165,249

PLANT SYSTEMS MAIN STEAM LINE ISOLATION VALVES (MSIVs)

LIMITING CONDITION FOR OPERATION 3.7.1.5 Two MSIVs shall be OPERABLE.

APPLICABILITY : MODE 1, and MODES 2, 3, and 4, except when all MSIVs are closed and deactivated.

ACTION:

MODE1 With one MSIV inoperable, restore the valve to OPERABLE status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or be in STARTUP within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

INSERT 1 MODES 2, 3 and 4 With one MSIV inoperable, close the valve within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months and verify the valve is closed once per 7 days. Otherwise, be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS Note: Required to be performed for entry into MODES 1 and 2 only.

4.7.1 .5 Each MSIV shall be demonstrated OPERABLE:

a. By verifying full closure within 8.0 seconds when tested pursuant to the INSERVICE TESTING PROGRAM.

b. By verifying each MSIV actuates to the isolation position on an actual or simulated actuation signal in accordance with the Surveillance Frequency Control Program.

WATERFORD - UNIT 3 3/4 7-9 AMENDMENT NO . 76 ,189,190,199, :M-9,250

PLANT SYSTEMS MAIN FEEDWATER ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3.7.1.6 Each Main Feedwater Isolation Valve (MFIV) shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

-INSERT

---~ 1 Note: Separate Condition entry is allowed for each valve.

ne or more MFIV inoperable, close and deactivate, or isolate the inoperable valve within 72 ho d verify inoperable valve closed and deactivated or isolated once every 7 days; otherwise ,

be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

The provisions of Specification 3.0.4 do not apply.

SURVEILLANCE REQUIREMENTS 4.7.1.6 Each main feedwater isolation valve shall be demonstrated OPERABLE:

a. By verifying isolation within 6.0 seconds when tested pursuant to the INSERVICE TESTING PROGRAM.

b. By verifying actuation to the isolation position on an actual or simulated actuation signal in accordance with the Surveillance Frequency Control Program.

WATERFORD - UNIT 3 3/4 7-9a AMENDMENT NO. 167,189,199,249, 250

3/4.7 PLANT SYSTEMS 3/4.7.1 .7 ATMOSPHERIC DUMP VALVES LIMITING CONDITION FOR OPERATION 3.7.1.7 Each Atmospheric Dump Valve (ADV) shall be OPERABLE*.

APPLICABILITY: MODES 1, 2, 3, and 4 ACTION:

a. With the automatic actuation channel for one ADV inoperable, restore the inoperable ADV to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or reduce power to less than or equal to 70% RATED THERMAL POWER within th t 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b. With the automatic actuation channels for both ADVs inop able, restore one ADV to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or reduce power to less h u I 70%

RATED THERMAL POWER within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . INSERT 1

c. With one ADV inoperable, for reason .....,- "~ <

~ -.-"'

OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months o in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

The provisions of Specification 3.0.4 are not applicable provided one ADV is OPERABLE .

SURVEILLANCE REQUIREMENTS 4.7.1.7 The ADVs shall be demonstrated OPERABLE:

a. By performing a CHANNEL CHECK in accordance with the Surveillance Frequency Control Program when the automatic actuation channels are required to be OPERABLE .

b. By verifying each ADV automatic actuation channel is in automatic with a setpoint of less than or equal to 1040 psia in accordance with the Surveillance Frequency Control Program when the automatic actuation channels are required to be OPERABLE.

c. By verifying one complete cycle of each ADV when tested pursuant to the INSERVICE TESTING PROGRAM .

ADV automatic actuation channels (one per ADV, in automatic with a setpoint of less than or equal to 1040 psia) are not required to be OPERABLE when less than or equal to 70% RATED THERMAL POWER for greater than 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

WATERFORD - UNIT 3 3/4 7-9b AMENDMENT NO. 499, ~ . 250

PLANT SYSTEMS EMERGENCY FEEDWATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.1.2 Three emergency feedwater (EFW) pumps and two flow paths shall be OPERABLE.

APPLICABILITY : MODES 1, 2, and 3.

INSERT 1 ACTION :

a. With one steam supply to the turbine-drive EFW erable, restore the steam supply to OPERABLE status within days r be in a least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUT OWN within the allowing 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b. With one steam supply to the turbine-dri en EFW pump and ne motor-driven EFW pump inoperable and the EFW flow p t able to deliver at I ast 100% flow to their respective steam generators, restore t team supply or m or-driven EFW pump to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months o be in at least HOT S ANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 ours .

c. With one steam supply to the turbine-driven EFW pump and th motor-driven EFW pumps inoperable and the EFW flow paths able to deliv rat least 100% flow to their respective steam generators, be in at least HOT STA BY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

d. With the EFW system inoperable for reasons other than those cribed in ACTION (a), (b), or (c), and able to deliver at least 100% flow to either am generator, restore the EFW system to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

e. With the EFW system inoperable for reasons other than those described in ACTION (a), (b), or (c), and able to deliver at least 100% combined flow to the steam generators, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

f. With the EFW system inoperable and unable to deliver at least 100% combined flow to the steam generators, immediately initiate action to restore the ability to deliver at least 100% combined flow to the steam generators. LCO 3.0 .3 and all other LCO ACTIONS requiring MODE changes are suspended until the EFW system is capable of delivering at least 100% combined flow to the steam generators.

g. Only as allowed by Surveillance Requ irements 4 .7.1.2(b) and 4.7.1.2(c), the provisions of Specifications 3.0.4 and 4.0.4 are not applicable to the turbine-driven EFW pump for entry into Mode 3.

WATERFORD - UNIT 3 3/4 7-4 Amendment No. 60 , 96, 11 ~, 173, 230

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64!1"-1 \ 5,&4\\ \ \ +!, +!,4\,- \(\/0\ \

PLANT SYSTEMS 3/4.7.3 COMPONENT COOLING WATER AND AUXILIARY COMPONENT COOLING WATER SYSTEMS LIMITING CONDITION FOR OPERATION 3.7.3 At least two independent component cooling water and associated auxiliary component cooling water trains shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION : INSERT 1 With only one comp ent cooling water and associated auxiliary component cooling water trai ERABLE, restore at least two trains to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months e in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.7.3 Each component cooling water and associated auxiliary component cooling water train shall be demonstrated OPERABLE :

a. In accordance with the Surveillance Frequency Control Program by verifying that each valve (manual, power-operated , or automatic) servicing safety-related equipment that is not locked, sealed , or otherwise secured in position, is in its correct position.

b. In accordance with the Surveillance Frequency Control Program by verifying that each automatic valve servicing safety-related equipment actuates to its correct position on SIAS and CSAS test signals .

c. In accordance with the Surveillance Frequency Control Program by verifying that each component cooling water and associated auxiliary component cooling water pump starts automatically on an SIAS test signal.

WATERFORD - UNIT 3 3/4 7-11 AMENDMENT NO. 200, 249

fLANT SYSTEMS 3L.!,1,! ULTIMATE HEAT SINK 11MIJING CONDITION FQB OP[RATION 3.7.4 Two independent trains of ultimate heat sink (UHS) cooling towers shall be OPERABLE with each train consisting of a dry cooling tower (OCT) and a wet aechanical draft cooling tower (WCT) and its associated water basin with:

a. A 11ini11wn water level in each wet tower basin of 971 (-9.77 ft HSL)

b. An average basin water temperature of less than or equal to eg*F.

c. Fans as r.equired by Table 3.7-3.

APPLICABILITY; HODES 1, 2, 3, and 4. INSERT 1 ACTION:

a. With 1 UHS train inoperable, res the inoperable train to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLO SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b. With both UHS trains inoperable, restore at least one UHS train to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

WATERFORD - UNIT 3 3/4 7-12 Amendment No.fl,1-2-3,254 FEB 1 j 199*1_

PLANT SYSTEMS LIMITING CONDITION FOR OPERATION (Contjnued)

ACTION: (Continued)

c. This action applies only when UHS tornado required equipment is inoperable.

With a Tornado Watch or Warning in effect with the forecast 7 day average ambient dry bulb temperature greater than 74°F, all 6 OCT tube bundles and all 9 DCTfans associated with the missile protected portion of both trains of the OCT shall be OPERABLE. With a Tornado Watch or Warning in effect with the forecast 7 day average ambient dry bulb temperature less than or equal to 74°F, all 6 OCT tube bundles and at least 8 DCTfans associated with the missile protected portion of both trains of the OCT shall be OPERABLE. If the number of tube bundles or fans OPERABLE is less than required, restore the inoperable tube bundle(s) or fan(s) to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , or be in at least HOT STANDBY within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

d. When Table 3.7-3 dry bulb temperature restrictions apply with UHS fan(s) inoperable, detennine the forecast ambient temperatures and verify that the minimum fan requirements of Table 3.7-3 are satisfied (required only if the associated UHS is OPERABLE). The more restrictive fan requirement shall apply when 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and 3 day average temperatures allow different configurations.

e. With either or both wet cooling tower basin cross-connect valves not OPERABLE for makeup, restore the valve(s) to OPERABLE status within 7 d or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or COLD OWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

INSERT 1 SURVEILLANCE REQUIREMENTS

4. 7.4. Each train of UHS shall be determined OPERABLE:

a. In accordance with the Surveillance Frequency Control Program by verifying the average water temperature and water level to be within specified limits.

b. In accordance with the Surveillance Frequency Control Program, by verifying that each wet tower and dry tower fan that is not already running, starts and operates for at least 15 minutes.

c. Verify that each wet tower basin cross-connect valve is OPERABLE in accordance with the INSERVICE TESTING PROGRAM.

WATERFORD - UNIT 3 3/4 7-13 AMENDMENT NO. 95, 123, -208,~. 254

PLANT SYSTEMS 3/4.7.6 CONTROL ROOM AIR CONDITIONING SYSTEM CONTROL ROOM EMERGENCY AIR FILTRATION SYSTEM LIMITING CONDITION FOR OPERATION 3.7.6.1 Two control room emergency air filtration trains (S-8) shall be OPERABLE. (Note 1)

APPLICABILITY: MODES 1, 2, 3, 4, 5, and 6 During load movements with or over irradiated fuel assemblies.

ACTION :

a. With one control room emergency air filtration train inoperable for reasons other than ACTION b, restore the inoperable train to OPERABLE status within 7 days. <( I INSERT 1 I

b. With one or more control room emergency air filtration trains inoperable due to inoperable control room envelope boundary in MODES 1, 2, 3, or 4, then perform the following:

1. Immediately initiate action to implement mitigating actions; and

2. Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , verify mitigating actions ensure control room envelope occupant exposures to radiological , chemical , and smoke hazards will not exceed limits; and

3. Within 90 days, restore the control room envelope boundary to OPERABLE status.

c. If the required ACTION and associated allowable outage times of ACTION a or b are not met in MODES 1, 2, 3, or 4, then be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

d. If the required ACTION and the associated allowable outage time of ACTION a is not met in MODES 5 or 6, or during load movements with or over irradiated fuel assemblies, then perform the following :

1. Immediately place OPERABLE control room emergency air filtration train in emergency radiation protection mode (or toxic gas protection mode if automatic transfer to toxic gas protection mode is inoperable); or

2. Immediately suspend load movements with or over irradiated fuel assemblies and operations involving CORE ALTERATIONS.

WATERFORD - UNIT 3 3/4 7-16 AMENDMENT NO. 115, 119, 170, 188, 218, 235

PLANT SYSTEMS CONTROL ROOM AIR TEMPERATURE - OPERATING LIMITING CONDITION FOR OPERATION 3.7.6.3 Two independent control room air conditioning units shall be OPERABLE.

APPLICABILITY*: MODES 1, 2, 3, and 4.

INSERT 1 ACTION:

a. With one control room air conditioning unit inop le, restore the inoperable unit to OPERABLE status within 7 days or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b. With two control room air conditioning units inoperable, return one unit to an OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.7.6.3 Each control room air conditioning unit shall be demonstrated OPERABLE:

a. In accordance with the Surveillance Frequency Control Program by verifying that the operating control room air conditioning unit is maintaining average control room air temperature less than or equal to 80 °F.

b. At least quarterly, if not performed within the last quarter, by verifying that each control room air conditioning unit starts and operates for at least 15 minutes.

During load movements with or over irradiated fuel assemblies, TS 3.7.6.4 is also applicable.

WATERFORD- UNIT 3 3/4 7-18a AMENDMENT NO. 115, 14g, 188, 218, 235, 249

PLANT SYSTEMS 3/4.7.12 ESSENTIAL SERVICES CHILLED WATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.12 Two independent essential services chilled water loops shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3, and 4 INSERT 1 ACTION:

With only one essential services chilled water loops to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months o e in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.7.12.1 Each of the above required essential services chilled water loop shall be demonstrated OPERABLE:

a. In accordance with the Surveillance Frequency Control Program by verifying that each valve (manual, power-operated , or automatic) servicing safety-related equipment that is not locked, sealed, or otherwise secured in position, is in its correct position .

b. In accordance with the Surveillance Frequency Control Program by verifying that the water outlet temperature is :,; 42 ° F at a flow rate of z 500 gpm.

c. Deleted

d. In accordance with the Surveillance Frequency Control Program , by verifying that each essential services chilled water pump and compressor starts automatically on a safety injection actuation test signal.

4.7.12.2 The backup essential services chilled water pump and chiller shall be demonstrated OPERABLE in accordance with Specification 4.7 .12.1 whenever it is functioning as part of one of the required essential services chilled water loops.

WATERFORD - UNIT 3 3/4 7-43 AMENDMENT~ 249

3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1 A.C. SOURCES OPERATING LIMITING CONDITION FOR OPERATION 3.8.1.1 As a minimum, the following A.C. electrical power sources shall be OPERABLE:

a. Two physically independent circuits between the offsite transmission network and the onsite Class 1E distribution system, and

b. Two separate and independent diesel generators, each with:

1. Diesel oil feed tanks containing a minimum one hour supply of fuel, and

2. A separate diesel generator fuel oil storage tank, and

3. A separate fuel transfer pump.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

a. With one offsite circuit of 3.8.1.1 a inoperable, demonstrate the OPERABILITY of the remaining offsite A.C. circuit by performing Surveillance Requirement 4.8.1.1.1 a within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter. Restore the offsite A.C. circuit to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN witht owing 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b. With one diesel generator of 3.8.1.1b inoperable: INSERT 1 (1) Demonstrate the OPERABILITY of the remaining A.C. circuits b performing Surveillance Requirements 4.8.1.1.1a (separately for each offsit A.C. circuit) within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter. If the die I generator became inoperable due to any cause other than an inoperable s port system, an independently testable component, or preplanned ma tenance or testing, demonstrate the OPERABILITY of the remaining OPERA LE diesel generator (unless it has been successfully tested in the last 24 ho rs) by performing Surveillance Requirement 4.8.1.1 .2a.4 within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months less the absence of any potential common mode failure for the remaining esel generator is demonstrated .

(2) Restore the diesel generator to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months r be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months , unless the following condition exists:

WATERFORD* UNIT 3 3/4 8-1 AMENDMENT NO. 23,92,126,166,199,216, 251

ELECTRICAL POWER SYSTEMS LIMITING CONDITION FOR OPERATION u ACTION: (Continued)

(a) The requirement for restoration to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months may be exten_ded to 1O days if a temporary emergency diesel generator is verified available, and (b) If at any time the temporary emergency diesel generator

availability cannot be met, either restore the temporary emergency diesel generator to available status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months (not to exceed

. to days from the time the permanent plant EDG originally became inoperable), or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLO SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

c. With one offsite A.C. circuit and one diesel generator of the above required A.C.

electrical power sources inoperable, demonstrate the OPERABILITY of the remaining off site A.C. circuit by performing Surveillance Requirement 4.B.1.1.1 a within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter; and, if the diesel generator became inoperable due to any cause other than an inoperable support system, an independently te~table component, or preplanned maintenance or testing, de*monstrate the OPERABILITY of the remaining OPERABLE diesel generator by performing Surveillance Requirement 4.8.1.1.2a.4 within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months (unle'ss it is already operating) unless the absence of any potential common mode failure for the remaining diesel generator is demonstrated. Restore at least one of the Inoperable sources to OPERABLE status within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTOO within the folloV'v'.ing 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . Restore the other A.C. power source (offsit circuit or, diesel generator) to OPERABLE status in accordance with the provisions of ACTION statement a orb, as appropriate, with the time requirement of tliat .ACTION statement based on the time of initial loss of the remaining inoperable A.C. power source. A successful test of diesel generator OPERABILITY per Surveillance Requirement 4.B.1.1.2a.4 performed under this ACTION statement satisfies the diesel generator test requirement of ACTION statement a or b.

I-----.

..INSERT 1

d. With one diesel generator inoperable, in addition to ACTION b. or c. above, verify that:

(1) All required systems, subsystems, trains, components, and devices that depend on the remaining <;)PERABLE diesel generator as a source of emergency power are also OPERABLE, and *

(2) When in MODE 1, 2, or 3, the steam-driven emergency feed pump is OPERABLE.

If these conditions are not satisfied within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLO SHUTDOWN within the f ollowjng 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

WATERFORD - UNIT 3 3/4 8-2 AMENDMENT NO. 23,126,166 V JUL 2 1 2000

ELECTRICAL POWER SYSTEMS INSERT 1 LIMITING CONDITION FOR OPERATION U ACTION: (Continued).

e. With two of the above required offsite A.C. circuits inoperable, res one of the inoperable offsite A.C, circuits to OPERABLE status within 24 hours o be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . Following restoration of one .

off site A.C. circuit, follow ACTION statement a with the time requirement of that ACTION statement based on the time of initial loss of the remaining inoperable offsite A.C. circuit. A successful test of diesel generator OPERABILITY per Surveillance Requirement 4.8.1.1.2a.4 performed under this ACTION statement satisfies the diesel generator test requirement of ACTION statement a.

f. With two of the above required diesel generators Inoperable, demonstrate the OPERABILITY of two offsite A.C. circuits by performing Surveillance Requirement 4.8.1.1.1 a within .1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter; restore one of the inoperable diesel generators to OPERABLE status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or be in* at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUT.DOWN _within the_ following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . Following restoration of one diesel generator, follow ACTION statement b with the time requirement of that ACTION statement based on the time of initial loss of the remaining inoperable diesel generator.

\..._)

WATERFORD - UNIT 3 3/4 8-2a AMENDMENT NO. 23,126, 166 JUL 2 1 2000

ELECTRICAL POWER SYSTEMS 3/4.8 .2 D.C. SOURCES OPERATING LIMITING CONDITION FOR OPERATION 3.8.2.1 As a minimum the following D.C. electrical sources shall be OPERABLE:

a. 125-volt Battery Bank No. 3A-S and one associated full capacity charger (3A1-S or 3A2-S).

b. 125-volt Battery Bank No. 3B-S and one associated full capacity charger (3B1-S or 3B2-S).

c. 125-volt battery Bank No. 3AB-S and one associated full capacity charger (3AB1-S or 3AB2-S) .

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION: INSERT 1

a. With one of the required battery banks inoperable, resto,e the inoperable battery bank to OPERABLE status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months o~

c::::::::::::=J at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b. With one of the required full capacity chargers inoperable, demonstrate the OPERABILITY of its associated battery bank by performing Surveillance Requirement 4.8.2.1a.1 within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , and at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter. If any Category A limit in Table 4 .8-2 is not met, declare the battery inoperable.

SURVEILLANCE REQUIREMENTS 4 .8.2.1 Each 125-volt battery bank and at least one associated charger shall be demonstrated OPERABLE:

a. In accordance with the Surveillance Frequency Control Program by verifying that:

1. The parameters in Table 4.8-2 meet the Category A limits, and

2. The total battery terminal voltage is greater than or equal to 125 volts on float charge.

WATERFORD - UNIT 3 3/4 8-9 AMENDMENT NO. 249

ELECTRICAL POWER SYSTEMS LIMITING CONDITION FOR OPERATION <Continued)

ACTION:

a. With one of the required divisions of A.C. ESF busses not fully energized, reenergize the division within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COL DOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b. With one A.C. SUPS bus either not energized from its associated INSERT 1 inverter, or with the inverter not connected to its associate D. _________.

bus: (1) reenergize the A.C . SUPS bus within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months and (2) reenergize the A.C . SUPS bus from its associated inverter connected to its associated D.C. bus within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

c. With one D.C. bus not connected to its associated battery bank, reconnect the D.C. bus from its associated OPERABLE battery bank within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months o~

6nc at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD S ~ ~ WN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

1INSERT 1 1 SURVEILLANCE REQUIREMENTS 4.8.3.1 The specified busses shall be determined energized in the required manner in accordance with the Surveillance Frequency Control Program by verifying correct breaker alignment and indicated voltage on the busses.

WATERFORD - UNIT 3 3/4 8-14 AMENDMENT NO. 249

REPLACE With INSERT 2 Pages 6-11 through page 6-13 not used WATERFORD - UNIT 3 6-10 AMENDMENT NO . 218 Next Page is 6--14

NEW PAGE INSERT

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INSERT 1 or in accordance with the Risk Informed Completion Time Program

INSERT 2 ADMINISTRATIVE CONTROLS 6.5.19 Risk Informed Completion Time Program This program provides controls to calculate a Risk Informed Completion Time (RICT) and must be implemented in accordance with NEI 06-09-A, Revision 0, "Risk-Managed Technical Specifications (RMTS) Guidelines."

The program shall include the following:

a. The RICT may not exceed 30 days;

b. A RICT may only be utilized in MODE 1 and 2;

c. When a RICT is being used, any change to the plant configuration, as defined in NEI 06-09-A, Appendix A, must be considered for the effect on the RICT.

1. For planned changes, the revised RICT must be determined prior to implementation of the change in configuration.

2. For emergent conditions, the revised RICT must be determined within the time limits of the Required Action Completion Time (i.e.,

not the RICT) or 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after the plant configuration change, whichever is less.

3. Revising the RICT is not required if the plant configuration change would lower plant risk and would result in a longer RICT.

d. For emergent conditions, if the extent of condition evaluation for inoperable structures, systems, or components (SSCs) is not complete prior to exceeding the Completion Time, the RICT shall account for the increased possibility of common cause failure (CCF) by either:

1. Numerically accounting for the increased possibility of CCF in the RICT calculation; or

2. Risk Management Actions (RMAs) not already credited in the RICT calculation shall be implemented that support redundant or diverse SSCs that perform the function(s) of the inoperable SSCs, and, if practicable, reduce the frequency of initiating events that challenge the function(s) performed by the inoperable SSCs.

e. The risk assessment approaches and methods shall be acceptable to the NRC. The plant PRA shall be based on the as-built, as- operated, and maintained plant; and reflect the operating experience at the plant, as specified in Regulatory Guide 1.200, Revision 2. Methods to assess the risk from extending the Completion Times must be PRA methods used to support this license amendment, or other methods approved by the NRC for generic use; and any change in the PRA methods to assess risk that are outside these approval boundaries require prior NRC approval.

Enclosure 1 W3F1-2021-0003 List of Revised Required Actions to Corresponding PRA Functions

W3F1-2021-0003 Page 1 of 21 List of Revised Required Actions to Corresponding PRA Functions

1. Introduction Section 4.0, Item 2 of the NRC Final Safety Evaluation (Reference 1) for NEI 06-09-A, Revision 0-A, Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines, (Reference 2) identifies the following needed content:

- The license amendment request (LAR) will provide identification of the TS Limiting Conditions for Operation (LCOs) and action requirements to which the RMTS will apply.

- The LAR will provide a comparison of the TS functions to the PRA modeled functions of the structures, systems, and components (SSCs) subject to those LCO actions.

- The comparison should justify that the scope of the PRA model, including applicable success criteria such as number of SSCs required, flow rate, etc., are consistent with licensing basis assumptions (i.e., 50.46 ECCS flowrates) for each of the TS requirements, or an appropriate disposition or programmatic restriction will be provided.

This enclosure provides confirmation that the Waterford 3 PRA models include the necessary scope of SSCs and their functions to address each proposed application of the Risk-Informed Completion Time (RICT) Program to the proposed scope TS LCO Conditions, and provides the information requested for Section 4.0, Item 2 of the NRC Final Safety Evaluation. The scope of the comparison includes each of the TS LCO conditions and associated required actions within the scope of the RICT Program. The WF3 PRA model has the capability to model directly or through use of a bounding surrogate the risk impact of entering each of the TS LCOs in the scope of the RICT Program.

Table E1-1 below lists each TS LCO Condition to which the RICT Program is proposed to be applied and documents the following information regarding the TSs with the associated safety analyses, the analogous PRA functions and the results of the comparison:

x Column "Proposed TS LCO Condition": Lists all of the LCOs and condition statements within the scope of the RICT Program.

x Column "SSCs Covered by TS LCO Condition": The SSCs addressed by each action requirement.

x Column "SSCs in PRA Model": Indicates whether the SSCs addressed by the TS LCO Condition are included in the PRA.

x Column "Function Covered by TS LCO Condition": A summary of the required functions from the design basis analyses.

x Column "Design Success Criteria": A summary of the success criteria from the design basis analyses.

x Column "PRA Success Criteria": The function success criteria modeled in the PRA.

W3F1-2021-0003 Page 2 of 21 x Column Disposition: Provides the justification or resolution to address any inconsistencies between the TS and PRA functions regarding the scope of SSCs and the success criteria.

Where the PRA scope of SSCs is not consistent with the TS, additional information is provided to describe how the LCO condition can be evaluated using appropriate surrogate events.

Differences in the success criteria for TS functions are addressed to demonstrate the PRA criteria provide a realistic estimate of the risk of the TS condition as required by NEI 06-09 Revision 0-A.

The corresponding SSCs for each TS LCO and the associated TS functions are identified and compared to the PRA. This description also includes the design success criteria and the applicable PRA success criteria. Any differences between the scope or success criteria are described in the table. Scope differences are justified by identifying appropriate surrogate events which permit a risk evaluation to be completed using the CRMP tool for the RICT program. Differences in success criteria typically arise due to the requirement in the PRA standard to make PRAs realistic rather than bounding, whereas design basis criteria are necessarily conservative and bounding. The use of realistic success criteria is necessary to accurately model the as-built as-operated plant, and to conform to capability Category II of the PRA standard (as required by NEI 06-09 Revision 0-A).

For those TS line items (and SSCs) at Waterford 3 that represent loss of TS-specified safety function identified in Table El-1, the success criteria assumed in the PRA model are consistent with the design basis criteria. Furthermore, there are no alternative SSCs (to those referenced in the TS) credited in the PRA model - used to fulfill a specified safety function for purposes of a PRA functional determination - where the SSCs referenced in the TS would be unavailable.

Examples of calculated RICT are provided in Table E1-2. The listed values are estimates.

Following program implementation, the actual RICT values will be calculated using the actual plant configuration and the current revision of the PRA model representing the as-built, as-operated condition of the plant (plus a CDF and LERF seismic penalty), as required by NEI 06-09 and the NRC safety evaluation and may differ from the RICTs presented.

References:

1. NRC Letter from Jennifer M. Golder (NRC) to Buff Bradley (NEI), "Final Safety Evaluation for Nuclear Energy Institute (NEI) Topical Report (TR) NEI 06-09, Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines. ADAMS Accession No. ML071200238. May 2007.

2. Nuclear Energy Institute (NEI) Topical Report (TR) NEI 06-09, "Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines,"

Revision 0-A. ADAMS Accession No. ML12286A322. October 2012.

W3F1-2021-0003 Page 3 of 21 Table E1-1 Proposed TS LCO SSCs Covered SSC in Functions Covered Design Success Criteria PRA Success Criteria Disposition Condition by TS LCO PRA by TS LCO Condition Condition Model Manual Trip Yes As a minimum, the Part of Diverse Reactor This function represents SSCs are modeled consistent with 3.3.1 (1) (push button reactor protective Trip System - success is actions to manually scram the TS scope and so can be directly Reactor Protective and trip/channel instrumentation a reactor trip (actuation of the reactor given an entry evaluated using the CRMP tool.

Instrumentation circuitry) channels and Control Element condition for the Standard bypasses of Technical Assemblies). Post Trip Actions and a The individual channels are not in Manual Reactor Trip Specification Table The manual trip function failure of the Reactor the PRA model, but failure (or setting 3.3-1 shall be is made up of two Protection System to to failure) of the manual action would OPERABLE. channels with a single automatically scram the bound the risk of an individual channel required to plant. This manual reactor channel.

1. Manual Reactor Trip initiate a trip. trip is in the PRA model.

Two channels of instruments See TS 3.3.1 and Table 3.3-1 3.3.2 SIAS Yes Safety Injection The manual SIAS function SIAS is in the PRA model. SSCs are modeled consistent with The Engineered Instrumentation Actuation Signal has 2 sets of 2 channels Automatic actuation is the TS scope and so can be directly Safety Features initiation given the with 1 set required to trip. triggered by either high evaluated using the CRMP tool.

Actuation System following signal The containment containment pressure or (ESFAS) demand: pressure, Pressurizer low pressurizer pressure. The Containment High Pressure and Instrumentation Pressure, and auto Both parameters include 3 Pressurizer Pressure Low

-Manually initiated actuation functions all out of 4 channel logic. The instrumentation/function will not have 1 Safety Injection SIAS have 4 channels each, SIAS actuation relays and a RICT. The TS action with one of Actuation (SIAS) -Containment High with two required to trip, manual initiation are these instrument channels Pressure and three is the minimum included in the logic. inoperable is to place the channel in

-Pressurizer Pressure number of operable bypass or the channel in trip in 1 Low channels. hour. This state (channel in bypass

-Automatic SIAS or trip) can be maintained beyond Actuation the 30-day backstop LCO limit.

3.3.2 CSAS Yes Containment Spray The manual CSAS CSAS is in the PRA model. SSCs are modeled consistent with The Engineered Instrumentation Actuation Signal function has 2 sets of 2 The actuation logic for the TS scope and so can be directly Safety Features initiation given the containment high pressure evaluated using the CRMP tool.

W3F1-2021-0003 Page 4 of 21 Table E1-1 Proposed TS LCO SSCs Covered SSC in Functions Covered Design Success Criteria PRA Success Criteria Disposition Condition by TS LCO PRA by TS LCO Condition Condition Model Actuation System following signal channels with 1 set includes 3 out of 4 channel (ESFAS) demand: required to trip. logic. The CSAS actuation *The Containment High Pressure Instrumentation The containment pressure relays and manual action instrumentation/function will not have

-Manually initiated and Auto actuation are also included in the a RICT. The TS action with one of 2 Containment Spray CSAS functions each have 4 logic. these instrument channels Actuation Signal -Containment High channels each, with two inoperable is to place the channel in (CSAS) Pressure* required to trip. bypass or trip in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . This state

-Automatic SIAS (channel in bypass or trip) can be Actuation maintained beyond the 30-day backstop LCO limit.

3.3.2 CIAS Yes Containment Isolation The manual CIAS CIAS is in the PRA model. SSCs are modeled consistent with The Engineered Instrumentation Actuation Signal given function has 2 sets of 2 The actuation logic for the TS scope and so can be directly Safety Features the following signal channels with 1 set containment and evaluated using the CRMP tool.

Actuation System demand: required to trip. pressurizer pressure (ESFAS) The containment include 3 out of 4 channel *The Containment High Pressure Instrumentation -Manually initiated pressure, Pressurizer logic. The CIAS actuation and Pressurizer Pressure Low CIAS Pressure, and Auto relays and manual action instrumentation/function will not have 3 Containment -Containment High actuation functions all are also included in the a RICT. The TS action with one of Isolation Actuation Pressure* have 4 channels each, logic. these instrument channels Signal (CIAS) -Pressurizer Pressure with two required to trip. inoperable is to place the channel in Low bypass or trip in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . This state

-Automatic CIAS (channel in bypass or trip) can be Actuation maintained beyond the 30-day backstop LCO limit.

3.3.2 MSI Yes Main Steam Line Both MSIVs close Both MSIVs close SSCs are modeled consistent with The Engineered Instrumentation Isolation Actuation the TS scope and so can be directly Safety Features Signal given the evaluated using the CRMP tool.

Actuation System following signal (ESFAS) demand: *The Containment High Pressure Instrumentation and Pressurizer Pressure Low

-Manually initiated MSI instrumentation/function will not have a RICT. The TS action with one of

W3F1-2021-0003 Page 5 of 21 Table E1-1 Proposed TS LCO SSCs Covered SSC in Functions Covered Design Success Criteria PRA Success Criteria Disposition Condition by TS LCO PRA by TS LCO Condition Condition Model Main Steam Isolation -Containment High these instrument channels (MSI) Pressure* inoperable is to place the channel in

-Pressurizer Pressure bypass or trip in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . This state Low* (channel in bypass or trip) can be

-Automatic CIAS maintained beyond the 30-day Actuation backstop LCO limit.

3.3.2 RAS Yes Switching safety The manual RAS function RAS is in the PRA model. SSCs are modeled consistent with The Engineered Instrumentation injection from the has 2 sets of 2 channels The actuation logic for low the TS scope and so can be directly Safety Features RWSP to the with 1 set required to trip. level of RWSP (which evaluated using the CRMP tool.

Actuation System containment sump The RWSP level and Auto signals RAS) includes 4 (ESFAS) given low level in the actuation functions each channels with 3 out of 4 Instrumentation RWSP. have 4 channels, with two logic required. The RAS required to initiate RAS. actuation relays and related 5 Safety Injection -Manually initiated RAS manual actions are also Recirculation (RAS) -RWSP Low Level included in the logic.

-Automatic RAS Actuation 3.3.2 EFAS Yes Initiation of EFW the The manual EFAS SG level and pressure SSCs are modeled consistent with The Engineered Instrumentation following signal. function and control valve instruments are in the PRA the TS scope and so can be directly Safety Features logic trip function each model (2 out of 4 channels evaluated using the CRMP tool.

Actuation System -Manually initiated have 2 sets of 2 channels need to function for (ESFAS) EFAS per SG required to trip. success). TS 3.3.2 7e Control Valve Logic is Instrumentation -SG Low Level with The SG level and Auto Manual initiation of EFAS is not explicitly in the PRA model.

high delta pressure actuation functions each also in the PRA model. However, the other related SG logic 7 Emergency -SG Low Level have 4 channels, with two and other instruments can be used a Feedwater Actuation - Automatic Actuation required to initiate EFAS. surrogate event to evaluate the LCO (EFAS) - Control Valve logic with the CRMP tool.

indicating SG low level 3.4.3.1 Pressurizer Yes The pressurizer shall Pressurizer Heaters are The Pressurizer Heaters The Pressurizer system (including Pressurizer Heaters be OPERABLE with: part of the RCS Pressure are included in the PRA heaters) is included in the PRA Control System (PPCS). model to ensure they shut model. The system related failures off/deenergize. Heater cause or contribute to high RCS

W3F1-2021-0003 Page 6 of 21 Table E1-1 Proposed TS LCO SSCs Covered SSC in Functions Covered Design Success Criteria PRA Success Criteria Disposition Condition by TS LCO PRA by TS LCO Condition Condition Model

b. At least two groups The design basis of the failure is modeled to cause pressure. Failure or inoperability of of pressurizer heaters PPCS is: high pressurizer pressure. the heaters is more likely to limit powered from Class 1E - Maintain reactor Pressurizer control and operators ability to add or maintain buses each having a coolant system pressurizer spray failure are pressurizer pressure.

nominal capacity of pressure at its also included in the PRA The PRA events in the model are not 150 kW. design setpoint of model. the ideal events to map this TS 2250 PSIA during function. Inoperability of the heaters small operating would result in a decrease in transients. pressure and limit operators ability

- Provide a means to maintain pressure in the of equalizing RCS/Pressurizer.

Pressurizer and A surrogate event of a stuck open RCS boron ADV will be used as a PRA concentrations. surrogate for this TS function for

- Minimize RICT evaluations. This represents deviations from conservative/bounding PRA setpoint using treatment for the loss of the heaters.

heaters and spray A stuck open ADV would (like failure flow control during heaters) limit the ability of the site to design maintain RCS pressure as desired.

This failure also has secondary side The heaters are impacts which ensures that the PRA coordinated with related risk results would be Pressurizer Spray flow to bounding for the loss of Pressurizer maintain reactor coolant Heater function for RICT system pressure. considerations.

Proportional Heaters and Backup Heaters work to maintain and/or increase system pressure. A high pressurizer pressure signal deenergizes all heaters.

W3F1-2021-0003 Page 7 of 21 Table E1-1 Proposed TS LCO SSCs Covered SSC in Functions Covered Design Success Criteria PRA Success Criteria Disposition Condition by TS LCO PRA by TS LCO Condition Condition Model 3.5.2 Emergency HPSI & LPSI Yes Two independent 1 of 2 HPSI pumps Same as design criteria SSCs are modeled consistent with Core Cooling components emergency core operating with injection the TS scope and so can be directly Systems (ECCS) - cooling system (ECCS) path for cold leg/hot leg evaluated using the CRMP tool.

ECCS Subsystems subsystems shall be injection. 1 of 2 HPSI OPERABLE with each pumps operating in High Pressure Safety subsystem comprised recirculation mode.

Injection of:

a. One OPERABLE 1 of 2 LPSI pumps high-pressure safety operating with injection Low Pressure Safety injection train, path for cold leg/hot leg Injection injection.

b. One OPERABLE low-pressure safety Shutdown Cooling One injection train, and LPSI Train with injection path and operating

c. An independent Shutdown Cooling Heat OPERABLE flow path Exchanger.

capable of taking suction from the refueling water storage pool on a safety injection actuation signal and automatically transferring suction to the safety injection system sump on a recirculation actuation signal.

3.6.1.3 Air Lock Doors Contain- 3.6.1.3 Each One of two containment Maintaining Primary The Air Lock doors are not explicitly Containment Air ment containment air lock airlock doors closed Containment Integrity in the PRA model, but the function Locks (Atmospheric shall be OPERABLE (same for each air lock) associated with this Technical and Dual) with: with minimal leakage. Specification is in the model.

W3F1-2021-0003 Page 8 of 21 Table E1-1 Proposed TS LCO SSCs Covered SSC in Functions Covered Design Success Criteria PRA Success Criteria Disposition Condition by TS LCO PRA by TS LCO Condition Condition Model

a. Both doors closed The PRA model includes the except when the air containment and containment lock is being used for leakage in the model. Since closure normal transit entry of the doors and leakage past the and exit through the doors are the attributes/functions containment, then at maintained in this Technical least one air lock door Specification, those attributes can be shall be closed, and explicitly mapped to existing PRA basic events and logic.

b. An overall air lock leakage rate in accordance with the Containment Leakage Rate Testing Program.

3.6.1.7 Containment Yes Each containment Containment Purge Purge lines are required to PRA model can map/simulate an Containment Purge, Supply, purge supply and valves must close/remain close/remain closed to have open/bypass containment condition.

Ventilation System & Exhaust exhaust isolation valve closed on demand (with successful containment For RICT evaluations, the RICT Valves shall be OPERABLE minimal leakage). isolation. calculation tool will provide a and may be open at no completion time based on the greater condition (failed valve/breached than the 52° open containment) independent of the 90 position allowed by the hour limit listed in the current mechanical stop for Technical Specification.

less than 90 hours0.00104 days 0.025 hours 1.488095e-4 weeks 3.4245e-5 months per 365 days.

3.6.3 Containment Yes Each containment Varies by Varies by valve/line/system Many containment isolation valves Containment Isolation Valves isolation valve shall be valve/line/system are directly included in the PRA Isolation Valves OPERABLE Open/remain open for model. The valves are modeled One of 2 isolation devices systems who remain open consistent with the TS scope and so per penetration. on CIAS signal. will be directly evaluated using the CRMP tool.

W3F1-2021-0003 Page 9 of 21 Table E1-1 Proposed TS LCO SSCs Covered SSC in Functions Covered Design Success Criteria PRA Success Criteria Disposition Condition by TS LCO PRA by TS LCO Condition Condition Model At least one of two valves Not all containment isolation closing/remaining closed for valves/lines are modeled for LERF systems/lines required to as some are screened due to release close. path size or other considerations.

Bounding surrogates can be used to ensure delta CDF and LERF results are included for CRMP evaluations.

Those SSCs that are modeled consistent with the TS scope can be directly evaluated using the CRMP tool.

The CRMP evaluation can and will vary depending on the penetration and containment isolation valve related to the specific LCO. If it is a system needed to stay or remain open for accident mitigation, the function can be mapped to a failure to open/remain open. If the valve is required to close or stay closed then such failures can be mapped. If the valve/system is not explicitly in the PRA model but the isolation function is to ensure containment integrity/isolation, containment breach or leakage can be mapped for the CRMP evaluation.

The success criteria in the PRA are consistent with the design basis criteria.

W3F1-2021-0003 Page 10 of 21 Table E1-1 Proposed TS LCO SSCs Covered SSC in Functions Covered Design Success Criteria PRA Success Criteria Disposition Condition by TS LCO PRA by TS LCO Condition Condition Model 3.6.2.1 Containment Yes Two independent One of two containment One of two containment SSCs are modeled consistent with Containment Spray Spray system containment spray spray trains. spray trains the TS scope and so can be directly components systems shall be evaluated using the CRMP tool.

OPERABLE with each The CS System is spray system capable designed to remove heat The success criteria in the PRA are of taking suction from from the containment consistent with the design basis the RWSP on a atmosphere during and success criteria.

containment spray following a LOCA or a actuation signal and MSLB inside containment.

automatically transferring suction to The CS System provides the safety injection capability for the fission system sump on a product scrubbing (reduce recirculation containment actuation signal. Each concentrations of spray system flow path elemental iodine isotopes) from the safety of the containment injection system sump atmosphere following a shall be via an LOCA to limit dose OPERABLE shutdown exposure to operators.

cooling heat exchanger.

3.6.2.2 Four Yes Two independent trains Two out of four cooling Two out of four cooling units SSCs are modeled consistent with Containment Cooling Containment of containment cooling units operating to: is required for success. the TS scope and trains can be System Cooling units shall be OPERABLE directly evaluated using the CRMP with one fan cooler to - Remove heat from the tool.

each train. containment atmosphere The success criteria in the PRA are following a loss of coolant consistent with the design basis accident (LOCA),

secondary system pipe rupture, or main steam line break (MSLB) inside containment.

W3F1-2021-0003 Page 11 of 21 Table E1-1 Proposed TS LCO SSCs Covered SSC in Functions Covered Design Success Criteria PRA Success Criteria Disposition Condition by TS LCO PRA by TS LCO Condition Condition Model

Maintain an acceptable containment pressure and temperature.

Limit off site radiation dose by reducing the pressure differential between containment atmosphere and the external environment.

3.7.1.5 MSIVs Yes Two MSIVs shall be Both MSIVs close Both MSIVs close SSCs are modeled consistent with Main Steam Isolation OPERABLE. the TS scope and so can be directly Valves (MSIV) Note: Separate evaluated using the CRMP tool. The Condition entry is success criteria in the PRA are allowed for each valve consistent with the design basis success criteria.

Failure of a single MSIV can only impact 1 of the 2 steam generators.

3.7.1.7 ADVs Yes Each Atmospheric These valves permit a 1 ADV per steam line to SSCs are modeled consistent with Atmospheric Dump Dump Valve (ADV) controlled release of open/remain open to reduce the TS scope and so can be directly Valves (ADV) shall be OPERABLE steam to provide the SG pressure. evaluated using the CRMP tool. The

a. With the automatic following: The PRA model includes success criteria in the PRA are actuation channel for - Reactor Coolant System local operation of ADVs as consistent with the design basis one ADV inoperable, cooling, and well as automatic function success criteria.

restore the - Main steam pressure of the valves.

inoperable ADV to control whenever the OPERABLE status MSIVs are closed or within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or when the condenser is reduce power to less not available for receiving than bypass steam.

or equal to 70% There is one ADV per RATED THERMAL steam line.

POWER within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

W3F1-2021-0003 Page 12 of 21 Table E1-1 Proposed TS LCO SSCs Covered SSC in Functions Covered Design Success Criteria PRA Success Criteria Disposition Condition by TS LCO PRA by TS LCO Condition Condition Model

b. With the automatic actuation channels for both ADVs inoperable, restore one ADV to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or reduce power to less than or equal to 70% RATED THERMAL POWER within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

c. With one ADV inoperable, for reasons other than above, restore the ADV to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

3.7.1.2 EFW pumps (2 Yes Three emergency At least one out of three One operating pump with SSCs are modeled consistent with Emergency electric motor feedwater (EFW) pumps (and associated operable flow path to the the TS scope and so can be directly Feedwater System driven, one pumps and two flow flow paths). steam generators. evaluated using the CRMP tool. The (EFW) steam turbine paths shall be success criteria in the PRA are driven), and OPERABLE consistent with the design basis injection flow success criteria paths 3.7.1.3 CSP Yes The condensate CSP aligned with CSP aligned with minimum SSCs are modeled consistent with Condensate Storage storage pool (CSP) minimum water volume water volume the TS scope and so can be Pool (CSP) shall be OPERABLE evaluated using the CRMP tool. The with: success criteria in the PRA are

W3F1-2021-0003 Page 13 of 21 Table E1-1 Proposed TS LCO SSCs Covered SSC in Functions Covered Design Success Criteria PRA Success Criteria Disposition Condition by TS LCO PRA by TS LCO Condition Condition Model

a. A minimum consistent with the design basis contained volume of at success criteria.

least 92% indicated level. The PRA does not explicitly model

b. A water temperature the impact of CSP temperature, but of greater than or equal this parameter can be addressed in to 55°F and less than the RICT program by assuming the or equal to 100°F. CSP is unavailable. Non-safety related backup/makeup to the CSP is also available and credited in the PRA model. Therefore, this LCO condition can be evaluated using the CRMP.

The cooling tower basin provides makeup to the CSP for EFW suction/operation. This would ensure the CSP below minimum level would not be a loss of function condition.

3.7.1.6 MFIVs Yes Each Main Feedwater Both MFIVs close. Both MFIVs close. SSCs are modeled consistent with Main Feedwater Isolation Valve (MFIV) the TS scope and so can be directly Isolation Valve shall be OPERABLE. MFIVs isolate the non- evaluated using the CRMP tool. The (MFIV) safety feedwater supply success criteria in the PRA are from safety related portion consistent with the design basis of the system. success criteria There is one MFIV per steam generator and they close on receipt of an MSIS 3.7.3 Two loops of Yes At least two One pump, one train. Same as design criteria. SSCs are modeled consistent with Component Cooling CCW with three independent Two independent trains Specific CCW success logic the TS scope and so can be Water (CCW) System CCW pumps. component cooling each with 100% capacity varies depending on PRA evaluated using the CRMP tool. The sequence. success criteria in the PRA are

W3F1-2021-0003 Page 14 of 21 Table E1-1 Proposed TS LCO SSCs Covered SSC in Functions Covered Design Success Criteria PRA Success Criteria Disposition Condition by TS LCO PRA by TS LCO Condition Condition Model Auxiliary Component Two water and associated to mitigate accident consistent with the design basis Cooling Water independent auxiliary conditions. success criteria.

System (ACCW) trains of ACCW component cooling ACCW design success is water trains shall be one out of two pumps is OPERABLE. required.

3.7.4 Two Yes Two Independent One of two trains (cooling One of two trains available SSCs are modeled consistent with Ultimate Heat Sink independent trains of ultimate heat towers and associated (same as design criteria) the TS scope and so can be (UHS) trains (1 wet sink (UHS) cooling water basin) evaluated using the CRMP tool. The and 1 dry towers shall PRA success criteria is more cooling tower be OPERABLE with complex than the design criteria with per train) each train consisting of a site specific thermal hydraulic a dry cooling tower evaluation used to determine the (DCT) and a wet various combinations of wet and dry mechanical draft tower fans needed for each accident cooling tower (WCT) condition. Exact, real time and its associated configurations can be evaluated water basin with: using the CRMP tool and PRA success criteria for RICT

a. A minimum water evaluations.

level in each wet tower basin of 97% (9.86 ft The action associated with basin MSL) level and temperature are not

b. An average basin included in the RICT program. Only water temperature of the fan operability is included.

less than or equal to 89°F.

c. fans as required by table 3.7-3 in Waterford 3 Technical Specifications 3.7.12 Two ECW Yes Two independent One of two ECW trains One of two ECW trains is SSCs are modeled consistent with trains essential services operating to provide required for PRA success. the TS scope and so can be Essential Chilled chilled water loops chilled water to safety evaluated using the CRMP tool Water (ECW) shall be OPERABLE. related air handling units.

W3F1-2021-0003 Page 15 of 21 Table E1-1 Proposed TS LCO SSCs Covered SSC in Functions Covered Design Success Criteria PRA Success Criteria Disposition Condition by TS LCO PRA by TS LCO Condition Condition Model With only one essential service chilled water loop OPERABLE, restore two loops to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

3.7.6.1 MCR Yes Two control room One of two One of two HVAC trains SSCs are modeled consistent with Control Room Air HVAC/Filtration emergency air filtration HVAC/filtration trains operating to maintain MCR the TS scope and so can be Conditioning System, system trains (S-8) shall be operable maintaining habitability. evaluated using the CRMP tool.

Control Room OPERABLE. habitable environment Emergency Air All the components in the MCR Filtration System HVAC/filtration system are not in the PRA model, but several Air Handling Units (AHUs), fans, and dampers are included allowing for a bounding evaluation using the CRMP.

3.7.6.3 MCR HVAC Yes Two independent One of two One of two HVAC trains SSCs are modeled consistent with Control Room Air control room air HVAC/filtration trains operating to maintain MCR the TS scope and so can be Conditioning System conditioning units shall operable maintaining habitability. evaluated using the CRMP tool.

be OPERABLE habitable environment All the components in the MCR HVAC/filtration system are not in the PRA model, but several AHUs, fans, and dampers are included allowing for bounding evaluation using the CRMP.

W3F1-2021-0003 Page 16 of 21 Table E1-1 Proposed TS LCO SSCs Covered SSC in Functions Covered Design Success Criteria PRA Success Criteria Disposition Condition by TS LCO PRA by TS LCO Condition Condition Model 3.8.1.1a Two sources of Yes As a minimum, the One qualified circuit One qualified circuit SSCs are modeled consistent with AC Sources - offsite AC following A.C. electrical between offsite between offsite the TS scope and so can be Operating transmission power sources shall be source/network and source/network and onsite evaluated using the CRMP tool.

OPERABLE: onsite Class 1E Class 1E distribution.

a. Two physically distribution.

independent circuits between the offsite transmission network and the onsite Class 1E distribution system, and 3.8.1.1b Diesel Yes As a minimum, the One of two EDGs capable One of two EDGs capable SSCs are modeled consistent with AC Sources - Diesel Generators, fuel following A.C. electrical of supplying one train of of supplying one train of the TS scope and so can be Generators storage and power sources shall be Class 1E electrical Class 1E electrical evaluated using the CRMP tool.

transfer OPERABLE: distribution distribution The success criteria in the PRA are components Two separate and consistent with the design basis independent diesel success criteria.

generators, each with:

1. Diesel oil feed tanks containing a minimum volume of 339 gallons of fuel, and

2. A separate diesel generator fuel oil storage tank, and

3. A separate fuel transfer pump.

3.8.2.1 Batteries and Yes All three divisions of One battery or one Each DC bus has two SSCs are modeled consistent with Battery DC power (batteries A, charger supplying DC chargers and a battery the TS scope and so can be DC Sources Chargers B, and AB) shall be power to 125V DC bus. providing power to it. Any evaluated using the CRMP tool.

(Batteries) -

operable. one of these three sources The success criteria in the PRA are Operating by itself can power the bus consistent with the design basis and its anticipated loads success criteria.

W3F1-2021-0003 Page 17 of 21 Table E1-1 Proposed TS LCO SSCs Covered SSC in Functions Covered Design Success Criteria PRA Success Criteria Disposition Condition by TS LCO PRA by TS LCO Condition Condition Model with minor exception (exceptions are explicitly modeled) 3.8.3.1a/b/c AC Busses Yes Engineered Safety One Train of AC power PRA success is the same SSCs are modeled consistent with Features (ESF) and supplying safety related as design success for the TS scope and so can be Onsite Power Static equipment power distribution evaluated using the CRMP tool.

Distribution Systems Uninterruptible Power The success criteria in the PRA are A/B/C Supply (SUPS) busses consistent with the design basis shall be energized in success criteria.

the specified manner.

a. Train A AC Busses

b. Train B AC Busses

c. Train AB AC 3.8.3.1d/e/f/g/h AC Yes Engineered Safety At least two ESF power PRA success is the same SSCs are modeled consistent with Uninterruptable Features (ESF) and division available. as design success for the TS scope and so can be Power Supply Static power distribution evaluated using the CRMP tool.

Onsite Power (SUPS) Uninterruptible Power The success criteria in the PRA are Distribution Systems Supply (SUPS) busses consistent with the design basis shall be energized. success criteria.

3.8.3.1j/k/l DC Busses Yes Engineered Safety At least one DC PRA success is the same SSCs are modeled consistent with Features (ESF) and distribution system as design success for the TS scope and so can be Onsite Power Static available to support safe power distribution evaluated using the CRMP tool.

Distribution Systems Uninterruptible Power shutdown. The success criteria in the PRA are Supply (SUPS) busses consistent with the design basis shall be energized. success criteria.

DC Busses connected to Battery Banks

W3F1-2021-0003 Page 18 of 21 Table E1-2: In Scope TS/LCO Conditions RICT Estimate Tech Spec LCO Condition RICT Estimate1, 2 3.3.1 (1) One of two manual reactor trip 30 Days Reactor Protective channels inoperable Instrumentation 3.3.2 One ESFAS (SIAS, CIAS, 30 Days The Engineered Safety CSAS, MSI, RAS, & EFAS)

Features Actuation module or measurement System (ESFAS) channel less than the total Instrumentation inoperable.

One ESFAS (SIAS, CIAS, 30 Days CSAS, MSI, RAS, & EFAS) module or measurement channel less than the minimum inoperable.

3.4.3.1 One group of pressurizer 24 Days Pressurizer heaters inoperable.

3.5.2 Emergency Core Cooling Systems One Low Pressure Safety 30 Days (ECCS) - ECCS Injection Train Inoperable Subsystems One ECCS subsystem 24 Days High Pressure Safety inoperable (other than one LPSI Injection train).

3.6.1.3 Containment Air Locks One containment air lock door 10 Days (Atmospheric and Dual) inoperable.

3.6.1.7 Containment One containment purge valve 10 Days Ventilation System inoperable.

3.6.3 Containment Isolation One containment isolation valve/ 30 Days Valves open system inoperable Containment Isolation One containment isolation valve/ 30 Days Valves closed system inoperable 3.6.2.1 Containment Spray One Containment Spray train 30 Days inoperable

W3F1-2021-0003 Page 19 of 21 3.6.2.2 Containment Cooling One Containment Cooling train 30 Days System inoperable 3.7.1.5 Main Steam Isolation One MSIV inoperable 15 Days Valves (MSIV) 3.7.1.7 Atmospheric Dump One required ADV inoperable 18 Days Valves (ADV) 3.7.1.2 Emergency Feedwater One steam supply to the turbine 18 Days System (EFW) driven pump inoperable One steam supply inoperable 4 Days and one motor driven pump inoperable (but 100% flow to a SG available)

One steam supply inoperable 12 Hours and both motor driven pumps inoperable (but 100% flow to a SG available)

EFW system inoperability (not 10 Days pump related) but 100% flow to either SG available 3.7.1.3 Condensate Storage CSP Inoperable due to low 6 Days Pool (CSP) water level 3.7.1.6 Main Feedwater One or more MFIVs inoperable 20 Days Isolation Valve (MFIV) (but open) 3.7.3 Component Cooling One Component Cooling train 16 Days Water (CCW) System inoperable Auxiliary Component One Auxiliary Component 24 Days Cooling Water System Cooling train inoperable (ACCW) 3.7.4 Ultimate Heat Sink One UHS train inoperable (due 20 Days (UHS) to fan operability) 3.7.12 Essential Chilled Water One essential chilled water loop 30 Days (ECW) inoperable

W3F1-2021-0003 Page 20 of 21 3.7.6.1 Control Room Air One control room emergency air 30 Days Conditioning System, filtration train inoperable (for Control Room equipment issues, not issues Emergency Air present with control room Filtration System envelope boundary) 3.7.6.3 Control Room Air One control room air 30 Days Conditioning System conditioning unit inoperable 3.8.1.1a AC Sources - One physically independent 20 Day Operating offsite transmission source (Class 1E) inoperable 3.8.1.1b AC Sources - Diesel One diesel generator of 3.8.1.1b 12 Days Generators inoperable 3.8.2.1 DC Sources (Batteries) One of the required battery 2 Days

- Operating banks inoperable With one of the required full 2 Days capacity chargers inoperable 3.8.3.1 One of the required divisions of 3 Days A.C. ESE busses not fully Onsite Power Distribution Systems energized One A.C. SUPS bus either not 12 Days energized from its associated inverter, or with the inverter not connected to its associated D.C.

bus One D.C. bus not connected to its associated battery bank 1 Day

1. Estimated RICTs are listed. Following program implementation, the actual RICT values will be calculated on a plant-specific basis, using the actual plant configuration and the current revision of the PRA model representing the as-built, as-operated condition of the plant, as required by NEI 06-09-A, Revision 0-A and the NRC safety evaluation, and may differ from the RICTs presented. RICTs evaluations utilize the internal events, internal flood, and internal fire PRA model calculations with seismic CDF and LERF penalties applied.

2. RICTs calculated to be greater than 30 days are capped at 30 days based on NEI 06 A, Revision 0-A.

Enclosure 2 W3F1-2021-0003 Information Supporting Consistency with Regulatory Guide 1.200, Revision 2

W3F1-2021-0003 Page 1 of 33 Information Supporting Consistency with Regulatory Guide 1.200, Revision 2 The Waterford 3 Internal Events PRA model (including internal flood) was peer reviewed in 2009 by the PWR Owners Group (PWROG). The review was conducted in May 2009 and the report was issued in August 2009 (Reference 1). The Internal Events PRA technical adequacy (including the 2009 peer review and follow up self-assessment results) has previously been reviewed by the NRC in previous LAR efforts associated with the NFPA 805 and TSTF 425 LARs, (June & July 2016, (Reference 2) (Reference 3))

The 2009 Peer Review was completed using the previous revisions of industry guidance.

Revision 1 of Regulatory Guide 1.200 was applied, and the review was conducted using the 2005 issued PRA standard (ASME RA-Sb-2005 - Reference 4). In development of the NFPA 805 LAR, a gap assessment was performed due to revisions to the guidance and the PRA standard. The goal of this assessment was to both identify the differences in the two documents, and to identify the impact the differences potentially have on existing Waterford 3 peer review findings. The results of this gap assessment are documented in PSA-WF3-08-01 (Reference 6) and following it all assessments were made relative to the 2009 PRA Standard (ASME/ANS RA-Sa-2009 - Reference 5) and Revision 2 of Regulatory Guide 1.200 (Reference 7).

The Fire PRA model was subject to a self-assessment and a full-scope peer review was completed in 2011. The review was in November 2010 and the resulting report was issued in February 2011 (Reference 8). Following a 2012 NRC Audit of the Waterford 3 NFPA 805 LAR and supporting documents, Waterford 3 revised the Fire PRA including several method changes. Waterford 3 had two focused scope peer reviews to ensure proper evaluation of the revised methods. One was in September 2012 (Reference 26) and another one was conducted in May 2013 (Reference 27).

The results of those peer reviews (for internal events and for the fire model) were the basis for the NFPA 805 LAR and Safety Evaluation and the TSTF 425 (surveillance frequency control program) LAR and Safety Evaluation.

In the time following those reviews, the Waterford 3 PRA models have been through several updates (with some technical upgrades) as well as peer reviews and formal Facts and Observation Close-out reviews. Findings for both the full power Internal Events and Fire PRA models (as of October 2017) were reviewed and closed using the process documented in Appendix X to NEI 05-04, NEI 07-12 and NEI 12-13, Close-out of Facts and Observations (F&Os) as accepted by NRC in the staff memorandum dated May 3, 2017 (ML17079A427). The results of this review have been documented (Reference 9) and are available for NRC audit.

The closure review concluded that the resolution of one finding resulted in application of a modeling upgrade. This one PRA upgrade as defined by the ASME PRA Standard RA-Sa-2009 (Reference 5) has occurred to the Internal Events PRA model since conduct of the PWROG peer review in 2009. That upgrade is associated with cooling tower success criteria. It was reviewed during the F&O closure review in October 2017 (as an imbedded focused peer review).

Following the closure review, both the internal events model and fire PRA model were revised to address findings that were not closed during the closure review (in addition to a routine model

W3F1-2021-0003 Page 2 of 33 update). During this update, the LERF model was revised with model upgrades to resolve issues and more thoroughly meet the PRA Standard. The LERF model was subject to a peer review in August 2019 as a result of the upgraded methods used. No unreviewed PRA upgrades (except those noted for Flood and LERF) were included in the internal events and internal fire PRA revision efforts.

The closure review conducted excluded open findings associated with the internal flood PRA model. The model had not been thoroughly updated since the original findings. The Waterford 3 internal flood PRA model was revised in 2019. The flood PRA model update also included methodology upgrades. A peer review was necessary due to the method upgrades and the time gap between the previous full scope flood model update (nearly ten years). The updated flood model was subject to a peer review in August 2019 for all PRA Standard elements relevant to internal flooding PRA models. Following the peer review, the flood PRA was updated in 2020 to address the peer review findings.

Table E2-1 lists the peer review efforts conducted in the past several years including dates and descriptions.

W3F1-2021-0003 Page 3 of 33 Table E2 Waterford 3 PRA Peer Reviews Review Description Review Document RG 1.200 PRA Peer Review Against the ASME PRA Standard LTR-RAM-II-09-39 Requirements For The Waterford Steam Electric Station, Unit 3 (Reference 1)

Probabilistic Risk Assessment (Westinghouse Owners Group August 2009)

Fire PRA Peer Review of Waterford Steam Electric Station Unit 3 Fire LTR-RAM-II-11-003 Probabilistic Risk Assessment Against the Fire PRA Standard (Reference 8)

Supporting Requirements from Section 4 of the ASME/ANS Standard (Westinghouse Owners Group February 2011)

Focused Scope Fire PRA Peer Review of Waterford Steam Electric PSA-WF3-08-03 Station Unit 3 Fire Probabilistic Risk Assessment Against the Fire PRA (Reference 26)

Standard Supporting Requirements from Section 4 of the ASME/ANS Standard (Robert Brady -URS Corporation - September 2012) 2nd Focused Scope Fire PRA Peer Review of Waterford Steam PSA-WF3-08-04 Electric Station Unit 3 Fire Probabilistic Risk Assessment Against the (Reference 27)

Fire PRA Standard Supporting Requirements from Section 4 of the ASME/ANS Standard (Robert Brady -URS Corporation - May 2013)

Waterford 3 Finding Level F&O Independent Technical Review PSA-WF3-08-02 (October 2017) (Reference 9)

Waterford 3 Internal Flood/LERF Focused Scope Peer Review PSA-WF3-08-05 (August 2019) (Reference 28)

Note - All of the Findings from the first four entries in the table were evaluated in the F&O Closure Review conducted in 2017. The only exception to this is the internal flooding and LERF related findings. All PRA Standard Elements for internal flooding and LERF were evaluated in the August 2019 flood/LERF peer review that was conducted on the updated flood model and LERF model.

W3F1-2021-0003 Page 4 of 33 Tables E2-2 through E2-4 provide a summary of the remaining findings and open items, including:

- Table E2-2 below lists the open Findings associated with the internal events PRA model and provides a disposition associated with the risk-informed applications for each open technical issue. This table includes open issues remaining from the 2017 closure review as well as the 2019 LERF related peer review.

- Table E2-3 below lists the open Findings associated with the August 2019 internal flood PRA model peer review.

- Table E2-4 below lists the open Findings associated with the internal Fire PRA model.

This table address the finding that remain open following 2017 F&O Closure Review.

All three tables provide a disposition for each entry with regard to a Risk-Informed Completion Time (RICT) program (and LAR).

The content of Tables E2-1, E2-2, E2-3, and E2-4 is contained in Waterford 3 PRA report PSA-WF3-08-06 PRA Technical Adequacy to Support Risk-Informed Applications, (Reference 10).

This includes listed references.

PRA Credit for FLEX The Waterford 3 PRA model does credit FLEX equipment and strategies. The credit for FLEX-related equipment is limited to specific extended loss of offsite power scenarios and is limited to permanently installed FLEX equipment or portable equipment stored in the RAB. The model changes to incorporate FLEX equipment and strategies referenced site procedures and is a direct representation of the as-built as-operated plant. The specific changes to add the FLEX equipment and strategies were not judged to be PRA upgrades, as existing modeling methods and techniques were used to update the model (no new or unique methods were applied).

The Waterford 3 PRA model credits a FLEX diesel generator to provide power to battery chargers (given an extended loss of offsite power) and to other FLEX equipment. This diesel unit is installed in the Reactor Auxiliary Building (RAB). Use of this equipment and actions necessary to start and align it are included in site procedures for loss of offsite power, and all necessary equipment (cables, panels, keys, etc.) is pre-staged. Existing model failure data type codes were used for diesel generator and circuit breaker failure data. As the equipment is permanently installed and procedurally controlled, generic failure data was judged applicable.

The FLEX Core Cooling Pump (FCCP) is a permanently installed pump that can be aligned to either provide Feedwater to the Steam Generators (SGs), makeup to the RCS, or backup cooling to the Spent Fuel Pool. The pump can be powered from a charging pump breaker, supplied by emergency power, or powered from the FLEX Diesel. Only alternate Feedwater to the Steam Generators is credited in the PRA, as this is the only applicable FLEX capability for a PRA modeled hazard. As with the FLEX diesel, all equipment and actions necessary to align and operate the pump for this function are driven by site procedures. Existing model failure data type codes were used for the FCCP. As the equipment is permanently installed and procedurally controlled, generic failure data was judged applicable.

The human actions added to the PRA model for FLEX deployment followed the same Human Error Probability (HEP) development methods as all other modeled actions. Credited actions are all procedure driven actions. Peer reviewed HEP methodology was applied to the added actions. Live timed field trials were used to support timing inputs for HEP development. The credited operator actions are procedure driven actions and are similar to other operator actions

W3F1-2021-0003 Page 5 of 33 evaluated using approaches consistent with the ASME/ANS PRA Standard as endorsed by RG 1.200, Revision 2.

The modeling of the credited FLEX equipment and actions:

- Does not represent any new methods.

- Does not change the scope of the model given that the equipment, dependencies, and type of accident sequences remain the same.

- Does not represent a change in capability of the PRA model given the original and updated models can both evaluate the risk associated with loss-of-offsite power and station blackout.

The changes implemented for the incorporation of the FLEX modifications were within the framework of the existing peer reviewed PRA model structure.

W3F1-2021-0003 Page 6 of 33 Table E2 Open Internal Events Peer Review Findings Assessed During F&O Closure Review Supporting Capability Finding Requirement Category Description Disposition for RICT Program Number (s) (CC)

AS-B3-01 AS-B3 Not Met The AS report (PRA-W3-01-001S01 Revision 1) This was graded by the 2017 F&O closure review team as includes discussion of the phenomenological partially resolved (with open documentation issues).

impacts of heating of the containment sump water (failure of HPSI recirculation due to loss of required WCAP-16679-P - Accident Sequence Phenomena was NPSH and pump cavitation) and large containment reviewed to determine if any additional phenomena other rupture (loss of safety injection due to the rapid needed to be addressed in the current Waterford 3 AS depressurization, flashing of hot water in the sump, analysis. All other phenomena have been addressed in the and loss of net positive suction head to the HPSI accident sequence and the system analyses, as necessary.

pumps) that can occur due to inadequate The effects of steam line and feed line breaks are evaluated containment heat removal. However, some events in the initiating event document.

such as steam line breaks and feedwater line breaks can result in harsh environments (especially Waterford 3 completed a review of the phenomenological steam and high temperature) where mitigating considerations in the AS report immediately following the equipment are located. original peer review (in 2010). That analysis and the results of it were not included in (or referenced) in the model Phenomenological impacts must be considered in Revision 5 AS report reviewed during the closure review.

order to ensure risk results are not underestimated. The considerations are included in the model - the updated documentation was noted to be insufficient. This disconnect Consider and document the phenomenological in documentation was the basis for grading this finding as conditions from the entire range of initiating events, partially resolved with open documentation issues.

especially high energy line break.

The Revision 6 update has a more thorough documentation of such phenomena and their treatment. The Revision 6 Accident Sequence and Success Criteria documentation contain the necessary details to satisfy AS-B3 (though a formal closure review has not been completed).

This Finding has no impact on quantified results and no impact on Risk Informed Completion Time application.

HR-B1-01 HR-B1 CC-I There is no pre-initiator identified for CCW, This was graded by the 2017 F&O closure review team as because of the CCW is a running system. partially resolved (with open documentation issues).

However, the CCW system may support the safety related standby system. The path of the CCW to This F&O has been resolved during the 2019 Revision 6 support this system may be failed due to pre- model update with a thorough review of pre-initiators and

W3F1-2021-0003 Page 7 of 33 Table E2 Open Internal Events Peer Review Findings Assessed During F&O Closure Review Supporting Capability Finding Requirement Category Description Disposition for RICT Program Number (s) (CC) initiator HFE. The potential for this type of pre- evaluation of the exact consequence of each failure.

initiator HFE needs to be re-evaluated and Restoration errors of CCW to a standby system are included documented. in the restoration of the associated standby system. For instance, CCW to the Containment Spray pumps is included in the Containment Spray restoration logic (not the CCW logic). Therefore, these restoration errors have been identified and evaluated in the current model (and correctly modeled to the right system consequence).

The closure review team judged that support system related pre-initiators are properly accounted for and modeled but the system notebooks were not explicit enough in the detailed explanation.

The Revision 6 PRA model documentation contains the necessary details and more thoroughly documents pre-initiator modeling. This has no impact on application of a RICT program. This is a documentation issue only and will have no impact on quantified results.

HR-F2-01 HR-F2 CC-I The cue of each HFE is not clearly addressed. The This was graded by the 2017 closure review team as timing of each cue should be included in the partially resolved.

description of the HFE time windows and included in the quantification of the HEPs. For HEP events, the cues (i.e., annunciators, EOP/AOP entry conditions) are explicitly discussed in each operator action in the model and are documented in the operator interview sheets. All of this is documented in the updated PRA HRA analysis PSA-WF3-01-HR (Reference 24).

The closure review team assessed that the HEP development spreadsheets (which are used or referenced in the HR analysis) lacked thorough detail and references associated with operator timing and cues. The timings used were/are reasonable and references/operator interviews could be found, but a direct tie in the development

W3F1-2021-0003 Page 8 of 33 Table E2 Open Internal Events Peer Review Findings Assessed During F&O Closure Review Supporting Capability Finding Requirement Category Description Disposition for RICT Program Number (s) (CC) documentation was not always explicitly provided. The spreadsheets used to calculate the HEP values lacked basis for all timings used in development. This limitation resulted in the finding remaining open.

The PRA Revision 6 update included application of the EPRI HRA calculator. The information associated with this finding is now more thoroughly documented for each modeled action.

This Finding has no impact on application of a RICT program. This is a documentation issue only and will have no impact on quantified results.

HR-G4-01 HR-G4 Not Met As seen in the HRA spreadsheet (hfe_cp.xls), the This was graded by the 2017 closure review team as time available to complete actions is based on a partially resolved (with open documentation).

range of references including plant-specific calculations. However, in some cases unjustified For HEP events, the cues (i.e., annunciators, EOP/AOP and/or inaccurate assumptions were used as a entry conditions) are explicitly discussed in each operator basis. The event timelines in the HRA action in the model and are documented in the operator spreadsheets also do not consistently identify the interview sheets. All of this is documented in the updated specific point in time relevant indications are PRA HRA analysis, PSA-WF3-01-HR (Reference 24).

received.

Inaccurate HFE time windows may result in under- The closure review team assessed that the HEP or over-estimation of HEPs. Ensure the time development spreadsheets (which are used/referenced in windows are justified based on realistic analyses or the HR analysis) lacked thorough detail and references simulation and identify in the HEP worksheets the associated with operator timing and cues (and timing of point in time relevant indications are received. specific cues). The timings used were/are reasonable and references/operator interviews could be found, but a direct tie in the development documentation was not always explicitly provided.

The 2019 PRA Revision 6 update included application of the EPRI HRA calculator. The information associated with this

W3F1-2021-0003 Page 9 of 33 Table E2 Open Internal Events Peer Review Findings Assessed During F&O Closure Review Supporting Capability Finding Requirement Category Description Disposition for RICT Program Number (s) (CC) finding is now more thoroughly documented for each modeled action.

This Finding has no impact on application of a RICT program. This is a documentation issue only and will have no impact on quantified results.

SC-C1-02 SC-C1 Not Met Throughout the document there are a number of This was graded by the 2017 closure review team as assumptions and statements made that directly partially resolved.

impact the success criteria but do not have any references identified to justify their bases. Querying The review team concluded that - while the revised report the PRA group determined that most of the did contain some improvement in providing statements were based on valid references, but references/bases for assumptions and SC treatments - it is they were not identified in the success criteria still not sufficient to meet CC-II of the Standard. This documentation. The references need to be includes some missing references, duplicate entries, and specifically identified and included. references to old/superseded documents.

The PRA model Revision 6 update included an updated SC report with a detailed review of references to ensure a valid basis was provided for documented success criteria.

This Finding has no impact on application of a RICT program. This is a documentation issue only and will have no impact on quantified results.

SY-A12b- SY-A12 Not Met Need to use the exclusion criteria in SY-A14 to This was graded by the 2017 closure review team as 01 (SY-13) justify excluding flow diversion pathways. Using partially resolved.

the criteria 2 normally closed valves should be SY-B14 easily justified using criteria SY-A14(a). The The 1/3 exclusion criteria is noted as standard treatment for (SY-15) criteria for excluding based on a 1 to 3 ratio systems, unless otherwise noted. However, it is not between the primary piping and the potential uniformly applied to all systems. Systems with different diversion piping needs to be backed up by modeling treatment and the basis for the different treatment pressure differentials. This exclusion criteria is are noted in the documentation. Additional flow diversion valid if the system pressures between the primary failures were credited following this finding for HPSI and and potential diversion piping is the same or LPSI based on meeting 1/3 criteria but having high pressure

W3F1-2021-0003 Page 10 of 33 Table E2 Open Internal Events Peer Review Findings Assessed During F&O Closure Review Supporting Capability Finding Requirement Category Description Disposition for RICT Program Number (s) (CC) similar. If the pressure differential is high, further differential, and CCW Makeup for having a finite/limited analysis is required to justify exclusion. volume. The CCW diversion would not fail system function Overall, the assumptions used to exclude specific but over time would reduce inventory below functional level types of failures needs to be reevaluated and before 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

justification provided on how the exclusion criteria is met. This Finding has no impact on application of a RICT program.

LE2-3(1) LE-E1* Met The LERF model includes events reflecting Continued operation of CCS and CSS under harsh conditional probability of the Component Cooling conditions/environment has an insignificant impact on LE-C9 Not Met System (CCS) and the Containment Spray (CSS) overall LERF results.

due to a harsh severe-accident environment.

Section 8.1.9 of the LERF Notebook (PSA-WF3- Vessel Rupture accounts for over 90% of total LERF.

01-LE, Rev. 2) identifies events P_CCSFAILS and ISLOCA is the second largest contributor accounting for

Finding P_CSSFAILS and refers to the Level 2 analysis for over 50% of the non-Vessel Rupture LERF total. CCS and Reference the development of the associated conditional CSS have no impact on these two LERF contributors. The this SR but it probabilities. The Level 2 report (PSA-WF3-01-L2- values applied to events P_CCSFAILS and P_CSSFAILS do is noted in the 01, Rev. 0), in turn, refers to the Waterford 3 not significantly impact LERF results. The revised LERF Report as Individual Plant Examination (IPE, Waterford 3 report shows a sensitivity case with the applied CSS and Met Probabilistic Risk Assessment Individual Plant CCS values increased by a factor of 10. This change Examination Submittal, August 1992) as the results in a less than 1% increase in LERF.

source. Table 4.6-2 of the IPE lists values for these events but provides no further analysis or Solving the fault tree for the loss of train A and B justification. containment spray system gate generates a 1.7E-03 result.

The applied P_CSSFAIL is 5E-03. The LERF input for loss of CSS (originally from the IPEE) is nearly a factor of 3 The SR requires that justification be provided for higher than the detailed fault tree logic. Similarly, the CCS continued operation of systems under adverse value is nearly a factor of 10 higher than calculated in the (severe-accident) environments. Although the fault tree.

potential for an elevated probability of failure has been incorporated for the CCS and CSS, no The systems in question are both designed to operate in a justification for the probabilities selected is harsh environment, and most of the active components for provided.

the system are outside containment (and wont be impacted by a harsh containment environment).

Develop a justification for the continued operation of CSS and CCS under adverse conditions.

W3F1-2021-0003 Page 11 of 33 Table E2 Open Internal Events Peer Review Findings Assessed During F&O Closure Review Supporting Capability Finding Requirement Category Description Disposition for RICT Program Number (s) (CC)

The model revision 6 LERF report (PSA-WF3-01-LE -

Reference 25) has been updated to provide additional basis for the values used. The applied values are conservative and have no significant impact on LERF results. This finding has no impact on quantified results and will not impact risk informed applications.

LE2-4(1) LE-E2 Not Met The treatment of accident-progression phenomena The LERF analysis has been updated to include additional relies almost entirely on predictions from MAAP discussion relevant to the limitations of MAAP 4.0.6, as well calculations. While this is appropriate for much of as to include additional sensitivities regarding in-vessel melt LE-F2* Cat-I/II/III the LERF analysis, the reliance on MAAP for retention issues (and several specific sensitivity cases crediting that core debris remains in the reactor related to the MAAP 4.0.6 limitation associated with this pressure vessel for a preponderance of scenarios phenomena). The updated documentation provides more is a potentially significant non-conservatism. It is detail on in-vessel retention and addresses the uncertainty

Finding difficult to determine what the impact on LERF of the cases and computer codes used in the evaluation.

References would be if these scenarios were permitted to lead this SR but it to relatively early breaching of the vessel. The following is an excerpt from the revised LERF document:

is noted in the Report as The reasonableness check did not address the Relative to the ability of MAAP 4.06 to model in-vessel Met appropriateness of MAAP 4.0.6 for certain accident retention, it is important to be aware of a Trouble Report phenomena, including in-vessel melt retention. posted for MAAP 4.08 and earlier releases. The error report indicates the following:

Investigate the potential for in-vessel retention in more detail, based on available literature and/or a

Reference:

TR 810 Ex-vessel cooling code error (Mod.

tool that is suitable for assessing this potential.

Package 833)

Adjust the LERF analyses to reflect more realistic As indicated in the reference above, there is a code error in treatment of in-vessel melt retention.

MAAP4.06 related to ex-vessel cooling:

Ensure that the treatment of in-vessel melt If the reactor cavity is flooded, there is substantial retention is considered during future checks of the debris in the lower head, and ex-vessel cooling is reasonableness of LERF contributors.

enabled, it is possible to melt through the reactor wall heat sink (i.e. mrvn(6,1) through mrvn(6,5) become 0), and the reactor vessel has not failed. If this occurs, the tabular output file will show primary

W3F1-2021-0003 Page 12 of 33 Table E2 Open Internal Events Peer Review Findings Assessed During F&O Closure Review Supporting Capability Finding Requirement Category Description Disposition for RICT Program Number (s) (CC) system energy imbalance, and RV failure will not have occurred for the MAAP sequence.

To address this error report, a sample of the existing MAAP cases indicating that vessel failure was prevented were reviewed to better understand if this error is occurring for those runs. As stated in the TR description above, the two vessel wall heat sinks should be examined to determine if the vessel had indeed failed, but that indication of breach was not acknowledged. In order to make this determination, those cases may need to be rerun to include plot variables associated with the heat sink mass as identified in the TR.

The Modification package includes a work-around that can be added to the input file to correct this error. For the selected sample cases, the mass of the heat sinks did not become zero as discussed above and therefore the error was not a factor in the determination of vessel breach.

The LERF documentation has been updated to more thoroughly address the issues identified in the Finding. This finding has no impact on quantified results and will not impact risk informed applications.

LE4-4(1) LE-G5 Not Met Section 9.6 of PSA-WF3-01-LE, Rev. 2 lists the The LERF documentation has been updated to address this types of PRA limitations to consider from the Finding and detail the impact of LERF assumptions on PRA ASME/ANS PRA Standard but provides no specific applications. Assumptions related to model development discussion of the actual limitations of the Waterford can and do impact results. However, the overall methods 3 LERF model nor how they may impact risk- and process used limit those impacts. Consensus modeling informed applications. Attachment A identifies the approaches are used in PRA model development. The assumptions in the Waterford 3 LERF model, but LERF model is also peer reviewed against the PRA does not identify impacts on applications from the Standard. Sensitivity cases for key sources of uncertainty treatment of the assumptions. have been developed to ensure the impact of assumptions and modeling decisions are known and documented. This Document the model limitations and potential thorough, peer reviewed, state-of-the-art approach to LERF impacts on PRA applications for Waterford 3 LERF modeling helps ensure the model and results maintain the model.

W3F1-2021-0003 Page 13 of 33 Table E2 Open Internal Events Peer Review Findings Assessed During F&O Closure Review Supporting Capability Finding Requirement Category Description Disposition for RICT Program Number (s) (CC) technical adequacy requirements to support risk-informed applications.

The LERF documentation has been updated to more thoroughly address the issues identified in the Finding. This finding has no impact on quantified results and will not impact risk informed applications.

LE4-5(1) LE-F2* Met Cat An analysis using the MAAP 4.0.6 computer code The LERF documentation has been updated to more I/II/III based on realistic inputs in PSA-WF3-01-LE, Rev. thoroughly address the appropriateness of MAAP 4.0.6 as it LE-D6* 2, calculates the plant specific conditions for the applies to TISGTR.

Met Cat III TISGTR accident sequences. The TISGTR analysis relies solely on MAAP 4.0.6 analyses to The updated reasonableness check is based on:

Finding make the determination of whether an accident EPRI Perspective on Thermally-Induced Steam Generator References progression sequence is classified as TISGTR and Tube Rupture Issues, NRC document ID: ML071340053, this SR but it does not apply split fractions based on industry Marc Kenton is noted in the guidance. The capabilities of the MAAP 4.0.6 code Report as should be validated for the appropriate use for this It is commonly believed that the TISGTR has a minor impact Met determination. Plant procedures are incorporated on the PRA results. Hot leg creep rupture has been found to into the operator actions governing the TISGTR either occur prior to tube rupture or immediately after. In response. both instances, the dominant flow path will be from the reactor vessel to containment, with a relatively small flow The reasonableness check did not address the through the failed SG tube.

appropriateness of MAAP 4.0.6 for certain accident phenomena, including TISGTR. Based on the reference above, the MAAP code can accurately predict the peak temperature in the SG tubes as well as the hot leg.

The LERF report was revised to update the reasonableness argument (including addition of the noted reference) to address this Finding. This Finding has no impact on quantifies LERF results or application of a RICT program.

Note - The two SRs listed in the Finding are both noted in the report as met for Capability Category II or III. Even

W3F1-2021-0003 Page 14 of 33 Table E2 Open Internal Events Peer Review Findings Assessed During F&O Closure Review Supporting Capability Finding Requirement Category Description Disposition for RICT Program Number (s) (CC) though the finding was issued the grading of the finding is that the analysis is sufficient for PRA applications as assessed (with a recommendation for improvement - which was completed).

Note 1 - These findings are the result of a peer review of the LERF model conducted in August 2019. The other listed findings are the ones that remained open following an F&O Closure Review in 2017.

W3F1-2021-0003 Page 15 of 33 Table E2 Internal Flooding Open Peer Review Findings Supporting Capability Finding Requirement( Category Description Disposition for RICT Program Number s) (CC)

FL1-3 IFSN-A15 Met CC-I/II/III Table B.3 of PSA-WF3-01-IF-WD lists piping The pipe segments excluded from the analysis have segments (flood sources) that require further been identified and evaluated. The updated Flood PRA analysis as to whether they could be screened. A documents account for the subject segments. Many Waterford 3 PRA Model Change Request (MCR screened but several resulted in new or updated

W3-6480) was written to track this issue. scenarios.

Complete the analysis of these piping segments to The updated flood analysis properly treats the determine whether they can be screened per the segments identified in this Finding. This issue has no criteria given in this SR. impact on quantified results and will not impact risk-informed applications.

Though a Finding was issued, the peer review graded the SR as Met CC-I/II/III.

FL-1-4 IFQU-B3* Met In Table A-2 of PSA-WF3-01-IF-SOU, item IF-C-9 A detailed engineering evaluation of the subject door dealing with the structural analysis of doors may was completed by Waterford 3/Entergy structural

Finding have a significant impact on internal flood CDF. engineers. The Flood PRA scenarios were updated References The configuration of doors within room 211 based on this more thorough evaluation of flood this SR but it is (vestibule area) may be such that the door leading propagation from room 211. The Flood PRA was noted in the to the outside environs may preferentially fail first, updated to explicitly address this Finding. Resolution of Report as Met which could have a significant impact on the this issue did result in a significant change in flood calculation of internal flood CDF (~50%). results. With the issue resolved, this Finding has no further impact on quantified results and will not impact risk-informed applications.

This Finding was issued even though IFQU-B3 was graded as Met Cat 1-3 in the peer review report.

FL 1-6 IFEV-A5* Met CC-I/II/III The frequency for event %FLD-TB_ALL used an The Flood model and documentation has been older data set (NUREG/CR-5750) that was updated. Updated initiating event frequencies have

Finding inconsistent with the data used throughout the rest been developed (in part to address Finding IFEV-A6 References of the plant. The EPRI reference that was primarily listed below in this table). Additional details on the this SR but it is used for other event frequencies outside the development of the Turbine Building frequency (%FLD-Turbine Building can also be used to develop plant- TB_ALL) have been added. Reasonableness checks of

W3F1-2021-0003 Page 16 of 33 Table E2 Internal Flooding Open Peer Review Findings Supporting Capability Finding Requirement( Category Description Disposition for RICT Program Number s) (CC) noted in the level flood frequencies based on generic industry resulting frequencies were completed to ensure Report as Met operating experience. compliance with Standard SR IE-C12 (and IFEV-A5).

This SR refers to the requirements in Section 2-2.1 This Finding has been addressed in the updated Flood of the Standard, which does involve SR IE-C12 PRA model and documentation. This issue has no that requires a comparison of results of the impact on quantified results and will not impact risk-initiating event analysis with generic data sources informed applications to provide a reasonableness check of the results.

The event frequency used from NUREG/CR-5750 This Finding was issued even though IFEV-A5 was is an older data source that was not compared with graded as Met Cat 1-3 in the peer review report.

the more recent data set used for the rest of the plant.

Compare the event frequency used for the Turbine Building floods with the plant-level data found in Section 8 of EPRI 3002012997, which covers a larger time period than the period used in NUREG/CR-5750. This will confirm the "reasonableness" of the selected event frequency for this initiating event and whether it should be deemed conservative as stated in Section 2.2.146.

FL3-2 IFPP-B1* Met There are discrepancies between what was The Flood PRA documentation has been updated to modeled in PSA-WF3-01-IF-AS and the information resolve discrepancies between the Accident Sequence in Table 2 of PSA-WF3-01-IF-FA. For example, and Flood Area Development documents. The revised

Finding RAB-21-211 lists Door D11 to RAB21-Q as not documents and tables contain consistent technical References having sufficient accumulation of water to fail, and information.

this SR but it is PRA equipment damage protected by D11 is listed noted in the as 'N/A'. The Flood PRA was updated to explicitly address this Report as Met Finding. With the issue resolved, this Finding has no Tables 1 and 2 of PSA-WF3-01-IF-FA should be impact on quantified results and will not impact risk-updated to be consistent with the accident informed applications.

sequence analysis described in Section 4 of PSA-WF3-01-IF-AS.

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This Finding was issued even though IFPP-B1 was graded as Met Cat 1-3 in the peer review report.

FL3-4 IFEV-A6 Cat I EPRI TR-3002012997 R4 Section 4.5 requires EPRI has rescinded the TR-3002012997 R4 guidance consideration of age correction for significant document. The flood PRA initiating events were scenarios. Aging factors from Table 4-19 were not updated to use EPRI TR-3002000079 R3 (which is still applied. It could be argued that age adjustment endorsed by EPRI). Following the TR-3002000079 R3 using EPRI TR-3002000079 R3 would not be guidance does not require age correction due to the applicable to Waterford 3 for another 2 to 3 years. age of the Waterford 3 site and piping systems (PRA-W3-01-IF-QU- Reference 23).

This SR requires consideration of material condition of fluid systems. EPRI TR-3002000079 The Flood PRA was updated to explicitly address this R3 does not require age correction for FP piping Finding. With the issue resolved, this Finding has no that has been in service less than 40 years. impact on quantified results and will not impact risk-However, EPRI TR-3002012997 R4 Table 4-19 informed applications.

provides age correction factors ranging from 10 to 50 years of service. Frequencies calculated in PRA-WF3-IF-QU were based on nominal values from EPRI TR-3002012997 R4 without age correction.

FL3-5 IFEV-A7 Not Met Section 2.1.1 of PRA-WF3-01-IF-IE includes an The flood documentation has been updated to correct invalid input to the equation used to calculate the the issue identified (including correction of the frequency of a maintenance induced flood event. erroneous value) in this Finding. Human induced flood The use of a maintenance unavailability value (or events are considered in the Waterford 3 flood PRA probability) was erroneously used as a frequency analysis. Human induced flood events have been in the justification for screening maintenance screened, and a valid basis for the screening is induced flooding events from the analysis. included in the updated documentation.

This SR requires consideration of human-induced The Flood PRA was updated to explicitly address this floods from maintenance. While consideration of Finding. With the issue resolved, this Finding has no maintenance induced flooding, events was impact on quantified results and will not impact risk-provided in Section 2.1.1 of PRA-WF3-01-IF-IE, the informed applications.

basis for screening these events is invalid.

Correct the qualitative argument to justify screening human induced flood events based on a

W3F1-2021-0003 Page 18 of 33 Table E2 Internal Flooding Open Peer Review Findings Supporting Capability Finding Requirement( Category Description Disposition for RICT Program Number s) (CC) review of generic industry data or apply the methodology given in Section 8 of EPRI 3002012997 R4.

FL3-7 IFQU-A10 Not Met Overlaying the internal flood logic and target sets The Flood PRA was updated following the peer review.

onto the LERF sequences produced erroneous The update included changes to key scenarios and cutsets relative to the corresponding CDF cutsets. initiating event frequencies. The update required For example, the top LERF cutset from flooding updated quantification of CDF and LERF. During the scenario %FLD-RAB_21-211FP-L includes AB update, an existing modeling error (not explicitly related Switchgear alignment events 'B_TO_AB' and to flood) was found and corrected. CDF and LERF

'HPIABISSTBY' while the corresponding CDF cutsets were reviewed for reasonableness and to cutset is more consistent with loss of equipment in ensure mapping/overlaying the CDF results onto LERF the AB and A Switchgear Rooms. was properly completed.

Justify the validity of the anomalous LERF cutsets Updated Flood LERF results were reviewed to check generated by internal flood initiators or modify the for erroneous results. With the issue resolved, this LERF analysis as necessary to account for the Finding has no impact on quantified results and will not unique flood-induced scenarios in accordance with impact risk-informed applications.

the applicable requirements described in 2-2.8.

Note - All Flood PRA findings are the result of an August 2019 Flood PRA peer review.

W3F1-2021-0003 Page 19 of 33 Table E2 Fire PRA Open Peer Review Findings Assessed During F&O Closure Review)

Supporting Capability Finding Requirement Category Description Disposition for RICT Program Number (s) (CC)

CS-A3-01 CS-A3 Met The component and cable selection report, Rev. 0, The 2017 F&O Closure review judged this finding as reviewed looking for effects of interlocks and Open ES-B4 Not Met permissive. There is discussion of the VCT in the MSO discussion, but there does not appear to be The Fire PRA model and documentation was updated in any other interlock or permissive discussion. 2019 to incorporate the internal event (R6) model update These items would be tank level interlocks with and to address open Findings.

valves and pumps, loss of one CCW, perhaps a flow switch that starts the pump in the other train. The mapping of all items on the SSEL was reexamined Various interlocks that are associated with the with attention paid to instruments to ensure that their starting of a pump such as adequate lube oil consequential impacts have been properly linked to the pressure, cooling water flow, etc. PRA model. The methodology and results of the analysis are documented in the Waterford 3 Component and Cable The instruments and cables associated with Selection Report, PRA-W3-03-ES-01( Reference 17).

permissives and interlocks do not appear to have been comprehensively addressed in the PRA. Specific rationale of impacts on permissives and Starting interlocks for pumps and breaker closure interlocks is documented for several components and or tank interlocks that open or close valves or flow cables in the MSO Expert Panel discussions (PRA-W3-switches that start pumps all could have fire effects 03-ES-01, Reference 17) that would adversely affect the success of various system functions. The updated documentation references the site Nuclear Safety Capability Assessment (NSCA). The NSCA contains a thorough, comprehensive review and treatment of interlocks and permissives. This finding has been addressed in the 2019 model update (though it has not been through formal closure review). This finding has no impact on PRA results or risk informed applications.

CS-B1-01 CS-B1 CC-I Electrical coordination is addressed in the scenario The original finding was judged by the F&O closure teams development report (R0247070001.06 Appendix as Open.

E). Appendix E of R0247070001.06 provides information concerning electrical coordination. Breaker coordination is addressed in the updated Fire However, it is incomplete because the PRA model and documentation. An updated analysis that supplemental coordination evaluation is missing supplements the Component and Cable Selection report from the document. Preliminary coordination and the Plant Response Model report has been review has been performed and exists in an email developed. The analysis (PSA-WF3-03-ES-02, WF3 Fire (though not formally documented).

W3F1-2021-0003 Page 20 of 33 Table E2 Fire PRA Open Peer Review Findings Assessed During F&O Closure Review)

Supporting Capability Finding Requirement Category Description Disposition for RICT Program Number (s) (CC)

PRA - Circuit Analysis Notebook, Reference 18) provides Appendix E of R0247070001.06 provides a more thorough analysis of electrical coordination.

information concerning electrical coordination.

However, it is incomplete because the The updated model and documentation provide a supplemental coordination evaluation is missing. thorough resolution to the breaker coordination issue.

Information was provided by email and needs to be This finding has been addressed in the 2019 model formally incorporated, (finding CS-B1-01) CC II/III update. This finding has no impact on PRA results or risk when finished. informed applications.

ES-A2-01 ES-A2 Not Met R0247070001.02, Rev. 0, reviewed. Three issues The original finding was judged by the 2017 F&O closure identified: 1. The loss of equipment due to a loss teams as Open of room cooling caused by a fire damper going closed could not be demonstrated. 2. Although the The first item identified in the Finding is no longer an R0247070001.06 credited breaker coordination issue. Room cooling has been removed from the model from the SSD analysis, the PRA model does not based on room heatup evaluations that determined appear to address the effect of a fire interrupting cooling was not needed to support operation through the the relay circuit that would inhibit the coordination mission time. Room cooling is no longer in the PRA and allow a fault to transfer upstream. 3. The loss model except for the Main Control Room.

of DC from a fire does not appear to be addressed fully. For example, a fire induced loss of DC to the DC breaker impacts to the RCP pumps have been RCP breakers would inhibit the operator action to explicitly addressed and a (ex-control room) Recovery trip the RCPs from the control room. Note: Items Action was added to the model to address cases where 1 and 3 have outstanding questions to Waterford fire damage could impact the ability of the operators to trip

3. the pumps from the control room.

Breaker coordination (including DC breakers) is addressed in an analysis that supplements the Component and Cable Selection report and the Plant Response Model report. The analysis (PSA-W3-03-ES-02, WF3 Fire PRA - Circuit Analysis and Failure Probability Development, Reference 18) provides a thorough analysis of electrical coordination and describes modeling that was added to the PRA to correctly account for DC power requirements for overcurrent protection.

W3F1-2021-0003 Page 21 of 33 Table E2 Fire PRA Open Peer Review Findings Assessed During F&O Closure Review)

Supporting Capability Finding Requirement Category Description Disposition for RICT Program Number (s) (CC)

The updated model and documentation provide a thorough resolution to the breaker coordination issue.

This finding has been addressed in the 2019 model update. This finding has no impact on PRA results or risk informed applications.

ES-A3-02 ES-A3 CC-I/II/III The loss of DC does not appear to be adequately The original finding was judged by the 2017 F&O closure addressed in the fire PRA. For example, a failure teams as partially resolved.

CS-A3 Met of DC to supply control power to the RCP breakers would inhibit the operator action to trip the Reactor DC breaker impacts to the RCP pumps have been Coolant Pumps (RCPs) in a loss of seal cooling explicitly addressed and a (ex-control room) Recovery scenario. This was compensated for by a spurious Action was added to the model to address cases where start of the RCPs which would affect the same fire damage could impact the ability of the operators to trip state in the model. Similarly, a loss of DC power the pumps from the control room.

could potentially transfer a fault due to inhibition of coordination. Breaker coordination (including DC breakers) is addressed in an analysis that supplements the The plant has redundant DC supplies to the two Component and Cable Selection report and the Plant breakers which makes this failure less probable. Response Model report. The analysis (PRA-W3-03-ES-However, additional documentation is required to 02, Circuit Analysis and Failure Probability Development) clarify the issue. The fire effects on DC could provides a more thorough analysis of electrical adversely affect coordination as well as remote coordination and describes modeling that was added to operation of breakers. the PRA to correctly account for DC power requirements for overcurrent protection.

The updated model and documentation provide a thorough resolution to the breaker coordination issue.

This finding has been addressed in the 2019 model update. This finding has no impact on PRA results or risk informed applications.

ES-C2-01 ES-C2 Not Met Component and Cable Selection Report The original finding was judged by the 2017 F&O closure R0247070001.02, Revision 0 in Section 2.6 states, teams as partially resolved.

"An instrumentation review was conducted using the simulator and operators to identify single

W3F1-2021-0003 Page 22 of 33 Table E2 Fire PRA Open Peer Review Findings Assessed During F&O Closure Review)

Supporting Capability Finding Requirement Category Description Disposition for RICT Program Number (s) (CC) instrument reliance and single The Fire PRA specific operator interviews (Engineering indication/instruments whose malfunction would Change EC 46718) have the following statement: With cause operators to take action that would result in respect to the potential for undesired operation actions or un-recoverable states. No single instrument errors of commission (EOCs) in response to fire-induced vulnerabilities were identified." instrument failures, the interviewed Senior Reactor Operators (SROs) indicated that it was extremely unlikely However, there is no documentation or discussion that any single instrument failure would cause an EOC if of this activity. Engineering standard EN-FP-S- the alarm response procedure were implemented with 008-Multi has a process for reviewing indication verbatim compliance. The communications and conduct needs for post fire in the simulator in 5.3.4, but this of operations protocols at Waterford 3 require does not specifically address spurious indications confirmation with redundant and/or diverse indications that cause unwanted actions. prior to changing the state of any equipment. There were no indicators or alarms identified during the simulator What is needed to meet Category II for this SR is control panel walk downs for which the operator would be to develop a process for how various indications expected to take an unrecoverable immediate action that are reviewed and screened and then considered would otherwise be undesired.

for inclusion into the FPRA model. There is a sample process that the PWROG did for ERGs for The assessment of fire impacts on instruments combined Westinghouse sites, this process is more detailed with the interviews of operators (included in the than required for meeting this SR for this documentation), and the simulator walkdowns represents application but does show the process. No a process for identifying vulnerabilities (including single evidence other than statement that a simulator instrument) that is sufficient to meet the Standard walkdown was performed. requirements for ES-C2. This finding has no impact on quantified results and will have no impact on application of a RICT program.

FQ-C1-01 FQ-C1 Not Met Dependencies on combinations of HFEs The original finding was judged by the 2017 F&O closure have been utilized from the internal events teams as Open.

PRA, but not addressed for specific combination on a scenario by scenario basis. This needs to be The NFPA 805 LAR, RAIs, and SE had additional details done to ensure all combinations have been that were not in the PRA documentation. The Fire PRA addressed. HEPs and combinations have been properly developed following NUREG-1921 (Reference 29). Any action with Provide additional justification for selection of the any instrument/control impact from the fire are failed (set methodology use for combo events to address fire to 1). Multipliers are applied to all other events/combos to impacts. account for increase failure rate during a fire. Any events

W3F1-2021-0003 Page 23 of 33 Table E2 Fire PRA Open Peer Review Findings Assessed During F&O Closure Review)

Supporting Capability Finding Requirement Category Description Disposition for RICT Program Number (s) (CC) where these are not applied were explicitly evaluated to document the basis for such treatment.

The revision 6 Fire PRA documentation (PSA-WF3 PRM - Reference 19) has been revised to include more detail to ensure FQ-C1 is met with more thorough and detailed documentation.

As finding is related to documentation, it has no impact on quantified results and will have no impact on application of a RICT program.

FSS-C1-01 FSS-C1 CC-I Two points or a range of heat release values were The 2017 F&O Closure review judged this finding as not assigned to the ignition sources. A lower Open.

overall CDF will likely be achieved by using a two-point analysis or additional fire modeling to Waterford 3 applied a single Heat Release Rate (HRR) represent HRR profiles from ignition thru burnout Modeling Treatment. This treatment offers a means for and the corresponding probabilities of damage. incorporation of fire modeling into the fire PRA in a manner that eliminates the need for separate scenario specific analyses which require significant effort for configuration control, review and update.

In reviewing Waterford 3 F&O resolution for the Fire PRA applications ( - Record of Review Dispositions to Waterford 3 Internal Events PRA Facts and Observations (F&Os) ) graded this finding as follows. The NRC staff finds that the resolution of the F&O, as described by the licensee in the LAR, would have a negligible effect on the evaluations relied upon to support fire risk evaluations and has no impact on the conclusions of the risk assessment and therefore the resolution of the F&O is acceptable or this application.

This modeling limitation would have no impact specific to the implementation of a RICT program.

W3F1-2021-0003 Page 24 of 33 Table E2 Fire PRA Open Peer Review Findings Assessed During F&O Closure Review)

Supporting Capability Finding Requirement Category Description Disposition for RICT Program Number (s) (CC)

FSS-D7-01 FSS-D7 CC-I Fire detection and suppression system generic This was graded by the 2017 closure review team as unavailability values were used, and outlier partially resolved (with open documentation).

behavior and system unavailability were not specifically analyzed. To obtain a higher category, Plant specific suppression unavailability is applied to a specific WSES3 maintenance history review to suppression data used in the Fire PRA. PSA-WF3 assess outlier behavior is to be documented. This FSS-06 (WF3 Fire PRA - Nonsuppression Factor capability assessment is the same as for the Model Calculation, Reference 22) uses plant specific data and of Record (MOR). generic data to develop failure values used for each detection/suppression system credited. Any plant specific To move from CC-I to CC-II, specific WSES outlier behavior is explicitly included in the values.

maintenance history review to assess outlier behavior is to be documented. The F&O closure team judged this finding a documentation issue only. The Fire PRA documentation has been updated to address this finding. This finding has no impact on quantified results and will have no impact on application of a RICT program FSS-E3-01 FSS-E3 CC-I Evaluate the validity of other parametric The 2017 F&O Closure review judged this finding as uncertainty probability distributions (other than Open Ignition Frequency, which is the only quantitative uncertainty addressed). Quantitative uncertainty intervals were generated and documented. The closure team determined that inputs to Only qualitative discussions were provided with the uncertainty calculation were overly simplistic (error respect to the uncertainty intervals for most fire factors of 5 were applied to all ignition frequency inputs -

modeling parameters. Provide quantitative actual error factor data was not applied). The closure uncertainty intervals. team evaluated the uncertainty completed as insufficient to close the finding.

The 2019 Fire PRA update included detailed quantitative uncertainty evaluation for Fire PRA parameters. This includes quantitative uncertainty intervals for fire frequencies, fire suppression factors, modeled circuit failure inputs, and credited recovery events. The updated uncertainty document (PSA-WF3-03-UNC-02 -Fire PSA Parametric Uncertainty Evaluation- Reference 15)

W3F1-2021-0003 Page 25 of 33 Table E2 Fire PRA Open Peer Review Findings Assessed During F&O Closure Review)

Supporting Capability Finding Requirement Category Description Disposition for RICT Program Number (s) (CC) includes the necessary parametric uncertainty content to resolve this finding.

The documentation has been updated to address this finding. This finding has no impact on quantified results and will have no impact on application of a RICT program.

FSS-F2-01 FSS-F2 CC-I Criteria for structural collapse or non-collapse was This was graded by the 2017 closure review team as not provided. Only judgment statements were partially resolved.

provided, and these statements do not appear to reflect reality. The closure review team assessed that the non-suppression of the structural collapse scenario is not Re-perform the analysis to address the situation sufficiently evaluated.

where a turbine building collapse occurs due to a large turbine lube oil fire. Revise the documents to In the 2019 Fire PRA update:

eliminate the implication that failure of structural The TGB oil fire leading to collapse - manual suppression steel is not a credible event. is not credited, since it is a fast-growing oil fire, but auto suppression is credited, since there is a dedicated deluge system for the TG Oil skid, FPM-5. The failure probability is based on actual plant data. This scenario fails all of the FPRA targets in the Turbine Building (except those in the Turbine Building Switchgear Room).

The model and documentation have been updated to address this finding. This finding has no impact on quantified results and will have no impact on application of a RICT program FSS-H2-01 FSS-H2 CC-I Section 4.0 (scenarios report) details damage This was graded by the closure review team as partially criteria used in the Fire Scenarios. No cases of resolved (with open documentation).

where plant specific thresholds or damage mechanisms were used. At the closure review, this finding remained open only due to documentation. The assessment of location and Plant specific target damage evaluations that quantities for thermoset cables was thoroughly addressed involve combinations of thermoset and in the NFPA 805 LAR, RAIs, and SER. It was not

W3F1-2021-0003 Page 26 of 33 Table E2 Fire PRA Open Peer Review Findings Assessed During F&O Closure Review)

Supporting Capability Finding Requirement Category Description Disposition for RICT Program Number (s) (CC) thermoplastic cable have not been made. Only explicitly documented in the documents supporting the generic damage mechanisms have been used. Fire PRA model. This limitation was noted by the review team and resulted in the finding only being partially closed.

The 2019 Fire PRA documentation contains the necessary documentation to fully resolve this finding.

PSA-WF3-03-FSS-02 (Fixed Ignition Source Zone of Influence Methods - Reference 20) and PSA-WF3 FSS-03 (Waterford 3 Transient Fire Scenario Report-Reference 21) calculate damage thresholds for both thermoplastic and thermoset cables for various fire scenarios. Generic treatments are used for most scenarios. More detailed fire modeling is applied to the Relay Rooms. Different damage thresholds were also developed and applied for sensitive (solid state) electronics.

This has no impact on the RICT program. This is a documentation issue only and will have no impact on quantified results.

FSS-H3-01 FSS-H3 Not Met The basis for using the FDT tools over other fire This was graded by the 2017 closure review team as modeling tools has not been provided. It is clear partially resolved (with open documentation).

that the FDT tools have been V&V'd, (verification and validation) but their application to the specific At the closure review, this finding remained open only due scenarios involved in the analysis must also be to documentation. Much of the details of the verification documented. and validation for FDT is addressed in the NFPA 805 LAR, RAIs, and SER. At the time of the F&O Closure The basis for using the FDT tools over other fire review some of these details were not explicitly in the modeling tools has not been provided. documents supporting the Fire PRA model. This limitation was noted by the review team and resulted in the finding only being partially closed.

W3F1-2021-0003 Page 27 of 33 Table E2 Fire PRA Open Peer Review Findings Assessed During F&O Closure Review)

Supporting Capability Finding Requirement Category Description Disposition for RICT Program Number (s) (CC)

The necessary details have been added to the Fire PRA documentation as part of the 2019 Fire PRA model update. PRA-W3-03-FSS-02 (Waterford 3 PSA Fixed Ignition Source Zone of Influence Methods, Reference 20)

, documents the verification and validation (V&V) of the fire modeling tools used in the Waterford 3 fire PRA including fixed ignition sources, transient ignition sources, multicompartment scenarios, and the main control room analysis through a comparison to published parameter values related to the various fire modeling tools.

This has no impact a RICT program. This is a documentation issue only and will have no impact on quantified results.

HRA-A4-01 HRA-A4 Not Met WSES3 Fire Probabilistic Risk Assessment, This was graded by the closure review team as partially Quantification Model Preparation and Database resolved (with open documentation).

Development, R0247070001.03, Rev. 0, Appendix A, Single HFE Screening Process Results, has a At the time of the 2017 F&O Closure review, the report Cue source/ Instrumentation field that identifies with the documented interviews was complete and for a number of records, the applicable available, the interview attachment was not included in procedure(s) that would be utilized to address the the relevant Fire PRA report.

respective HFE. Additionally, Appendix D Detailed HRA for Selected HFEs, Fire PRA specific interview (for operator actions during does identify for certain HFEs, that there were fire events) have been conducted and documented. The limited STA reviews of the certain aspects of the interviews include discussion or procedures, cues, HFEs, namely the cues. However, documented instrumentation, as well as the potential for fire impacts to interviews with Operations to support the use of the actions. These interviews are documented in these cues/instruments were not found. Report Engineering Change EC 46718. (DOCUMENTATION OF PRAW3-01-001S03, Rev. 1, OPERATOR INTERVIEWS FROM APPENDIX E OF Operator Interview Sheets document the PRA PSA-W3-03-PRM)

Internal Events HRA events that were reviewed with Operations and includes the name of the This has no impact on a RICT program. This was judged Operator interviewed. No equivalent Operator a documentation issue. The documentation has been Interview sheets were located to address the resolved to address the issue. This Finding has no impact of fire.

W3F1-2021-0003 Page 28 of 33 Table E2 Fire PRA Open Peer Review Findings Assessed During F&O Closure Review)

Supporting Capability Finding Requirement Category Description Disposition for RICT Program Number (s) (CC) impact on quantified PRA results or application of a RICT program.

HRA-E1-01 HRA-E1 Not Met A part of the documentation uses future tense and This was graded by the 2017 closure review team as requires correction. The missing dependency partially resolved (with open documentation).

analysis needs to be added as well as the documentation for operator and training interfaces. The assessment of internal events PRA HRA actions used in the FPRA is discussed in the Fire PRA Plant Documentation issues with R0247070001.03, Response Model (PRM) Notebook (PSA-WF3-03-PRM -

Revision 0 section 5: Page 5-8 has a paragraph Reference 19). The impact of the fire on the action explaining that ex-mcr HFEs are set to true and (instruments, increased stress) is also evaluated. The then the risk significant HFEs are analyzed in more current/updated document provides a clearer description detail. If the HFE is set to true, then it would not of the applied methodology for HFEs.

show up in the cutsets. This paragraph needs a rewrite. Page 59 has a table explaining various The finding was judged partially closed due to issues treatments of HFEs in the model. Two of the concerning documentation of the HEP dependency columns conflict; one recommends a course of treatment for the fire PRA. The review team concluded action and the resolution takes another course with through the referenced report, RAI responses, and review no explanation of the differences. Also, there of the model recovery rule files that the treatment is should be some discussion about thermal acceptable. The open issue is that the details should all hydraulic analysis on any new sequences. be in the report and review of RAI and recovery rule files should not be necessary to make such judgments.

The Fire PRA update in 2019 provides a more thorough documentation of HFE treatment. The assessment of internal events PRA HRA actions used in the FPRA is discussed in the Fire PRA Plant Response Model (PRM)

Notebook (PSA-WF3-03-PRM - Reference 19). The impact of the fire on the action (instruments, increased stress) is also evaluated. The current document provides a clearer description of the applied methodology for HFEs.

The 2019 Fire PRA update provides more thorough documentation of the issue involved in this Finding. This

W3F1-2021-0003 Page 29 of 33 Table E2 Fire PRA Open Peer Review Findings Assessed During F&O Closure Review)

Supporting Capability Finding Requirement Category Description Disposition for RICT Program Number (s) (CC) finding has no impact on quantified PRA results or application of a RICT program.

IGN-A10-01 IGN-A10 CC-I The Fire Frequency results presented in Table 4-1 This was graded by the 2017 closure review team as are mean values only. Previous uncertainties Open.

associated with the Bin frequencies and the results of the Bayesian update are described. Quantitative uncertainty intervals were generated and Uncertainties associated with the application of the documented. The closure team determined that inputs to weighting factors which multiply the Bin the uncertainty calculation were overly simplistic (error frequencies are not discussed anywhere. There is factors of 5 were applied to all ignition frequency inputs -

no discussion of uncertainties in this area in the actual error factor data was not applied). The closure final quantification report. team evaluated the uncertainty completed as insufficient to close the finding.

Documentation of sources of uncertainty covers uncertainties associated with NUREG/CR-6850 The 2019 Fire PRA update included detailed quantitative Bin elements, and uncertainties associated with uncertainty evaluation for Fire PRA parameters. This the Bayesian update. No discussion on includes quantitative uncertainty intervals for fire uncertainties associated with partitioning and frequencies, fire suppression factors, modeled circuit weighting factor applications. The uncertainty failure inputs, and credited recovery events. The updated analysis is incomplete. Provide either a numerical parametric uncertainty document (PSA-WF3-03-UNC-02 uncertainty analysis or qualitative discussion of - Fire PSA Parametric Uncertainty Evaluation -

other sources of uncertainty as required by the Reference 15) includes the necessary parametric standard. uncertainty) includes the necessary parametric uncertainty content to resolve this finding.

A second uncertainty document, PSA-WF3-03-UNC-01 (Reference 14) documents sources of uncertainty and examines the impact the sources have on the model and results. The PSA-WF3-03-UNC-01 characterizes sources of uncertainty from all relevant Fire PRA tasks.

The documentation has been updated to address this finding. This finding has no impact on quantified results and will have no impact on application of a RICT program.

W3F1-2021-0003 Page 30 of 33 Table E2 Fire PRA Open Peer Review Findings Assessed During F&O Closure Review)

Supporting Capability Finding Requirement Category Description Disposition for RICT Program Number (s) (CC)

UNC-A1-01 UNC-A1 Not Met QU-E1 and QU-E2 requires identification of This was graded by the 2017 closure review team as sources of model uncertainty and assumptions. In Open.

general, WSES had an assumption section in each report. However, a simple search on "assume" A detailed uncertainty evaluation for each task in the showed that there were many more assumptions methodology is provided in the uncertainty report (PSA-than were listed in the assumption sections. At the WF3-03-UNC-01, Reference 14). The document provides CC-I level, QU-E3 requires estimation of the qualitative and quantitative uncertainties with explanations uncertainty interval of the overall CDF results. for the associated analyses.

WSES does not provide an estimation of the uncertainty interval for CDF. QU-E4 requires that The 2019 Fire PRA update includes two documents for each source of model uncertainty and related dedicated to uncertainty. One document covers assumption identified in QU-E1 and QUE2, quantitative uncertainty and includes a much more respectively; IDENTIFY how the PRA model is thorough assessment of parametric uncertainty.

affected. WSES only identifies how the model is Quantitative uncertainty intervals were generated and impacted for some of the assumptions and sources documented.

of uncertainty. The Uncertainty and Sensitivity Matrix in Appendix D of R0247070001.07. Per The 2019 Fire PRA update includes quantitative LE-F2, WSES should review LERF contributors for uncertainty intervals for fire frequencies, fire suppression reasonableness (e.g., to assure excessive factors, modeled circuit failure inputs, and credited conservatisms have not skewed the results, level recovery events. The updated parametric uncertainty of plant specificity is appropriate for significant document (PSA-WF3-03-UNC Reference 15) contributors, etc.). There is no evidence that such includes the necessary parametric uncertainty content to a review was performed.

resolve this finding.

Additional explanation should be provided for the documented entries. The meaning of the A second uncertainty document, PSA-WF3-03-UNC-confidence intervals for the different values 01(Reference 14) - Waterford 3 Fire PRA Sensitivity and (mean, 5th/95th, and median) would not be Uncertainty Report - documents sources of uncertainty obvious to most readers. It is suggested that this and examines the impact the sources have on the model discussion be corrected in the next update of the and results. This report (as the finding suggests) documentation. The usefulness of the qualitative evaluates uncertainty topics from EPRI 1026511. The evaluation of modeling uncertainties could be PSA-WF3-03-UNC-01 characterizes sources of significantly enhanced by additional comments uncertainty from all relevant Fire PRA tasks and examines regarding their potential impacts. EPRI 1026511 the impact of assumptions made in those tasks.

provides a tabulation of sources of modeling uncertainty associated with fire. There is no

W3F1-2021-0003 Page 31 of 33 Table E2 Fire PRA Open Peer Review Findings Assessed During F&O Closure Review)

Supporting Capability Finding Requirement Category Description Disposition for RICT Program Number (s) (CC) indication that this report was considered in The documentation has been updated to address this identifying the sources discussed in the finding. This finding has no impact on quantified results documentation. Because this has become a and will have no impact on application of a RICT program.

standard reference, and is a companion document to NUREG-1855, it would be worthwhile to check EPRI 1026511 to identify any additional sources that should be discussed.

Note - All findings in Table 2-4 are F&Os that remain open/partially open following a 2017 F&O Closure review. The original findings evaluated during the closure review were from a 2011 full scope fire peer review and two focused scope peer reviews conducted in September 2012 and May 2013.

W3F1-2021-0003 Page 32 of 33 References

1. LTR-RAM-II-09-39 RG 1.200 PRA Peer Review Against the ASME PRA Standard Requirements For The Waterford Steam Electric Station, Unit 3 Probabilistic Risk Assessment (Westinghouse Owners Group August 2009).

2. Waterford Steam Electric Station, Unit 3 - Issuance of Amendment Regarding Transition to a Risk-Informed, Performance-Based Fire Protection Program in Accordance with 10 CFR 50.48(c) (CAC No. ME7602) (ADAMS Accession No ML16126A033). June 2016.

3. Waterford Steam Electric Station, Unit 3 - Issuance of Amendment RE: Adoption of TSTF-425, Revision 3 Relocate Surveillance Frequencies to Licensee Control - RITSTF Initiative 5b (CAC No. MF6366) (ADAMS Accession No ML16159A419). July 2016.

4. ASME RA-Sb-2005 - Addenda to ASME RA-S-2002: Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications. July 2007.

5. ASME/ANS RA-Sa-2009, Standard for Level l/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, Addendum A to RA-S-2008, ASME, New York, NY, American Nuclear Society, La Grange Park, Illinois, dated February 2009.

6. PSA-WF3-08-01, Waterford 3 PRA Peer Review Gap Assessment to 2009 PRA Standard.

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. LTR-RAM-II-11-003 Fire PRA Peer Review of Waterford Steam Electric Station Unit 3 Fire Probabilistic Risk Assessment against the Fire PRA Standard Supporting Requirements from Section 4 of the ASME/ANS Standard.

9. PSA-WF3-08-02, Waterford 3 Finding Level F&O Independent Technical Review.

10. PSA-WF3-08-06, Revision 0 - PRA Technical Adequacy to Support Risk-Informed Applications.

11. PSA-WF3-01-QU, Revision 3 - WF3 PSA At-Power Level 1 Integration and Quantification Analysis.

12. PSA-WF3-01-QU-01, Revision 2 - WF3 PSA Uncertainty and Sensitivity Analysis.

13. PSA-WF3-01-IF-SOU - Waterford 3 Internal Flooding Sources of Uncertainty.

14. PSA-WF3-03-UNC WF3 Fire PRA Sensitivity and Uncertainty Report.

15. PSA-WF3-03-UNC Fire PSA Parametric Uncertainty Evaluation.

16. PSA-WF3-03-FQ-01 Waterford 3 Fire PRA Quantification Report.

W3F1-2021-0003 Page 33 of 33

17. PSA-WF3-03-ES-01, R0, Fire PRA Equipment and Cable Selection Notebook.

18. PSA-WF3-03-ES-02, R0, Circuit Analysis and Failure Probability Development.

19. PSA-WF3-03-PRM, R0, Fire PRA Plant Response Model (PRM) Notebook.

20. PSA-WF3-03-FSS-02, R0, Waterford 3 PSA Fixed Ignition Source Zone of Influence Methods.

21. PSA-WF3-03-FSS-03, R0, Waterford 3 Transient Fire Scenario Report.

22. PSA-WF3-03-FSS-06, R0, Nonsuppression Factor Calculation.

23. PSA-WF3-01-IF-QU, Revision 0 - WF3 Internal Flooding Quantification Report -

EC76543.

24. PSA-WF3-01-HR, WF3 At-Power Human Reliability Analysis Notebook

25. PSA-WF3-01-LE, WF3 PSA Large Early Release Analysis Notebook

26. PSA-WF3-08 Focused Scope Fire PRA Peer Review of Waterford Steam Electric Station Unit 3 Fire Probabilistic Risk Assessment Against the Fire PRA Standard Supporting Requirements from Section 4 of the ASME/ANS Standard (Robert Brady -URS Corporation - September 2012)

27. PSA-WF3-08 2nd Focused Scope Fire PRA Peer Review of Waterford Steam Electric Station Unit 3 Fire Probabilistic Risk Assessment Against the Fire PRA Standard Supporting Requirements from Section 4 of the ASME/ANS Standard (Robert Brady -URS Corporation - May 2013)

28. PSA-WF3-08 Waterford 3 Internal Flood/LERF Focused Scope Peer Review.

August 2019.

29. NUREG 1921, EPRI/NRC-RES Fire Human Reliability Analysis Guidelines - Final Report.

July 2012.

Enclosure 3 W3F1-2021-0003 Information Supporting Technical Adequacy of PRA Models without PRA Standards Endorsed by Regulatory Guide 1.200, Revision 2 This Enclosure is Not Applicable to the Waterford 3 submittal. Waterford 3 is not proposing to use any PRA models in its Risk-Informed Completion Time Program for which a PRA standard, endorsed by the NRC in RG 1.200, Revision 2 does not exist.

Enclosure 4 W3F1-2021-0003 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models

W3F1-2021-0003 Page 1 of 13 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models

1. Introduction and Scope Topical Report NEI 06-09, Revision 0-A (Reference 1), as clarified by the Nuclear Regulatory Commission (NRC) final safety evaluation (Reference 2), requires that the License Amendment Request (LAR) provide a justification for exclusion of risk sources from the Probabilistic Risk Assessment (PRA) model based on their insignificance to the calculation of configuration risk as well as discuss conservative or bounding analyses applied to the configuration risk calculation.

This enclosure addresses this requirement by discussing the overall generic methodology to identify and disposition such risk sources. This enclosure also provides the Waterford 3 Nuclear Power Plant (WF3) specific results of the application of the generic methodology and the disposition of impacts on the WF3 Risk Informed Completion Time (RICT) Program. Table E4-1 of this enclosure presents the plant-specific bounding analysis of seismic risk to WF3. Table E4-1 also presents the justification for excluding analyses of other external hazards from the WF3 PRA.

Topical Report NEI 06-09 does not provide a specific list of hazards to be considered in a RICT Program. However, NUREG-1855, "Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decision Making" (Reference 4), provides regulatory guidance on risk-informed decision making relative to hazards that are not considered in the PRA model.

Specifically, Section 6 of NUREG-1855 provides the following list of external hazards that should be addressed either via a bounding analysis or included in a PRA calculation:

Minimum Scope of External Hazards to be Considered x Seismic Events x Accidental Aircraft Impacts x External Flooding x Extreme Winds and Tornados (including generated missiles) x Turbine-Generated Missiles x External Fires x Accidents From Nearby Facilities x Release of Chemicals Stored at the Site x Transportation Accidents x Pipeline Accidents (e.g., natural gas)

Waterford 3 completed an analysis in 2017 that re-examined the external events considered in the site Individual Plant Examination of External Events (IPEEE). That evaluation of external events was based on the IPEEE contents and is not exactly the same as the NURGEG-1855 list. However, the NUREG-1855 phenomena are covered in the updated evaluation (see table E4-1 for details). The scope of this enclosure is consideration of external hazards and the primary source for the conclusions is the update IPEEE external hazard evaluation. As explained in subsequent sections of this enclosure, risk contributions from seismic events are evaluated quantitatively, while all the other listed external hazards are evaluated and screened as having low risk.

W3F1-2021-0003 Page 2 of 13

2. Technical Approach The guidance contained in NEI 06-09 states that all hazards that contribute significantly to incremental risk of a configuration must be quantitatively addressed in the implementation of the RICT Program. The following approach focuses on the risk implications of specific external hazards in the determination of the risk management action time (RMAT) and RICT for the Technical Specification (TS) Limiting Conditions for Operation (LCOs) selected to be part of the RICT Program.

Consistent with NUREG-1855, external hazards may be addressed by:

1. Screening the hazard based on a low frequency of occurrence,

2. Bounding the potential impact and including it in the decision-making or

3. Developing a PRA model to be used in the RMAT/RICT calculation.

The ASME/ANS PRA Standard (Reference 5) has endorsed the following set of five external hazard screening criteria:

1. The hazard would result in equal or lesser damage than the events for which the plant has been designed. This requires an evaluation of plant design bases to estimate the resistance of plant structures and systems to a particular external hazard.

2. The hazard has a significantly lower mean frequency of occurrence than another event (taking into account the uncertainties in the estimates of both frequencies), and the hazard could not result in worse consequences than the other event.

3. The hazard cannot occur close enough to the plant to affect it. Application of this criterion needs to take into account the range of magnitudes of the hazard for the recurrence frequencies of interest.

4. The hazard is included in the definition of another event.

5. The hazard is slow in developing, and it can be demonstrated that sufficient time exists to eliminate the source of the threat or to provide an adequate response.

The review of external hazards considers two aspects of the contribution to risk. The first is the contribution from the occurrence of beyond design basis conditions (i.e., winds greater than design). These beyond design basis conditions challenge the functionality of the systems, structures, and components (SSCs) to support safe shutdown of the plant. The second aspect addressed are the challenges caused by external conditions that are within the design basis, but still require some plant response to assure safe shutdown (i.e., high winds causing loss of offsite power). While the plant design basis assures that the safety-related equipment necessary to respond to these challenges are protected, the occurrence of these conditions nevertheless cause a demand on these systems and can impact configuration risk. Note that when the effect of a particular hazard is not mitigatable using the plant SSCs, then there is no impact on the changes in risk calculated to support the RICT Program, and so these hazards can be screened as well. Only events which create a demand for mitigation equipment are potentially relevant to the RICT Program.

The review and disposition of each external hazard is addressed in Table E4-1. Table E4-1 includes disposition of external hazards for both screened hazards and ones that will impact RICT evaluations (not screened). Unless otherwise specified, all information is based on PSA-WF3-07-01, Waterford 3 - Re-Examination of External Events Evaluation in the IPEEE (Reference 7). This report was completed in 2017 to re-evaluate IPEEE related external events to consider updated inputs and insights.

W3F1-2021-0003 Page 3 of 13 Table E4-1 External Hazard Evaluation External Hazard Evaluation Disposition for WF3 RICT Program Aircraft Impacts There are five airports within 10 miles of the Waterford 3 There are only five airports within 10 miles of site. The MYU Heliport, the River Parish Hospital Heliport the plant and they have been screened out and Port of South Louisiana Executive Regional Airport fall from further consideration or analyzed to be so within the 10-mile boundary. All listed airports/heliports small as to not pose a hazard for WF3. Of the within this boundary have limited flight traffic and screen airports greater than 10 statute miles from the from detailed consideration. Louis Armstrong New Orleans WF3 site, they have either been screened out International Airport, while outside the 10-mile boundary, is or demonstrated to not pose a hazard for close to the 10 mile cutoff and has a high volume of air WF3. There are no military facilities or military traffic. training routes close to the plant.

The bounding analysis associated with the risk The Louis Armstrong New Orleans International Airport is of aircraft impact on the WF3 site is located approximately 11.5 miles from the site at the closest documented in Reference 7.

point. This is an international airport for a large city and has The assessment concludes that aircraft hazard significant flight traffic. The risk of aircraft impact to the can be screened out from an external events WF3 from Louis Armstrong Airport is evaluated in the WF3 PRA for WF3.

Re-Examination of External Events Evaluation in the No unique PRA model for aircraft impact IPEEE (Reference 7). The evaluation concludes that scenarios is required in order to assess flights from the airport do not pose a substantial risk to the configuration risk for the RICT Program.

site. There are no additional large airfields near the site and no military installations near the site.

External Flooding The WF3 FSAR indicates that major floods have occurred In summary, the updated examination of on average every seven years. The FSAR cites the most external flood risk, including the updated plant recent floods to have occurred in 1973 and 1975. Flooding data, flood history and new measures for risk events following that time are as follows: management validate the current flood x On August 28, 2005, Waterford 3 shut down due to mitigation strategy of the current design Hurricane Katrina approaching and declared an basis. External flooding events will cause no unusual event. Shortly after Katrina, Waterford 3 flooding of WF3 safety-related structures, restarted and resumed normal operation. While systems and components (SSCs).

significant flooding occurred in the general area, the WF3 site did not experience any significant flooding Based upon the above information, a from Hurricane Katrina. quantitative evaluation of the external flooding

W3F1-2021-0003 Page 4 of 13 External Hazard Evaluation Disposition for WF3 RICT Program x The Mississippi River floods in April and May 2011 hazard is not required in order to assess were among the largest and most damaging configuration risk for the RICT Program.

recorded along the U.S. waterway in the past century, comparable in extent to the major floods of Additionally, in NRC letter to Waterford 1927 and 1993. During this period, Waterford 3 (ML17171A128) WATERFORD STEAM entered a refueling outage on April 6, 2011. The ELECTRIC STATION, UNIT 3- FLOOD plant was restarted on May 12, 2011. FOCUSED EVALUATION ASSESSMENT (Reference 12), the NRC staff The Army Corps of Engineers now actively creates and concluded that the Waterford 3 Flood maintains spillways and floodways to divert periodic water Evaluation is consistent with approved surges into backwater channels and lakes, as well as route guidance (NEI-16-05) and that the site has part of the Mississippi's flow into the Bonnet Carre Spillway effective flood protection against hurricane, and/or the Atchafalaya Basin and from there to the Gulf of storm surge, damn/levee failure, and river and Mexico, bypassing New Orleans. Some of the pre-1927 stream related beyond design basis flood strategy is also still in use today, with the Corps actively events.

cutting the necks of horseshoe bends, allowing the water to move faster and reducing flood heights.

Extreme Winds and The licensing basis for extreme winds and tornadoes for WF3 has been designed for extreme winds Tornados (including WF3 is described in FSAR Section 2.3.1.2.4. Extreme and tornado loadings that are substantially generated missiles) winds at WF3 site come from thunderstorms, hurricanes higher than the licensing basis events and tornadoes. Seismic Category I structures are also presently required. The safety related designed to withstand the postulated spectrum of tornado structures, systems and components (SSCs) missiles. Re-evaluating IPEEE events found no significant are protected from tornado missiles using changes since the FSAR. barriers with thicknesses exceeding the x All Seismic Category I structures are designed for current requirements based on recent tornado the 100-year wind speed of 110 mph (equivalent to 3 hazard analysis. It is concluded that the second gust speed of 130 mph). Tornado loadings hazard event of extreme winds and are based on a 300-mph rotational wind speed and tornadoes may be screened out for WF3 and a 60-mph translational wind speed, with a no unique PRA model is required in order to simultaneous maximum atmospheric pressure drop assess configuration risk for the RICT of 3 psi in 3 seconds. Non-Category I structures Program.

have been designed to not collapse on or impact Seismic Category I structures.

W3F1-2021-0003 Page 5 of 13 External Hazard Evaluation Disposition for WF3 RICT Program x The American Society of Civil Engineers standard Tornados and high wind hazards are ASCE 7-10 provides requirements for general considered for the initiating events analysis for structural design and the means for determining the probabilistic risk analysis (PRA) model for wind loads, based on wind speed data. The 100- loss of offsite power (LOOP) per year wind speed of 130 mph (3 second gust value) PSA-WF3-01-IE-01 (Reference 9) specified in ASCE 7-10 is less than or equal to the WF3 licensing basis wind speed (Reference 7).

Regulatory Guide 1.76 Revision 3 has placed the Waterford 3 site in Region I for licensing basis tornado. The characteristics of the licensing basis tornado for Region I are: Maximum rotational speed = 184 mph; Maximum translational speed = 46 mph; Maximum wind speed = 230 mph (= 184 mph + 46 mph); Maximum pressure drop = 1.2 psi.

x The tornado missile spectrum specified in RG 1.76 Revision 3 differs somewhat from that used in the design of WF3; however, the minimum concrete thicknesses for wall and roof obtained using the RG 1.76 Revision 3 missile spectrum for Region I are less than those provided in the WF3 FSAR.

x The maximum wind speed for the Waterford 3 site corresponding to an exceedance probability of 1E-7 per year is 204 mph on the enhanced Fujita Scale (NUREG/CR-4461, Revision 2, Table 6 Reference 14). This value is less than that specified in RG 1.76 and is much less compared to the licensing basis tornado for WF3.

x No new construction or modification of existing plant structures were identified that would impact the safety-related structures in the event of a tornado. Therefore, safety-related structures, systems and components at Waterford Unit 3 are

W3F1-2021-0003 Page 6 of 13 External Hazard Evaluation Disposition for WF3 RICT Program adequately protected against tornado generated missiles.

External Fires The IPEEE and re-examination of IPEEE (Reference 7) External fires (wild fires) not related to considered external fires and explosions. This included transportation accidents or nearby facilities industrial facilities and transportation pathways (river, road, was not addressed in the IPEEE or the and rail) near the site. The potential hazards from these updated external hazard evaluation.

facilities and paths include fire and explosion The location and geography of the Waterford 3 hazards. Licensing basis events are potential accidents site makes wild fire/forest fire type threats to that have a probability of occurrence equal to or greater plant a remote hazard that does not need than 1E-7 per year whose consequences can result in further evaluation. The plant is surrounded by radionuclide releases more than 10 CFR 50.67 guidelines. a river, swamp/wetlands, and other industrial facilities.

External Fires from nearby facilities, mobile transient fire hazards as, and explosion/deflagration events are evaluated in other sections (Nearby Facilities/pipeline/transportation accidents) in this table.

Accidents From Accidents from nearby facilities were evaluated in the Accidents involving fire can be evaluated by Nearby Facilities IPEEE and reconsidered in PSA-WF3-07-01 (Reference thermal effects at the scene of an accident and 7). leakage or spillage of flammable materials The referenced evaluations considered hazards posed forming clouds that drift towards the plant, with by both stationary and transient sources and considered all delayed ignition and deflagration of said vapor relevant toxins known to be stored (within 5 miles) or clouds possibly affecting safety-related transported near the site. The evaluation was updated in structures. Explosion hazards generate 2017 to evaluate the original methodology and consider overpressure events which could endanger current data and insights. safety-related structures. Both explosion and deflagration hazards diminish due to Review of all combustible materials transported or stored increasing distance from the plant at a rate of within five miles of Waterford 3 revealed that several 1/r2, with r being the distance from the event to the site. The PSA-WF3-07-01(Reference 7)

W3F1-2021-0003 Page 7 of 13 External Hazard Evaluation Disposition for WF3 RICT Program sources presented hazards which merited closer report documents that the thermal fluxes and investigation. The sources of hazards are: time durations of the vapor cloud fireballs are

1. Liquified Petroleum Gas (LPG) shipments by insufficient to produce appreciable heating of truck passing the site at a closest distance of plant structures and equipment and that the 634 ft. from the main control room (see limiting acceptable overpressure of 3.0 psi discussion in Transportation Accidents would be caused by an explosion of a truck below), carrying LPG on Louisiana Route 18

2. river transport of gasoline along the shipping (PSA-WF3-07-01). That was the most severe channel which passes 1200 ft. north of the safety hazard evaluated. No unique PRA model for related structures (see discussion in external fire scenarios is required in order to Transportation Accidents below), assess configuration risk for the RICT

3. nearby gas pipelines and LPG lines (see Program.

discussion in Pipeline Accidents below),

4. flammable stationary sources, and Based on the evaluations reported in

5. potential hazards from externally generated EC-S97-025 (Reference 10) and the FSAR on missiles. storage and handling of toxic chemicals near the site, it is concluded that this hazard group The re-examination of IPEEE (Reference 7) covers specific does not pose a credible threat to WF3.

sources of fire and explosives from nearby facilities and Therefore, the external hazard from industrial concludes that all may be screened from detailed risk and military facilities accidents could be evaluation (and RICT calculations). screened out from the WF3 PRA and no unique PRA model for accidents from nearby facilities is required in order to assess configuration risk for the RICT Program.

Pipeline Accidents The Re-Examination of External Events Evaluation in the Based on the recent (2017) re-evaluation of (e.g., natural gas) IPEEE (Reference 7) included considerations of pipeline IPEEE hazards (including pipelines), it is accidents and their potential impact to the site. concluded that this hazard group does not The 12 pipeline operators within 2 miles of WF3 were pose a credible threat to WF3. Therefore, the identified and included in the evaluation. The evaluation external hazard from pipeline accidents could looked at the size and content of the pipelines and the be screened out from the WF3 PRA and WF3 proximity of the hazards to the site. This Risk Informed Applications.

analysis includes considering the most significant/bounding accidents based on the content of the pipelines. The

W3F1-2021-0003 Page 8 of 13 External Hazard Evaluation Disposition for WF3 RICT Program updated analysis concluded that the original IPEEE remains bounding with regard to pipeline accidents and that explicit consideration in PRA models is not necessary (Reference I 7).

Release of Section 5.1 of the IPEEE states Waterford 3 further The 2017 report examining IPEEE conclusions Chemicals concludes that the plant is in conformance with the 1975 confirmed the original IPEEE assessment Stored at the Site SRP that pertains to high winds, on-site storage of regarding on site chemicals. Based on the hazardous materials, and off-site developments. evaluations reported in EC-S97-020 (Reference 11) and the FSAR on storage and The chemicals stored on the WF3 site were analyzed with handling of toxic chemicals onsite, it is other stationary and transient toxic chemicals in EC-S97- concluded that this hazard group does not 020. Therefore, onsite sources of toxic chemicals do not pose a credible threat to WF3. Therefore, pose a threat to control room habitability in 95 percentile the internal hazard from chemical meteorological conditions. release accidents may be screened out from the WF3 PRA and no unique PRA model is required in order to assess configuration risk for the RICT Program.

Transportation Site risk from local transportation accidents were evaluated Based on the evaluations reported in Accidents in 2016 and 2017 to update IPEEE evaluations with current references 7 and 8, and the WF3 FSAR on data. References 7 and 8 detail truck, rail, barge, and the storage, handling, and transport of toxic potential hazardous contents of each related to the WF3 chemicals near the site, it is concluded that site. this hazard group does not pose a credible threat to WF3. Therefore, the external hazard These engineering reports contains an updated survey of from transportation accidents could be chemicals transported via railroad, Mississippi River screened out from the WF3 PRA and no transportation and trucks on Louisiana 18. The frequency of unique PRA model is required in order to shipment of toxic chemicals is another screening criterion assess configuration risk for the RICT for transient sources per Reg. Guide 1.78: ten shipments Program.

per year for trucks, 30 per year for rail cars and 50 per year for barges. Transient chemicals which did not screen out were analyzed in EC-S97-020 and EC-S97-025. Both calculations concluded that sources of toxic chemicals

W3F1-2021-0003 Page 9 of 13 External Hazard Evaluation Disposition for WF3 RICT Program within 5 miles of WF3 do not pose a threat to control room habitability in 95 percentile meteorological conditions. The new Annual Radiological Release Probability increased slightly to 6.450E-7, which includes the 0.1 probability of release given incapacitation. When this probability is compared to the guidance in Section 2.2.3 of Regulatory Guide 1.70 (Reference 16), it is below the 1E-6 per year criterion. Therefore, the results indicate that the protective features described in the FSAR provide adequate protection for the control room operators.

Turbine-Generated The turbine-generators for WF3 are manufactured by The Waterford 3 turbine design has an Missiles Westinghouse and are described in FSAR Section 10.2. acceptably low probability for destructive Both plant specific (PSA-WF3-07-01) and vendor missile generation and corresponding (Westinghouse) Calculations concluded that there is a low damage.

frequency of core damage due to turbine-generated missile related events. Based on the recent (2017) re-evaluation of IPEEE hazards (including turbine-generated Some WF3 related information regarding turbine-generated missiles), it is concluded that this hazard group missiles: does not pose a credible threat to WF3.

x Tables 3.5-9g and 3.5-9h in the FSAR include plant unacceptable turbine missile strike and damage No unique PRA model for turbine missiles is event probabilities. The value for a design required in order to assess configuration risk overspeed event for is 2.64E-8 per year. For Low for the RICT Program.

Trajecting Missiles (LTM) and High Trajecting Missiles (HTM) destructive overspeed events, those values are 3.36E-8 per year and 3.39E-8 per year, respectively. These values are all less than the acceptable value of 1E-7 per year.

x Westinghouse performed an updated calculation in 2006 that included normally running, as well as design, intermediate, and destructive overspeed cases. The normally running case is a new turbine missile scenario added for this analysis. All

W3F1-2021-0003 Page 10 of 13 External Hazard Evaluation Disposition for WF3 RICT Program quantified cases in the analysis had per year results for missile damage probabilities in the 1E-08 to 1E-11 range. These values are lower than the acceptable value specified in Regulatory Guide 1.115 of 1E-7 per year (Reference 7).

Seismic Events Due to the plant being located in a region of very low Seismic risk is not a significant contributor to seismicity, a seismic PRA was not developed for Waterford configuration risk calculations of the RICT

3. Reviewing past Seismic CDF (SCDF) estimates and Program for Waterford 3; however, a bounding updated seismic information - it was found that seismic risk estimate for it has been developed to include cannot be excluded from the calculation of configuration- in RICT evaluations.

specific risk. An updated (bounding) seismic risk estimate was developed. Updated seismic hazard information from Seismic risk will be included in RICT and the Near Term Task Force (NTTF) in response to RMAT calculations by adding an Fukushima was used. incremental 3.2E-06/year and 3.2E-07/year bounding seismic contribution to the A plant high-confidence low-probability of failure (HCLPF) configuration-specific delta CDF/delta LERF, was obtained for seismic capacity. The plant level fragility respectively, attributed to internal and fire as reported in EPRI Letter Fleet Seismic Core Damage events contributions.

Frequency Estimates for Central and Eastern U.S. Nuclear Power Plants Using New Site-Specific Seismic Hazard This method ensures that an incremental Estimates (Reference 15) is 0.25g. seismic CDF/LERF equal to the bounding SCDF/SLERF is added to internal and fire A convolution was performed over the full range of the Peak events incremental CDF/LERF contribution for Ground Acceleration (PGA), 1, 5 and 10 Hz hazard curves every RICT occurrence.

with a 0.25g plant HCLPF. The bounding seismic evaluation concludes that an estimate of 3.2E-06/yr CDF A permanent SCDF of 3.2E-06/yr and SLERF should be used in RICT evaluations. of 3.2E-07/yr will also be added to the W3 configuration risk management model to There are several sources of conservatism inherent in this quantify instantaneous CDF/LERF whenever a approach. The seismic contribution to delta risk for the RICT RICT is in effect (Reference 6).

calculation for any configuration is taken as the full seismic plant CDF / LERF. That is, for the purpose of any RICT calculation, delta seismic CDF is assumed to be equal to

W3F1-2021-0003 Page 11 of 13 External Hazard Evaluation Disposition for WF3 RICT Program the estimated seismic CDF from the convolution of the seismic hazard curves with the limiting HCLPF. The full annual seismic frequency is applied to the seismic contribution for all RICT calculations, regardless of the duration of the RICT. Since the maximum duration for a RICT is limited to the 30-day backstop, the estimated seismic CDF is roughly 12 times the contribution applicable during any RICT. The presumption of the plant HCLPF leading directly to seismic core damage means that no other failures, whether related to components in an LCO or not, and whether treated as correlated or not, are required for the calculation of seismic CDF (i.e., the use of plant level HCLPF assumes that enough equipment failures occur to lead directly to core damage.), and these would not increase the estimated seismic CDF (Reference 6).

A factor 0.1 is used to calculate SLERF from SCDF. The W3 internal events LERF is over two orders of magnitude lower than CDF (so the 0.1 factor is conservative relative to calculated site risk from other hazards). The potential exists for a seismic event to contribute to a bypass of containment, such as impacts to containment isolation valve(s) or spurious operations produced from contact chatter. However, detailed evaluation is expected to result in containment piping and penetrations being more seismically robust than other equipment. The applied 0.25g HCLPF value is likely overly conservative for containment related equipment and structures and detailed evaluation would expect to result in less than a 0.1 CDF to LERF factor.

W3F1-2021-0003 Page 12 of 13

6. Conclusions Based on this analysis of external hazards for WF3, no additional external hazards other than seismic events are quantitatively considered. The evaluation concluded that the hazards either do not present a design-basis challenge to WF3, the challenge is adequately addressed in the PRA, or the hazard has a negligible impact on the calculated RICT and can be excluded.

The ICDP/ILERP acceptance criteria of 1E-5/1E-6 will be used to calculate the resulting RICT and RMAT based on the total configuration-specific delta CDF/LERF attributed to internal events (including flood) and internal fire, plus the seismic bounding delta CDF/LERF values. These CDF/LERF values are based on the bounding seismic assessment for CDF and a conservative conditional large early release probability of 0.1 (Reference 6).

5. References

1. Nuclear Energy Institute (NEI) Topical Report (TR) NEI 06-09, "Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines,"

Revision 0-A, (ADAMS Accession No. ML12286A322). October 2012.

2. NRC Letter from Jennifer M. Golder (NRC) to Buff Bradley (NEI), "Final Safety Evaluation for Nuclear Energy Institute (NEI) Topical Report (TR) NEI 06-09, 'Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines,"'" (ADAMS Accession No. ML071200238). May 2007

3. WCAP-1 6952-NP, "Supplemental Implementation Guidance for the Calculation of Risk Informed Completion Time and Risk Managed Action Time for RITSTF Initiative 4B," August 2010.

4. NUREG-1855, "Guidance on the Treatment of Uncertainties Associated with PRAs in Risk- Informed Decision Making," Volume 1, March 2009.

5. NUREG-75/087, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, LWR Edition," 1975.

6. PSA-WF3-04-01, Revision 1, Waterford 3 Seismic Risk Bounding Evaluation

7. PSA-WF3-07-01, Revision 0, Re-Examination of External Events Evaluation in the IPEEE (2017)

8. WF3-IC-15-00001 2016 Survey of Hazardous Chemicals Stored, Processed, Or Transported in the Vicinity of Waterford 3 Steam Electric Station August 2016

9. PSA-WF3-01-IE-01 "WF3 PSA Loss of Offsite Power"

10. EC-S97-025 Control Room Habitability Following Accidental Chlorine Release

11. EC-S97-020 Toxic Chemical Analysis to Assess Control Room Habitability

12. NRC Letter, WATERFORD STEAM ELECTRIC STATION, UNIT 3- FLOOD FOCUSED EVALUATION ASSESSMENT (CAC NO. MF9710; EPID L-2017-JLD-0009). (ADAMS Accession No. ML12121A128). February 2018.

W3F1-2021-0003 Page 13 of 13

13. The American Society of Civil Engineers standard ASCE 7-10

14. NUREG/CR-4461, Revision 2, Tornado Climatology of the Contiguous United States 2007.

15. EPRI Letter Fleet Seismic Core Damage Frequency Estimates for Central and Eastern U.S. Nuclear Power Plants Using New Site-Specific Seismic Hazard Estimates, (ML14083A586). March 2014

16. Regulatory Guide 1.70, Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants, LWR Edition, USNRC, Revision 3, November 1978.

Enclosure 5 W3F1-2021-0003 Baseline CDF and LERF

W3F1-2021-0003 Page 1 of 2 Baseline CDF and LERF

1. Introduction Section 4.0, Item 6 of the U.S. Nuclear Regulatory Commissions (NRC) Final Safety Evaluation (Reference 1) for NEI 06-09, Revision 0-A, "Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines," (Reference 2) requires that the license amendment request (LAR) provide the plant-specific total CDF and LERF to confirm applicability of the limits of Regulatory Guide (RG) 1.174, Revision 1 (Reference 3). (Note that RG 1.174, Revision 2 (Reference 4), issued by the NRC in May 2011, did not revise these limits.)

The purpose of this enclosure is to demonstrate that the Waterford 3 total Core Damage Frequency (CDF) and total Large Early Release Frequency (LERF) are below the guidelines established in RG 1.174. RG 1.174 does not establish firm limits for total CDF and LERF but recommends that risk-informed applications be implemented only when the total plant risk is no more than about 1E-4/year for CDF and 1E-5/year for LERF. Demonstrating that these limits are met confirms that the risk metrics of NEI 06-09 can be applied to the WF3 Risk Informed Completion Time (RICT) Program.

2. Technical Approach Table E5-1 lists the Waterford 3 CDF and LERF values that resulted from a quantification of the baseline internal events model, internal flooding model, and fire PRA model. This table also includes an estimate of the seismic contribution to CDF and LERF based on the methodology detailed in Enclosure 4. Other external hazards are below accepted screening criteria and therefore do not contribute significantly to the totals.

Table E5-1 Waterford 3 CDF and LERF Values Baseline Baseline Hazard Model Reference CDF LERF At Power Internal Events Rev6 3.03E-06 3.04E-08 Reference 5 Internal Fire Rev6 Fire 2.01E-05 2.05E-07 Reference 6 Internal Flood Rev6 Flood .49E-06 7.47E-09 Reference 7 Bounding Seismic 3.2E-06 3.2E-07 Reference 8 Estimate Total 2.78E-05 5.6E-07

W3F1-2021-0003 Page 2 of 2

3. References

1. Letter from Jennifer M. Golder (NRC) to Biff Bradley (NEI), "Final Safety Evaluation for Nuclear Energy Institute (NEI) Topical Report (TR) NEI 06-09, 'Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines,'"

(ADAMS Accession No. ML071200238). dated May 17, 2007.

2. Nuclear Energy Institute (NEI) Topical Report (TR) NEI 06-09, "Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines,"

Revision 0-A, (ADAMS Accession No. ML12286A322). dated October 12, 2012.

3. Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment In Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," Revision 1, (ADAMS Accession No. ML023240437). November 2002.

4. 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, (ADAMS Accession No. ML10091006). May 2011.

5. PSA-WF3-01, Revision 2 - WF3 Internal Events Model R5B Summary Report -

EC71543.

6. PSA-WF3-03-FQ-02 Waterford 3 Fire PRA Quantification Report.

7. PRA-WF3-01-IF-QU, Revision 0 - WF3 Internal Flooding Quantification Report -

EC76543.

8. PSA-WF3-04-01, Seismic Risk Evaluation to Support the TSTF-505 LAR, Rev. 1.

Enclosure 6 W3F1-2021-0003 Justification of Application of At-Power Models to Shutdown Modes This Enclosure is not applicable to the Waterford 3 submittal. Entergy is proposing to apply the Risk-Informed Completion Time Program only in Modes 1 and 2 and not in the shutdown Modes.

Enclosure 7 W3F1-2021-0003 PRA Model Update Process

W3F1-2021-0003 Page 1 of 3 PRA Model Update Process

1. Introduction Section 4.0, Item 8 of the Nuclear Regulatory Commission's (NRC) Final Safety Evaluation (Reference 1) for NEI 06-09, Revision 0-A, "Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines," (Reference 2) requires that the license amendment request (LAR) provide a discussion of the licensee's programs and procedures which assure the PRA models which support the RMTS are maintained consistent with the as-built/as-operated plant.

This enclosure describes the administrative controls and procedural processes applicable to the configuration control of PRA models used to support the Risk-Informed Completion Time (RICT)

Program, which will be in place to ensure that these models reflect the as-built/as-operated plant. Plant changes, including physical modifications and procedure revisions, will be identified and reviewed prior to implementation to determine if they could impact the PRA models per EN-DC-151, PSA Maintenance and Update (Reference 3). The configuration control program will ensure these plant changes are incorporated into the PRA models as appropriate.

The process will include discovered conditions associated with the PRA models, which will be addressed by the applicable site Corrective Action Program.

Should a plant change or a discovered condition be identified that has a significant impact to the RICT Program calculations as defined by the above procedures, an unscheduled update of the PRA model will be implemented. Otherwise, the PRA model change is incorporated into a subsequent periodic model update. Such pending changes are considered when evaluating other changes until they are fully implemented into the PRA models. Periodic updates are typically performed every four years.

2. PRA Model Update Process Internal Event, Internal Flood, and Fire PRA Model Maintenance and Update The Entergy fleet PSA maintenance and update process ensures that the applicable PRA model used for the RICT Program reflects the as-built/as-operated plant for Waterford 3. The PRA configuration control process delineates the responsibilities and guidelines for updating the full power internal events, internal flood, and fire PRA models, and includes both periodic and unscheduled PRA model updates.

The process includes provisions for monitoring potential impact areas affecting the technical elements of the PRA models (e.g., due to plant changes, plant/industry operational experience, or errors or limitations identified in the model), assessing the individual and cumulative risk impact of unincorporated changes, and controlling the model and necessary computer files, including those associated with the configuration risk management program (CRMP) model.

The Entergy PRA maintenance and update processes are governed by procedures. Industry best practices and consensus modeling techniques are also reviewed and monitored to ensure Entergy PRA is using state of the art processes and methods.

W3F1-2021-0003 Page 2 of 3 Review of Plant Changes for Incorporation into the PRA Model

1. Plant changes or discovered conditions are reviewed for potential impact to the PRA models, including the CRMP model and the subsequent risk calculations which support the RICT Program (NEI 06-09, Section 2.3.4, Items 7.2 and 7.3, and 2.3.5, Items 9.2 and 9.3).

2. Plant changes that meet the criteria defined in Reference 3 (including consideration of the cumulative impact of other pending changes) will be incorporated in the applicable PRA model(s), consistent with the NEI 06-09 guidance. Otherwise, the change is assigned a priority and is incorporated at a subsequent periodic update consistent with procedural requirements. (NEI 06-09, Section 2.3.5, Item 9.2)

3. PRA updates for plant changes are typically performed every four years, consistent with the guidance of NEI 06-09 (NEI 06-09, Section 2.3.4, Item 7.1, and 2.3.5, Item 9.1).

4. If a PRA model change is required for the CRMP model, but cannot be immediately implemented for a significant plant change or discovered condition, either:

a. Interim analyses to address the expected risk impact of the change will be performed. In such a case, these interim analyses become part of the RICT Program calculation process until the plant changes are incorporated into the PRA model during the next update. The use of such bounding analyses is consistent with the guidance of NEI 06-09.

b. Appropriate administrative restrictions on the use of the RICT Program for extended Completion Times are put in place until the model changes are completed, consistent with the guidance of NEI 06-09.

These actions satisfy NEI 06-09, Section 2.3.5, Item 9.3.

W3F1-2021-0003 Page 3 of 3

3. References

1. Letter from Jennifer M. Golder (NRC) to Buff Bradley (NEI), "Final Safety Evaluation for Nuclear Energy Institute (NEI) Topical Report (TR) NEI 06-09-A, 'Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS)

Guidelines,'" (ADAMS Accession No. ML071200238). dated May 17, 2007.

2. Nuclear Energy Institute (NEI) Topical Report (TR) NEI 06-09, "Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS)

Guidelines," Revision 0-A, (ADAMS Accession No. ML12286A322). dated October 12, 2012.

3. EN-DC-151, "PSA Maintenance and Update." Entergy Fleet PSA Procedure Governing the process and responsibilities for maintaining and updating the risk models

Enclosure 8 W3F1-2021-0003 Attributes of the CRMP Model

W3F1-2021-0003 Page 1 of 3 Attributes of the CRMP Model

1. Introduction Section 4.0, Item 9 of the Nuclear Regulatory Commission's (NRC) Final Safety Evaluation (Reference 1) for NEI 06-09, Revision 0-A, "Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines," (Reference 2) requires that the license amendment request (LAR) provide a description of PRA models and tools, including identification of how the baseline probabilistic risk assessment (PRA) model is modified for use in the configuration risk management program (CRMP) tools, quality requirements applied to the PRA models and CRMP tools, consistency of calculated results from the PRA model and the CRMP tools, and training and qualification programs applicable to personnel responsible for development and use of the CRMP tools. This item should also confirm that the CRMP tools can be readily applied for each Technical Specification (TS) limiting condition for operation (LCO) within the scope of the plant-specific submittal.

This enclosure describes the necessary changes to the peer-reviewed baseline PRA models for use in the CRMP software to support the Risk-Informed Completion Time (RICT) Program. The process employed to adapt the baseline models for CRMP use is demonstrated:

1. to preserve the core damage frequency (CDF) and large early release frequency (LERF) quantitative results;

2. to maintain the quality of the peer-reviewed PRA models; and

3. to correctly accommodate changes in risk due to configuration-specific considerations.

Quality controls and training programs applicable for the CRMP are also discussed in this enclosure.

2. Translation of Baseline PRA Model Update Process for Use in CRMP The baseline PRA models for internal events, including the internal flood and internal fire models, are the peer-reviewed models, updated when necessary to incorporate plant changes to reflect the as-built/as-operated plant. The internal events, internal fire, and internal flood models are maintained as separate models. These models will be used in the RICT Program. A current effort is in process to merge the models into a single one-top model for easier use in applications.

The CRMP software will be used to facilitate all configuration-specific risk calculations and support the RICT Program implementation. The baseline PRA models are modified as follows for use in configuration risk calculations:

x The unit availability factor is set to 1.0 (unit available).

x Maintenance unavailability is set to zero/false unless unavailable due to the actual (at the time) configuration.

x Mutually exclusive combinations, including normally disallowed maintenance combinations, are adjusted to allow accurate analysis of the configuration.

x For systems where some trains are in service and some in standby the Real Time Risk model addresses the actual configuration of the plant including defining in service trains as needed. Additionally, the impact of outside temperatures on system requirements

W3F1-2021-0003 Page 2 of 3 such as Cooling Tower Fan combination/requirements is addressed in the Real Time Risk model. There are no changes in success criteria based on the time in the core operating cycle.

The configuration risk software is designed to quantify the unit-specific configuration for both internal events, including internal flooding and fire, and includes the seismic risk contribution when calculating the RMAT and RICT.

3. Quality Requirements and Consistency of PRA Model and CRMP Tools The approach for establishing and maintaining the quality of the PRA models, including the CRMP model, includes both a PRA maintenance and update process (described in Enclosure 7), and the use of self-assessments and independent peer reviews (described in Enclosure 2).

The information provided in Enclosure 2 demonstrates that the site's internal event, internal flood, and internal fire PRA models conform to the associated industry standards endorsed by Regulatory Guide 1.200 (Reference 3). This information provides a robust basis for concluding that the PRA models are of sufficient quality for use in risk-informed licensing actions.

For maintenance of an existing CRMP model, changes made to the baseline PRA model in translation to the CRMP model will be controlled and documented. An acceptance test is performed after every CRMP model update. This testing also verifies correct mapping of plant components to the basic events in the configuration risk model. These actions are procedurally controlled.

4. Training and Qualification The PRA staff is responsible for development and maintenance of the configuration risk model. Operations and Work Control staff will use the configuration risk tool under the RICT Program. PRA Staff and Operations are trained in accordance with a program using National Academy for Nuclear Training (ACAD) documents, which is also accredited by INPO.

5. Application of the CRMP Tool to the RICT Program Scope The EPRI PHOENIX software will be used to facilitate all configuration-specific risk calculations and support the RICT Program implementation. This program is specifically designed to support implementation of RMTS. PHOENIX will permit the user to evaluate all configurations within the scope of the RICT Program using appropriate mapping of equipment to PRA basic events. The RICT program will meet RG 1.174 (Reference 4) and Entergy software quality assurance requirements.

W3F1-2021-0003 Page 3 of 3

6. References

1. Letter from Jennifer M. Golder (NRC) to Buff Bradley (NEI), "Final Safety Evaluation for Nuclear Energy Institute (NEI) Topical Report (TR) NEI 06-09-A, 'Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS)

Guidelines,'" (ADAMS Accession No. ML071200238). dated May 17, 2007.

2. Nuclear Energy Institute (NEI) Topical Report (TR) NEI 06-09, "Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS)

Guidelines," Revision 0-A, (ADAMS Accession No. ML12286A322). dated October 12, 2012.

3. Regulatory Guide 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," Revision 2, (ADAMS Accession No. ML090410014). March 2009.

4. 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, (ADAMS Accession No. ML10091006). May 2011.

Enclosure 9 W3F1-2021-0003 Key Assumptions and Sources of Uncertainty

W3F1-2021-0003 Page 1 of 16 Key Assumptions and Sources of Uncertainty

1. Introduction This Attachment examines the sources of uncertainty in the Waterford 3 Probabilistic Risk Assessment (PRA) models and key assumptions in the models. This section documents the PRA technical adequacy of the Waterford 3 PRA model as it applies to PRA applications. In addition to the adequacy of the PRA models with regard to industry peer assessments, modeling uncertainty also needs to be thoroughly examined and understood to support risk-informed applications.

The Waterford 3 PRA at power internal events model uncertainty document (Reference 1),

internal flooding uncertainty document (Reference 2) and fire PRA uncertainty document (Reference 3) each contain a detailed, thorough evaluation of uncertainty for the model of record. Each report follows EPRI/NUREG guidance and considers a variety of potential sources of uncertainty. Key sources of uncertainty are identified, and relevant sensitivity cases are documented to examine the key assumptions and sources of uncertainty and their impact on results.

The purpose of this attachment is to disposition the impact of uncertainty in the PRA models for the Risk Informed Completion Time (RICT) Application. The baseline internal events PRA model, the internal flooding model, and Fire PRA model each document assumptions and sources of uncertainty. These models and the associated documentation have all been reviewed during the model peer reviews. The completeness uncertainty associated with scope and level of detail are documented in the models but are only considered for their impact on a specific application. No specific issues of PRA completeness have been identified in the results of the internal events PRA, internal flood, and fire PRA peer reviews. The approach for this application is, therefore, to review these documents to identify the key assumptions and sources of uncertainty which may be directly relevant to the RICT Program calculations and to identify needed sensitivity analyses where appropriate.

The Waterford 3 PRA models are continuously maintained and periodically updated under the guidance of fleet procedures. This includes continuous identification, documentation and tracking of open issues (the site maintains a Model Change Request (MCR) database for identified model issues). Entergy PRA guidance requires periodic model update as well as self-assessments and peer reviews. Model issues ranging from MCRs to open peer review Findings will be reviewed during RICT efforts to ensure the open issues and the potential impact they may have on the program are understood. This may include additional RICT specific sensitivity cases if they are judged necessary. Note, Waterford 3 does not have a seismic PRA model, but will apply a seismic CDF and LERF penalty for RICT evaluations.

2. Internal Events and Internal Flood Uncertainty The process used to identify at power internal events related PRA model uncertainties and their impact is described in the Waterford 3 PRA uncertainty documentation (Reference 1). The internal flooding PRA model uncertainty document, PSA-WF3-01-IF-SOU (Reference 2),

contains details related to key assumptions and sources of uncertainty relevant to the flooding risk. In the uncertainty documentation, NUREG-1855 (Reference 4) and EPRI report 1016737 (Reference 5) were used to provide guidance for a structured process for addressing

W3F1-2021-0003 Page 2 of 16 uncertainties in PRA inputs and results in the context of risk-informed decision-making.

Appendix A of EPRI 1016737 is used as a template to document plant-specific issue characterization and assessments to fully satisfy the related supporting requirements.

Uncertainty considerations for each model element (success criteria, human actions, data, etc.)

were also documented to ensure a comprehensive evaluation of PRA uncertainty was completed.

The parametric uncertainty analysis for the Waterford 3 at power internal events model is documented in Reference 1, and the Reference 2 contains the parametric flood uncertainty evaluation. These parametric uncertainty analyses address the State-of Knowledge Correlation (SOKC) by the use of system level type codes for basic events. This applies the same variability of all components of that type within a system during the analysis. The parametric uncertainty analyses in the PRA quantification model and documentation demonstrate that the point estimate mean values provide a close representation of the propagated mean values reflecting SOKC and the propagated mean total CDF and LERF values were confirmed to meet RG 1.174 Revision 3 (Reference 6).

Table E9-1 contains key assumptions and sources of uncertainty from the baseline internal events and internal flooding models and assesses the impact of the item(s) on application of a RICT program.

W3F1-2021-0003 Enclosure 9 Page 3 of 16 Table E9-1 Internal Events / Internal Flooding PRA Assumptions & Sources of Uncertainty Assumption/ Uncertainty Discussion RICT Disposition The model has been The uncertainty report This does not represent updated to include credit for documents a sensitivity a key source of uncertainty in the FLEX equipment. The evaluation for the added FLEX RICT application.

added FLEX modeling was changes. The sensitivity case the subject of a sensitivity shows that FLEX has an impact The Revision 6 Waterford 3 evaluation to compare and reduces Station Blackout Uncertainty and Sensitivity results with and without the (SBO) contribution. Analysis contained a sensitivity on added FLEX equipment. While this is documented in the FLEX equipment. This sensitivity uncertainty and sensitivity examined the impact the FLEX documentation, it is not changes had on the model. As considered a source of the current model reflects the as-uncertainty. This was a built, as-operated plant, this is not sensitivity evaluation to gauge considered a source of the impact of a change to the uncertainty.

model. This was a necessary change to accurately update the model to match plant response and the potential use of FLEX equipment and strategies.

Environmental impacts on Local environmental impacts This was noted as a source of initiating events (for can increase or decrease the model uncertainty for the Revision example, intake, offsite frequency of some initiators. An 6 model update but will not have power) alignment for severe weather is an appreciable impact the RICT included for Waterford 3 that application.

addresses the impact of severe weather on both Loss of The environmental impacts can Offsite Power (LOOP) frequency impact the online risk monitor or and recovery. influence changes to IE The other environmental impact frequency. However, RICT risk applicable for Waterford 3 is evaluations will use real time extremely high temperatures. actual plant configuration Waterford 3 has Technical conditions for assessing risk on Specifications that limit plant equipment and systems.

operation with excessive temperature in the ultimate heat sink. Forced shutdowns due to high temperature would be included in the reactor trip data.

Credit for repair and Waterford 3 does not credit No significant impact on the RICT Recovery repair of any equipment. Off- application. Including credit for site equipment recovery would reduce This was noted as a source power recoveries are based on risk results. The current treatment of uncertainty in both the standard practices. is conservative. Conservative Success Criteria and treatment is more likely to result in System Analysis sections. Not crediting equipment

W3F1-2021-0003 Enclosure 9 Page 4 of 16 Table E9-1 Internal Events / Internal Flooding PRA Assumptions & Sources of Uncertainty Assumption/ Uncertainty Discussion RICT Disposition recovery was noted as a source shorter, more restrictive of uncertainty tied to Success completion times.

Criteria. Applying recoveries or completing more detailed off- Lack of credit for potential site power recoveries could equipment recovery can be a lower risk results. source of uncertainty (the model A sensitivity case for EDG not reflecting the real response to recovery is included in the an event). However, in this case quantification/sensitivity the results bias in the direction of documentation. higher risk and conservativism.

Several bounding inputs Bounding conditions for the Since the treatment of induced-and conditions for LERF Waterford 3 LERF model SGTR and scrubbing are primarily analysis were identified as include a phenomenological uncertainty potential sources of induced SGTRs and credit for issue for LERF, this will not have uncertainty scrubbing. an appreciable impact on the

- Use of generic value Several sensitivity evaluations RICT program. Additionally, due for SG age/wear, were included in the LERF to the fact that LERF is dominated

- In-vessel core melt documentation to evaluate the by vessel rupture (~90%), the impact on TI-SGTR, impact the bounding and results are not sensitive to these and potentially uncertain inputs have considerations.

- Credit for scrubbing on LERF results. All noted issues are treated For example, average SG wear conservatively in the LERF model models were used even as and are not expected to uniquely Waterford 3 recently replaced impact RICT results.

S/Gs.

Sensitivity evaluations are documented in the LERF analysis for the impact of credited scrubbing, as well as in-vessel core melt on TI-SGTR Human actions credited The Waterford 3 LERF model Treatment of post core damage during severe accident contains 3 operator actions after human actions will not have an conditions core damage: bump the pump, appreciable impact on the RICT late depressurization, and program given the fact that LERF LOOP recovery. results are dominated by vessel Due to the uncertain nature of rupture (~90%) and not sensitive post-accident/post core damage to these operator actions.

conditions, credit for these actions was identified as a source of uncertainty. A sensitivity case is included in the documentation with no post core damage human actions credited.

W3F1-2021-0003 Enclosure 9 Page 5 of 16 Table E9-1 Internal Events / Internal Flooding PRA Assumptions & Sources of Uncertainty Assumption/ Uncertainty Discussion RICT Disposition Maintenance/operational Waterford 3 includes numerous This was noted in the PRA activities (for example, alignment flags that can revision 6 documentation as a switchyard work, system determine the impact of potential uncertainty candidate.

testing) maintenance or operational However, RICT risk evaluations activities on risk. In addition, will use real time plant the risk associated with different configuration inputs when types of switchyard work can assessing risk on equipment and also be evaluated. systems.

Model may over-estimate Conservative criteria in LERF No significant impact on a RICT contribution of pressure modeling may cause over application. Conservative induced SGTR (PI-SGTR) estimation of LERF risk. treatment will result in shorter, to LERF. In modeling, induced SGTR - more limiting completion times.

bounding conditions were applied.

The PI-SGTR and TI-SGTR values used the average wear models for the steam generator but the steam generators were recently replaced (Waterford 3s new SGs have less wear than industry average).

The Following Entries are Related to Internal Flooding The information shown on Different pipe lengths would No significant impact on the RICT the walkdown data sheets directly impact the initiating application. Risk significant in the original internal event frequencies which in turn scenarios were checked for flooding analysis was would affect the core damage errors/discrepancies between assumed to be correct frequency. For risk-significant actual and documented pipe unless changed in the Fire flooding events, length data was lengths. Variations in pipe length PRA Walkdown Notebook. based on drawing information or inputs are in the conservative The legacy information in confirmed during walkdowns. direction and leading conservative the initial flooding walkdown Therefore, the overall effect on impact on the RICT results.

was spot-checked to verify the base model is expected to accuracy. be small.

For areas where fluid drains This assumption is reasonable This was noted in the document away quickly, sealed and backed by regulatory as a key assumption, but no penetrations are assumed requirements. It is also an sensitivity cases were developed to be effective at preventing industry consensus approach. as the treatment was judged to be propagation between areas Penetrations/barriers with a acceptable and not a source of such that the propagation specific concern would be significant uncertainty.

would not directly impinge addressed on an individual on equipment in the basis. Additionally, WF3 applied industry adjoining area and result in consensus methods to help limit equipment failure. For potential uncertainty.

W3F1-2021-0003 Enclosure 9 Page 6 of 16 Table E9-1 Internal Events / Internal Flooding PRA Assumptions & Sources of Uncertainty Assumption/ Uncertainty Discussion RICT Disposition example, the penetration This is a flood specific issue and seals will prevent spray and will not significantly impact the splash impacts from application of a RICT program.

causing equipment failures in the adjoining area.

Generic Letter 86-10 specifies that fire barriers must be capable of withstanding significant spray and splash after being exposed to fire. For areas where fluids drain away quickly, use of fire barriers is assumed to prevent propagation for the time needed for the fluid to drain away.

Flood-induced failure of Due to multiple potential flood- No significant impact on the RICT AOVs involves the valve induced failure modes (e.g., application. Conservative operators loss of function valve shorts out, instrument air treatment will result in shorter, but would also involve the supply line fails, water affects more limiting completion times.

AOV failing to its designed valves diaphragm, etc.), it is not fail position. However, if the guaranteed that AOV will always fail-safe position is the fail to its fail-safe position.

desired position for the Therefore, credit for fail-safe is PRA, it is assumed that the not taken. This assumption is valve fails as-is. conservative, but reasonable.

The overall effect on the base model is expected to be small.

It is considered improbable The selection of 25 gpm is No significant impact on the RICT that flooding events with based on engineering application. Conservative small flow rates will spray judgement, it is about the same treatment will result in shorter, enough PRA related flow rate as three or four more limiting completion times.

equipment housed in the household garden hoses and is affected zone to initiate a considered reasonable.

PRA transient. Therefore, the smallest discharge rate This was identified in as a key from any break is assumed assumption. This assumption is to be 25 gpm unless noted reasonable and expected to be otherwise [for specified conservative. The impact on spray scenarios]. the base model is expected to be small.

W3F1-2021-0003 Page 7 of 16

3. Fire PRA (FPRA) Model:

The process used to identify uncertainties and their impact is described for the fire PRA in the Waterford 3 FPRA uncertainty documentation (Reference 3).

The evaluation examines sources of uncertainty for each of the FPRA development follows NUREG-1855 (Reference 4). The Fire PRA Uncertainty Report contains discussions on topics outlined in from EPRI report 1026511 Appendix B (Reference 7) arranged by tasks from NUREG/CR-6850 (Reference 8). Key sources of uncertainty are noted, and sensitivity cases are completed to evaluate them.

The WF3 Fire PRA Sensitivity and Uncertainty Report is a thorough and comprehensive assessment of uncertainty. The report evaluates over seventy uncertainty topics and discusses/documents how each topic is addressed in the Waterford 3 Fire PRA model.

The parametric uncertainty analysis for the Waterford 3 Fire PRA is provided in the uncertainty documentation. The parametric uncertainty analysis addresses the State-of-Knowledge Correlation (SOKC) by the use of system level type codes for basic events, and additional SOKC factors included in quantitation. This applies the same variability of all components of that type within a system during the analysis. The parametric uncertainty analyses in the PRA quantification model/model documentation demonstrate that the point estimate mean values provide a close representation of the propagated mean values reflecting SOKC and the propagated mean total CDF and LERF values were confirmed to meet RG 1.174 Revision 3 (Reference 6).

The Fire PRA uncertainty document identifies key assumptions and sources of uncertainty relevant to development and quantification of the fire models. The items identified in that document are included in Table 6-2 below. The table includes items judged to be key assumptions and sources of uncertainty relevant to the fire PRA. Table 6-2 assess whether the key fire PRA assumptions and uncertainty topics are specifically relevant to the use of that model for the RICT application.

As noted in the uncertainty/sensitivity documentation, the WF3 Fire PRA analysis is believed to represent a somewhat conservative estimation of fire risk, within the constraints of the requirements for a model acceptable for the NFPA-805 program. As the model is somewhat conservative, its application for a RICT program will likely slightly bias results toward shorter completion times. The evaluation of sources of uncertainty in the FPRA are documented in the table below including consideration for impact on a RICT application.

The Fire PRA uncertainty document identifies key assumptions and sources of uncertainty relevant to development and quantification of the fire models. The items identified in that document are included in Table 6-2 below. The table includes items judged to be key assumptions and sources of uncertainty relevant to the fire PRA. Table E9-2 assess whether the key fire PRA assumptions and uncertainty topics are specifically relevant to the use of that model for the RICT application.

W3F1-2021-0003 Enclosure 9 Page 8 of 16 Table E9-2 Fire PRA Sources of Model Uncertainty Uncertainty Topic Sources of Uncertainty RICT Disposition Scope of equipment Lack of credit for some systems The TEDG is credited in the PRA credited in fire PRA could mask the risk associated with model. This is represented in the model those systems in some model and results and represents applications. Additionally, that the as-built-as-operated plant.

same lack of credit could overestimate the importance of Besides the EDG/TEDG related other credited systems. systems, this topic is not a source A sensitivity case was completed to of uncertainty for the RICT program evaluate this. The Temporary (the sensitivity case resulted in less Emergency Diesel Generator than 1% change in fire CDF and (TEDG) credit was adjusted. The LERF).

TEDG was made permanent during a plant modification and was added to the PRA model in the last internal events update. There is some uncertainty with regard to how quickly the TEDG would be used in the event of a fire. For this sensitivity, it is assumed the TEDG is not available.

Exclusion of certain Lack of credit for some systems The current approach used systems due to lack of (systems with limited cable data - (assume equipment lacking cable data all assumed failed) could mask the detailed cable data is failed) will risk associated with those systems result in conservative evaluations.

in some applications. Additionally, This conservativism will tend to that same lack of credit could result in shorter, more limiting overestimate the importance of completion times.

other credited systems.

A sensitivity case was completed to evaluate this uncertainty. The model was evaluated with credit for Unlocated (assumed failed)

Equipment. In the case run, the same equipment was assumed available in all locations except Turbine Generator Building (TGB).

Development of fire Present NUREG/CR-6850 results This is a fire frequency specific frequencies for each in different fire frequencies for the issue. This will not have an fire area and ignition same equipment in different plants. appreciable impact on the RICT source For example, older BWRs with less evaluations or results.

equipment than a new PWR may result in a factor of 2 higher fire frequencies for pumps or electrical

W3F1-2021-0003 Enclosure 9 Page 9 of 16 Table E9-2 Fire PRA Sources of Model Uncertainty Uncertainty Topic Sources of Uncertainty RICT Disposition equipment. This is a form of parameter uncertainty. The Waterford 3 FPRA uses the ignition frequencies from EPRI Supplement 1 to NUREG/CR-6850. The source of data for the ignition frequencies is a source of uncertainty.

A sensitivity case was completed to evaluate this uncertainty. The model was run with ignition frequencies from different sources (NUREG/CR-6850 vs. EPRI Supplement 1).

Credit for fire wrap No credit is taken for fire wrap The inclusion/exclusion of fire wrap (qualified 3M wrap) in the has a negligible impact on the Waterford 3 FPRA. overall results. Due to the small A sensitivity case was completed to impact demonstrated by sensitivity evaluate this uncertainty. case, the uncertainty sensitivity Credit for Wrap in Risk Significant cases associated with this model locations where wrap exists but is uncertainty is negligible and not not credited was evaluated (it relevant for the RICT application.

assumed the wrap prevents failure of FPRA targets that are wrapped).

Treatment of In the Waterford 3 FPRA cable The sensitivity cases were permissive signals, selection, the majority of the circuit performed to measure the risk interlocks, and analysis is performed using associated with the treatment of associated logic detailed and conservative safe these cables that in select fire shutdown analysis. Any additional scenarios. Although the sensitivity cable selection is performed in a shows a small impact on Fire CDF similar manner. Automatic and Fire LERF, the uncertainty actuation logic signals are modeled document concluded that the as separate pseudo-components modeling treatment applied to with their own cable selection. these cables is acceptable and Thus, associated circuits are preferred to alternative modeling accounted for in the cable treatments.

selection. Some component This will not have an appreciable specific changes were made in fire impact on the RICT program.

modeling/quantification (with documented basis for each decision). Due to the special treatment applied to certain signal cables. This treatment was

W3F1-2021-0003 Enclosure 9 Page 10 of 16 Table E9-2 Fire PRA Sources of Model Uncertainty Uncertainty Topic Sources of Uncertainty RICT Disposition identified as a source of modeling uncertainty.

Two sensitivity cases were documented to examine this uncertainty. In one case, the cables with special treatment were assumed to always fail. In the other case, the same cables were assumed to never fail.

Circuit failure Failure probabilities utilized in The use of hot short failure probabilities circuit failure analysis could be a probability given fire induced failure form of parameter uncertainty, but probability is based on fire test data the choice of the representative set and associated consensus of values is a form of model methodology published in NUREG-uncertainty. The Waterford 3 7150, Volume 2. Based on a FPRA utilizes circuit failure review of the assumptions and probabilities from NUREG/CR- potential uncertainty related to this 7150. The aggregate failure element, it is concluded that the probabilities are used, which methodology for the Circuit Failure combines intra and inter cable Mode Likelihood Analysis task faults, and is specific to the type of does not introduce any epistemic circuit and material type, e.g., uncertainties that will affect a RICT single break control circuit, program.

thermoset cables, ungrounded DC Solenoid Operated Valves (SOVs).

A sensitivity case was completed to examine this uncertainty. In the sensitivity, all Circuit Failure Mode Likelihood (CFMLA) values were set to 1.0, i.e., no credit for hot short probabilities.

Availability of power In the Waterford 3 FPRA, a This is a fire PRA specific issue for spurious operations simplifying assumption is made and is limited to the RCPs. The after initial cable failure that the power is available to a treatment of this issue adds cable damaged by a fire and thus additional detail to the model such can spuriously operate, when in that this issue will not have an fact the fire damage may cause the appreciable impact on the RICT power supply to be interrupted. evaluations or results.

Power is assumed to be initially available to allow spurious operation to occur. This is a conservative assumption. An

W3F1-2021-0003 Enclosure 9 Page 11 of 16 Table E9-2 Fire PRA Sources of Model Uncertainty Uncertainty Topic Sources of Uncertainty RICT Disposition exception to this modeling assumption is in the Turbine Building, if power is lost to the 6.9kV switchgears, the RCPs are assumed to be tripped and cannot spuriously start or fail to be tripped by the operators, despite a spurious failure of the RCP control cables.

A sensitivity case was completed to examine this uncertainty. The case removes this treatment for the RCPs (they can spuriously start/operate with loss of 6.9kV power).

Credit for detection In the Waterford 3 FPRA, manual This is a fire detection/suppression and suppression suppression is not credited for oil specific issue. The treatment of fire scenarios, since it was this issue adds additional detail to determined that oil fires damage the model such that this issue will targets quickly, and it is assumed not have an appreciable impact on that target damage occurs prior to the RICT evaluations or results.

the fire brigade responding. For electrical and cable fires, a bounding time to damage is determined that is applicable to all such fire scenarios, and this time to damage is used to establish time available to credit manual suppression utilizing generic non-suppression probabilities, as provided in FAQ 08-0050 in Supplement 1 of NUREG/CR-6850.

Several Sensitivity cases were run to examine the uncertainty and model sensitivity to credited detection and suppression. The following cases were completed.

- Assume failure of all automatic suppression systems (no credit for automatic suppression.

W3F1-2021-0003 Enclosure 9 Page 12 of 16 Table E9-2 Fire PRA Sources of Model Uncertainty Uncertainty Topic Sources of Uncertainty RICT Disposition

- Credit automatic suppression in the general areas of the Turbine Building, PAU TGB

- A case that assumes failure of all automatic detection systems and thus manual suppression in all areas that are not continuously manned, such as the MCR.

- A case was run to examine manual detection/suppression.

The case assumed failure of all manual suppression by the dedicated fire watch in hot work scenarios, as well as manual suppression in the MCR and other continuously manned locations.

Effectiveness of The fire barrier failure probabilities This is a fire barrier specific issue.

passive fire barriers are based on the partitioning Consensus methods are used such between compartments elements in the barrier. In the that This will not have appreciable Waterford 3 FPRA, the failure impact on the RICT program probabilities for passive fire barrier evaluations.

partitioning elements are based on the generic values listed in NUREG/CR-6850 Appendix A.

The barrier failure probability for the barrier is the sum of the partitioning elements applicable to that barrier.

A couple of sensitivity cases were run to examine uncertainty associated with credited barrier failure probabilities. In one case it was assumed all barriers fail with a probability of 1.0 (i.e., they are not effective due to a door being propped open, a penetration not being filled after maintenance, etc.).

In a second case, barrier failure probabilities were reduced. It was

W3F1-2021-0003 Enclosure 9 Page 13 of 16 Table E9-2 Fire PRA Sources of Model Uncertainty Uncertainty Topic Sources of Uncertainty RICT Disposition assumed all barriers are better by a factor of 10 (i.e., they are less conservatively modeled, unless the failure is currently 1.0).

Treatment of structural The potential impacts of structural This is a fire specific issue steel failures steel failures were considered and associated with large fires in the assessed at Waterford 3, and it Turbine Building. This will not have was determined that the TGB is the appreciable impact on the RICT only PAU noted to contain potential program. It is also a conservative high-hazard fire sources. All oil fire treatment and any impacts would scenarios in the TGB were slightly bias completion times evaluated, and those that could resulting in shorter, more limiting possibly impact a structural results.

member were assumed to fail the entire TGB outside the Switchgear Room. This assumption could contain a large amount of conservatism.

A sensitivity case was completed to examine this uncertainty.

A case was run to reduce conservatism with structural collapse fires in the turbine building. Some hydrogen fires and oil fires were adjusted to only impact the source of the fire and nearby targets, but not collapse the Turbine Building.

Treatment of fire- For all operator actions credited in The Waterford 3 Fire PRA model is induced instrument the Waterford 3 FPRA, there is a based on industry consensus failures discussion of the cues and/or modeling approaches for its HEP required instrumentation. Some calculations, so this is not actions model the required considered a significant source of instrumentation. For other operator uncertainty.

actions, redundancy and diversity of instruments is credited as a means to assume that sufficient indication exists. In this case, the assumption that it does not need to be modeled becomes a source of model uncertainty.

W3F1-2021-0003 Enclosure 9 Page 14 of 16 Table E9-2 Fire PRA Sources of Model Uncertainty Uncertainty Topic Sources of Uncertainty RICT Disposition A sensitivity case was completed to examine this uncertainty.

Indication for the Condensate Storage Pool (CSP) Makeup or Alternate EFW Suction are among the most important indications for operators. This is because the system window is several hours long, it is assumed alternate indications for CSP level could be used as an operator cue; however, if alternate indication is not available or provides false readings, these actions would be hindered. This sensitivity case assumes failure probabilities of CSP makeup and alternate CSP suction are increased by factor of 10 (this includes relevant combination HFEs).

HEP Methodology The basis for the HEP The Waterford 3 Fire PRA model, methodology utilized needs to be including HEP methodology, is consistent with the internal events based on industry consensus PRA standard requirements for modeling approaches for its HEP HRA. Model uncertainty exists on calculations, so this is not the actual methodology utilized; considered a significant source of this is recognized as a generic uncertainty.

source of model uncertainty. The Waterford 3 FPRA utilized EPRI Several HEP sensitivities have HRAC per the guidance in been completed and documented.

NUREG-1921 (Reference 9) in the No HEP specific issues are development of fire-specific HEP expected to significantly impact values. application of an RICP program.

Several Fire HEP related sensitivity cases were run to examine relevant uncertainties. The cases included:

A case examining - Manual Control of EFW Flow (Long term ex-MCR actions are assumed to have the same failure probability as the

W3F1-2021-0003 Page 15 of 16 Table E9-2 Fire PRA Sources of Model Uncertainty Uncertainty Topic Sources of Uncertainty RICT Disposition internal events model; however, if these actions would be hindered by the fire in some way, the failure probabilities would be higher; assume failure probabilities of these actions are increased by a factor of 10, including relevant combination HFEs.)

A case examining - Manual Actuation of EFAS if Auto-Actuation Fails (This is the most risk-significant single operator action; this is a simple action that is assumed to be a factor of 10 worse than the internal events probability, per the guidance in NUREG-1921; however, since this is the most important action and is a very simple action. The case assumes that the probability is equal to that calculated for internal events, and reduce the relevant action and combination HFEs by a factor of 10.)

A case examining - Tripping of the RCPs Following Loss of Seal Cooling (Combination of tripping from the MCR, which is assumed to be a factor of 10 worse than the internal events value, and locally tripping the breaker in TGB, which has a detailed fire-specific calculation). This action is known by the operators to be an important action and is very simple. This case reduces the failures associated with tripping the RCPs (and relevant HFE combination events) by a factor of 10.

W3F1-2021-0003 Page 16 of 16 4.0 References

1. Entergy Report, PSA-WF3-01-QU-01, Revision 2 , "WF3 PSA Uncertainty and Sensitivity Analysis," dated February 27, 2019

2. Entergy Report, PSA-WF3-01-IF-SOU, "Waterford 3 Internal Flooding Sources of Uncertainty, dated May 6, 2020

3. Entergy Report, PSA-WF3-03-UNC-01, "WF3 Fire PRA Sensitivity and Uncertainty Report," dated December 23, 2019

4. NRC Report NUREG-1855, "Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decision Making," Revision 1, (ADAMS Accession No. ML17062A466), dated March 2017

5. EPRI Technical Report 1016737, "Treatment of Parameter and Modeling Uncertainty for Probabilistic Risk Assessments," dated December 19, 2008

6. NRC 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, (ADAMS Accession No. ML17317A256), dated January 2018

7. EPRI Technical Update 1026511, "Practical Guidance on the use of Probabilistic Risk Assessment in Risk-Informed Applications with a Focus on the Treatment of Uncertainty," dated December 4, 2012

8. NRC Report NUREG/CR-6850, EPRI 1011989, "EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities," (ADAMS Accession Nos. ML052580075, ML052580118),

dated September 2005

9. NRC Report NUREG-1921, "EPRI/NRC-RES Fire Human Reliability Analysis Guidelines," (ADAMS Accession No. ML12216A104), dated July 2012

Enclosure 10 W3F1-2021-0003 Program Implementation

W3F1-2021-0003 0 Page 1 of 3 Program Implementation

1. Introduction Section 4.0, Item 11 of the NRC Final Safety Evaluation (Reference 1) for NEI 06-09, Revision 0-A (Reference 2) requires that the license amendment request (LAR) provide a description of the implementing programs and procedures regarding the plant staff responsibilities for the Risk Managed Technical Specifications (RMTS) implementation, and specifically discuss the decision process for risk management action (RMA) implementation during a Risk-Informed Completion Time (RICT).

This enclosure provides a description of the implementing programs and procedures regarding the plant staff responsibilities for the Risk-Informed Completion Time (RICT) Program, including training of plant personnel, and specifically discusses the decision process for RMA implementation during extended Completion Times (CT).

2. RICT Program and Procedures Entergy will develop a program description and implementing procedures for the RICT Program.

The program description will establish the management responsibilities and general requirements for risk management, training, implementation, and monitoring of the RICT program. More detailed procedures will provide specific responsibilities, limitations, and instructions for implementing the RICT program. The program description and implementing procedures will incorporate the programmatic requirements for RMTS included in NEI 06-09-A. The program will be integrated with the online work control process. The work control process currently identifies the need to enter an LCO Action statement as part of the planning process and will additionally identify whether the provisions of the RICT program are required for the planned work. The risk thresholds associated with 10 CFR 50.65(a)(4) will be coordinated with the RICT limits. The maintenance rule performance monitoring provisions and Mitigating System Performance Index (MSPI) thresholds will assist in controlling the amount of risk expended in use of the RICT program.

The Operations Department (licensed operators) is responsible for compliance with the TS and will be responsible for implementation of RICTs and RMAs. Entry into the RICT program will require management approval prior to pre-planned activities and as soon as practicable following emergent conditions.

The procedures for the RICT program will address the following attributes consistent with NEI 06-09-A:

x Plant management positions with authority to approve entry into the RICT Program.

x Important definitions related to the RICT Program.

x Departmental and position responsibilities for activities in the RICT Program.

x Plant conditions for which the RICT Program is applicable.

x Limitations on implementing RICTs under voluntary and emergent conditions.

x Implementation of the RICT Program 30-day back stop limit.

x Use of the Configuration Risk Management Program (CRMP) tool.

x Guidance on recalculating RICT and risk management action time (RMAT) within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or within the most limiting front-stop CT after a plant configuration change.

W3F1-2021-0003 0 Page 2 of 3 x Requirements to identify and implement RMAs when the RMAT is exceeded or is anticipated to be exceeded.

x Guidance on the use of RMAs including the conditions under which they may be credited in RICT calculations.

x Guidance on crediting probabilistic risk assessment (PRA) functionality.

x Conditions for exiting a RICT.

x Requirements for training on the RICT Program.

x Documentation requirements related to individual RICT evaluations, implementation of extended CTs, and accumulated annual risk.

3. RICT Program Training Training will be carried out in accordance with Entergy training procedures and processes that utilize the Systematic Approach to Training. These procedures were written based on the Institute of Nuclear Power Operations (INPO) Accreditation (ACAD) requirements, as developed and maintained by the National Academy for Nuclear Training.

Participation Departments that will receive training appropriate to their level of program responsibilities include:

x Operations x Operations Training x Work Management x Outage Management x Planning and Scheduling Personnel x Work Week Managers x Regulatory Assurance x Maintenance x Engineering x Risk Management x Other Selected Management Scope of Training For those individuals who will be directly involved in the implementation of the RICT Program the training topics will be developed with consideration for:

x Specific training on the revised TS x Record keeping requirements x Case studies x Hands-on experience with the CRMP tool for calculating RMAT and RICT x Identifying appropriate RMAs x Determining PRA functionality x Common cause failure considerations x Other detailed aspects of the RICT Program For management positions with authority to approve entry into the RICT Program, as well as supervisors, managers, and other personnel who will closely support RICT implementation, the

W3F1-2021-0003 0 Page 3 of 3 training will provide a broad understanding of the purpose, concepts, and limitations of the RICT Program.

4. References

1. Letter from Jennifer M. Golder (NRC) to Buff Bradley (NEI), "Final Safety Evaluation for Nuclear Energy Institute (NEI) Topical Report (TR) NEI 06-09, 'Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines,'"

(ADAMS Accession No. ML071200238). dated May 17, 2007.

2. Nuclear Energy Institute (NEI) Topical Report (TR) NEI 06-09, "Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines,"

Revision 0-A, (ADAMS Accession No. ML12286A322). dated October 12, 2012.

Enclosure 11 W3F1-2021-0003 Monitoring Program

W3F1-2021-0003 1 Page 1 of 2 Monitoring Program

1. Introduction Section 4.0, Item 12 of the NRC Final Safety Evaluation (Reference 1) for NEI 06-09 Revision 0-A (Reference 2) requires that the license amendment request (LAR) provide a description of the implementation and monitoring program as described in Regulatory Guide (RG) 1.174, An Approach For Using Probabilistic Risk Assessment In Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, Revision 1, (Reference 3) and NEI 06-09-A. (Note that RG 1.174, Revision 2 (Reference 4), issued by the NRC in May 2011, made editorial changes to the applicable section referenced in the NRC safety evaluation for Section 4.0, Item 12.)

This enclosure provides a description of the process applied to monitor the cumulative risk impact of implementation of the Risk-Informed Completion Time (RICT) Program, specifically the calculation of cumulative risk of extended Completion Times (CTs). Calculation of the cumulative risk for the RICT Program is discussed in Step 14 of Section 2.3.1 and Step 7.1 of Section 2.3.2 of NEI 06-09, Risk Informed Technical Specifications Initiative 4b.

General requirements for a Performance Monitoring Program for risk-informed applications are discussed in Regulatory Guide (RG) 1.174, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant Specific Changes to the Licensing Basis, Element 3.

2. Description of Monitoring Program The RICT Program will require calculation of cumulative risk impact at least every refueling cycle, not to exceed 24 months, consistent with the guidance in NEI 06-09, Revision 0-A. For the assessment period under evaluation, data will be collected for the risk increase associated with each application of an extended CT for both core damage frequency (CDF) and large early release frequency (LERF), and the total risk will be calculated by summing all risk associated with each RICT application. This summation is the change in CDF or LERF above the zero maintenance baseline levels during the period of operation in the extended CT (i.e., beyond the front-stop CT). The change in risk will be converted to average annual values.

The total average annual change in risk for extended CTs will be compared to the guidance of RG 1.174, Revision 2, Figures 4 and 5, acceptance guidelines for CDF and LERF, respectively. If the actual annual risk increase is acceptable (i.e., not in Region I of Figures 4 and 5 of RG 1.174, Revision 2), then RICT Program implementation is acceptable for the assessment period. Otherwise, further assessment of the cause of exceeding the acceptance guidelines of RG 1.174 and implementation of any necessary corrective actions to ensure future plant operation is within the guidelines will be conducted under the corrective action program.

The evaluation of cumulative risk will also identify areas for consideration, such as:

x RICT applications that dominated the risk increase x Risk contributions from planned vs. emergent RICT applications x Risk Management Actions (RMA) implemented but not credited in the risk calculations x Risk impact from applying RICT to avoid multiple shorter duration outages

W3F1-2021-0003 1 Page 2 of 2 x Any specific RICT application that incurred a large proportion of the risk Based on a review of the considerations above, corrective actions will be developed and implemented as appropriate. These actions may include:

x Administrative restrictions on the use of RICTs for specific high-risk configurations x Additional RMAs for specific configurations x Rescheduling planned maintenance activities x Deferring planned maintenance to shutdown conditions x Use of temporary equipment to replace out-of-service systems, structures, or components (SSC) x Plant modifications to reduce risk impact of future planned maintenance configurations In addition to impacting cumulative risk, implementation of the RICT Program may potentially impact the unavailability of SSCs. The existing Maintenance Rule (MR) monitoring programs under 10 CFR 50.65(a)(1) and (a)(2) provide for evaluation and disposition of unavailability impacts which may be incurred from implementation of the RICT Program. The SSCs in the scope of the RICT Program are also in the scope of the MR, which allows the use of the MR Program. RG 1.177, An Approach for Plant-Specific, Risk-Informed Decision Making: Technical Specifications (Reference 5), Section 3.2, Maintenance Rule Control, discusses that the scope of evaluations required under the Maintenance Rule should include prior related TS changes, such as extension of CTs.

The monitoring program for the MR, along with the specific assessment of cumulative risk impact described above, serve as the Implementation and Monitoring Program for the RICT Program as described in Element 3 of RG 1.174 and NEI 06-09.

3. References

1. Letter from Jennifer M. Golder (NRC) to Buff Bradley (NEI), "Final Safety Evaluation for Nuclear Energy Institute (NEI) Topical Report (TR) NEI 06-09-A, 'Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS)

Guidelines,'" (ADAMS Accession No. ML07I1200238). dated May 17, 2007.

2. Nuclear Energy Institute (NEI) Topical Report (TR) NEI 06-09, "Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines,"

Revision 0-A, (ADAMS Accession No. ML12286A322). dated October 12, 2012.

3. Regulatory Guide 1.174, "An Approach For Using Probabilistic Risk Assessment In Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," Revision 1, (ADAMS Accession No. ML023240437). November 2002.

4. 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, (ADAMS Accession No. ML10091006). May 2011.

5. Regulatory Guide 1.177, "An Approach for Plant-Specific, Risk-Informed Decision Making: Technical Specifications," Revision 1, May 2011.

Enclosure 12 W3F1-2021-0003 Risk Management Action Examples

W3F1-2021-0003 2 Page 1 of 3 Risk Management Action Examples

1. Introduction This enclosure describes the process for identification and implementation of Risk Management Actions (RMA) applicable during extended Completion Times (CT) and provides examples of RMAs. RMAs will be governed by plant procedures for planning and scheduling maintenance activities. The procedures will provide guidance for the determination and implementation of RMAs when entering the Risk-Informed Completion Time (RICT) Program consistent with the guidance provided in NEI 06-09, Revision 0-A (Reference 1).

2. Responsibilities For planned entries into the RICT Program, Work Management is responsible for developing the RMAs with assistance from Operations and Risk Management. Operations is responsible for approval and implementation of RMAs. For emergent entry into extended CTs, Operations is also responsible for developing the RMAs.

3. Procedural Guidance For planned maintenance activities, implementation of RMAs will be required if it is anticipated that the risk management action time (RMAT) will be exceeded. For emergent activities, RMAs must be implemented if the RMAT is reached. Also, if an emergent event occurs requiring recalculation of a RMAT already in place, the procedure will require a reevaluation of the existing RMAs for the new plant configuration to determine if new RMAs are appropriate. These requirements of the RICT Program are consistent with the guidance of NEI 06-09.

RMAs will be implemented in accordance with current procedures (Entergy RICT/RMAT procedures are not yet developed) no later than the time at which an incremental core damage probability (ICDP) of 1E-6 is reached, or no later than the time when an incremental large early release probability (ILERP) of 1E-7 is reached. If, as the result of an emergent condition, the instantaneous core damage frequency (ICDF) or the instantaneous large early release frequency (ILERF) exceeds 1E-3 per year or 1E4 per year, respectively, RMAs are also required to be implemented. These requirements are consistent with the guidelines of NEI 06-09.

By determining which structures, systems, or components (SSCs) are most important from a CDF or LERF perspective for a specific plant configuration, RMAs may be created to protect these SSCs. Similarly, knowledge of the initiating event or sequence contribution to the configuration-specific CDF or LERF allows development of RMAs that enhance the capability to mitigate such events. The guidance in NUREG-1855 (Reference 2) and EPRI TR-1026511 (Reference 3) will be used in examining PRA results for significant contributors for the configuration, to aid in identifying appropriate compensatory measures (e.g., related to risk significant systems that may provide diverse protection, or important support systems or human actions). Enclosure 9 identifies several areas of uncertainty in the internal events and fire PRAs that will be considered in defining configuration-specific RMAs when entering a RICT.

If the planned activity or emergent condition includes an SSC that is identified to impact Fire PRA, as identified in the current Configuration Risk Management Program (CRMP), Fire PRA

W3F1-2021-0003 2 Page 2 of 3 specific RMAs associated with that SSC shall be implemented per the current plant procedure.

Approved equipment-specific RMAs for risk significant SSCs within the scope of the RICT program will be contained in the procedure.

It is possible to credit RMAs in RICT calculations, to the extent the associated plant equipment and operator actions are modeled in the PRA; however, such quantification of RMAs is neither required nor expected by NEI 06-09. Nonetheless, if RMAs will be credited to determine RICTs, the procedure instructions will be consistent with the guidance in NEI 06-09.

NEI 06-09 classifies RMAs into the three categories described below:

1) Actions to increase risk awareness and control.

x Shift brief x Pre-job brief x Training x Presence of system engineer or other expertise related to the activity x Special purpose procedure to identify risk sources and contingency plans

2) Actions to reduce the duration of maintenance activities.

x Pre-staging materials x Conducting training on mock-ups x Performing the activity around the clock x Performing walk-downs on the actual system(s) to be worked on prior to beginning work

3) Actions to minimize the magnitude of the risk increase.

x Suspend or minimize activities on redundant systems x Suspend or minimize activities on other systems that adversely affect the CDF or LERF x Suspend or minimize activities on systems that may cause a trip or transient to minimize the likelihood of an initiating event that the out-of-service component is meant to mitigate x Use temporary equipment to provide backup power, ventilation, etc.

x Reschedule other risk-significant activities

4. Examples Example RMAs that may be considered during a RICT Program entry for an Emergency Diesel Generator (EDG) or a Containment Spray (CS) Pump to reduce the risk impact and ensure adequate defense-in-depth are:

A. Diesel Generator:

1. Contact the Transmission System Operator (TSO) to determine the reliability of offsite power supplies prior to entering a RICT, and implement RMAs during times of high grid stress conditions, such as during high demand conditions.

2. Evaluate weather conditions for threats to the reliability of offsite power supplies.

W3F1-2021-0003 2 Page 3 of 3

3. Defer elective maintenance in the switchyard, on the station electrical distribution systems, and on the main and auxiliary transformers associated with the unit.

4. Defer planned maintenance or testing that affects the reliability of the operable EDGs and their associated support equipment. Defer planned maintenance activities on station blackout mitigating systems. Treat these as protected equipment.

5. Verify availability of TEDG.

6. Defer planned maintenance or testing on redundant train safety systems. If testing or maintenance activities must be performed, a review of the potential risk impact will be performed.

7. Implement 10 CFR 50.65(a)(4) fire-specific RMAs associated with the affected EDG.

8. Brief the on-shift operations crew concerning the unit activities, including any compensatory measures established, and review the appropriate emergency operating procedures for a Loss of Offsite Power.

B. Containment Spray Pump:

1. Defer planned maintenance or testing activities on the redundant CS train and its associated support equipment and treat those systems as protected equipment.

2. Defer planned maintenance or testing that affects the reliability of those safety systems that provide a defense-in-depth, such as Containment Cooling System or Emergency Core Cooling Systems (ECCS). If testing or maintenance activities must be performed, a review of the potential risk impact will be performed. Ensure all required materials, tools, and personnel are available, prior to entering the RICT, and perform maintenance activities around the clock.

3. Brief the on-shift operations crew concerning the unit activities, including any compensatory measures established.

9. References

1. Nuclear Energy Institute (NEI) Topical Report (TR) NEI 06-09, "Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines,"

Revision 0-A, (ADAMS Accession No. ML12286A322). dated October 12, 2012.

2. NUREG-1855, "Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decision Making," U.S. Nuclear Regulatory Commission, dated March 2009.

3. EPRI TR-1026511, "Practical Guidance on the Use of Probabilistic Risk Assessment in Risk-Informed Applications with a Focus on the Treatment of Uncertainty," Technical Update, Electric Power Research Institute, dated December 2012.

Enclosure 13 W3F1-2021-0003 Waterford 3 to Standard Technical Specification Cross Reference

W3F1-2021-0003 3 Page 1 of 6 Waterford 3 to Standard Technical Specification Cross Reference Technical Specification TSTF-505 Waterford Apply Comments Description Rev 2 3 RICT?

TS TS Completion Times 1.3 NA Example 1.3-8 1.3-8 NA NA Waterford 3 TS do not contain a section with examples.

This example will not be added to the Waterford 3 TS.

Reactor Protective System 3.3.4 (RPS) Logic and Trip Initiation One Matrix Logic Channel 3.3.4 A1 3.3.1 Yes Waterford 3 including actuation and instrumentation inoperable actions per Table 3.3-1 ACTION 1 Engineered Safety Features 3.3.6 Actuation System (ESFAS)

Logic and Manual Trip One or more Functions with one 3.3.6 B1 3.3.2 b Yes Waterford 3 including actuation and instrumentation Manual Trip or Initiation Logic actions per Table 3.3-3 ACTIONS 12, 15, 16, 19b, and 20b channel inoperable One or more Functions with one 3.3.6 D1 3.3.2.b Yes Waterford 3 including actuation and instrumentation Actuation Logic channel inoperable actions per Table 3.3-3 ACTIONS 12, 15, 16, 19b, and 20b RCS Loops - Mode 3 System 3.4.5 3.4.1.2 One RCS loop inoperable 3.4.5 A.1 3.4.1.2 a No Waterford 3 is only including Modes 1 and 2 TS in the scope of the RICT program Pressurizer 3.4.9 3.4.3.1 One required group of pressurizer 3.4.9 B.1 3.4.3.1 a Yes TSTF-505 changes are incorporated heaters inoperable Note that the Waterford 3 TS wording is different, but functionally equivalent to the NUREG-1432 verbiage ECCS - Operating 3.5.2 3.5.2 One LPSI subsystem inoperable 3.5.2 A.1 3.5.2 a Yes TSTF-505 changes are incorporated One or more trains inoperable for 3.5.2 B.1 3.5.2 b Yes TSTF-505 changes are incorporated reasons other than Condition A

W3F1-2021-0003 3 Page 2 of 6 Technical Specification TSTF-505 Waterford Apply Comments Description Rev 2 3 RICT?

TS TS Containment Air Locks 3.6.2 3.6.1.3 One or more containment air locks 3.6.2 C.3 3.6.1.3 b Yes TSTF-505 changes are incorporated inoperable for reasons other than Note that the Waterford 3 TS wording is different, but Condition A or B functionally equivalent to the NUREG-1432 verbiage Containment Isolation Valves 3.6.3 3.6.3 One or more penetration flow 3.6.3 A.1 Yes Waterford 3 is proposing to include actions for closed paths with one containment systems and other penetrations, consistent with TSTF 505 isolation valve inoperable (Only action 3.6.3. Not all penetrations are modeled in the PRA.

applicable to containment sump supply valves to the ECCS and containment spray pumps)

One or more penetration flow 3.6.3 B.1 Yes Waterford 3 is proposing to include actions for closed paths with one containment systems and other penetrations, consistent with TSTF 505 isolation valve inoperable for action 3.6.3. Not all penetrations are modeled in the PRA.

reasons other than A, E and F.

(Only applicable to penetration flowpaths with two (or more) containment isolation valves)

One or more penetration flow 3.6.3 D.1 Yes Waterford 3 is proposing to include actions for closed paths with one containment systems and other penetrations, consistent with TSTF 505 isolation valve inoperable. (Only action 3.6.3. Not all penetrations are modeled in the PRA.

applicable to penetration flowpaths with only one containment isolation valve and a closed system)

Containment Ventilation System 3.6.1.7 With a containment purge supply 3.6.3 3.6.1.7 b Yes Containment purge valves are part of the containment and/or exhaust isolation valve(s) isolation valve technical specification in NUREG-1432, but having a measure leakage Waterford 3 has a separate LCO for containment purge.

exceeding the limits of Surveillance Requirement 4.6.1.7.2

W3F1-2021-0003 3 Page 3 of 6 Technical Specification TSTF-505 Waterford Apply Comments Description Rev 2 3 RICT?

TS TS Containment Spray and Cooling 3.6.6 3.6.2.1 (CS)

Systems 3.6.2.2 (CCS)

One containment spray train 3.6.6 A.1 3.6.2.1 a Yes TSTF-505 changes incorporated inoperable One containment cooling train 3.6.6 C.1 3.6.2.2 Yes TSTF-505 changes incorporated inoperable One containment spray and one 3.6.6 D.1 Yes The Waterford 3 TS do not explicitly contain this Condition.

containment cooling train 3.6.6 D.2 However, the RICT program will provide the RICT for inoperable situations where 3.6.2.1 a and 3.6.2.2 are simultaneously entered.

Main Steam Isolation Valves 3.7.2 3.7.1.5 (MSIVs)

One MSIV inoperable in Mode 1 3.7.2 A.1 3.7.1.5 Yes TSTF-505 changes incorporated Mode 1 Action Main Feedwater Isolation Valves 3.7.1.6 (MFIVs)

With one or more MFIV inoperable 3.7.1.6 Yes TS 3.7.3 (NUREG 1432) is not in TSTF-505 however, the Waterford 3 TS and actions for the MFIV are consistent with the MSIV TS (3.7.1.5), which is in scope of TSTF-505.

This is acceptable because the TSTF states that there may also be plant-specific TS to which changes of the type presented in the TSTF may be applied. This meets criteria 1 explained in Attachment 1, Section 2.3.

W3F1-2021-0003 3 Page 4 of 6 Technical Specification TSTF-505 Waterford Apply Comments Description Rev 2 3 RICT?

TS TS Atmospheric Dump Valves 3.7.4 3.7.1.7 (ADVs)

One required ADV line inoperable 3.7.4 A.1 3.7.1.7 a Yes Waterford 3 TS has two actions (a and c) to cover the (auto TSTF-505 action. The TS wording is different, but actuation functionally equivalent to the NUREG-1432 verbiage channel inoperable) 3.7.1.7 c (inoperable for other reasons)

Auxiliary Feedwater (AFW) 3.7.5 3.7.1.2 Waterford 3 uses Emergency Feedwater (EFW)

System terminology and has site specific limiting conditions Turbine driven AFW train 3.7.5 A.1 3.7.1.2 a Yes Waterford 3 TS 3.7.1.2 Action a matches NUREG-1432, inoperable due to one inoperable TSTF-505. changes incorporated steam supply One AFW train inoperable in Mode 3.7.5 B.1 3.7.1.2 b Yes Waterford 3 TS 3.7.1.2 Action b is site specific but aligns 1,2, or 3 for reasons other than with NUREG-1432 action 3.7.5 B.1 Condition A With the EFW system inoperable 3.7.1.2 d Yes Waterford 3 TS 3.7.1.2 d does not have a TSTF-505 for reasons other than (a), (b), or counterpart.

(c), and able to deliver at least 100% flow to either steam This is acceptable because the TSTF states that there may generator also be plant-specific TS to which changes of the type presented in the TSTF may be applied. This meets criteria 1 explained in Attachment 1 Section 2.3.

Component Cooling Water 3.7.7 3.7.3 (CCW) System One CCW train inoperable 3.7.7 A.1 3.7.3 Yes Waterford 3 TS combine CCW and ACC into a single TS (3.7.3). This is an acceptable plant specific application of the type presented in TSTF-505

W3F1-2021-0003 3 Page 5 of 6 Technical Specification TSTF-505 Waterford Apply Comments Description Rev 2 3 RICT?

TS TS Service Water System (SWS) 3.7.8 3.7.3 One SWS train inoperable 3.7.8 A.1 3.7.3 Yes Waterford 3 TS combine CCW and ACC into a single TS (3.7.3). This is an acceptable plant specific application of the type presented in TSTF-505 Ultimate Heat Sink (UHS) 3.7.9 3.7.4 One or more cooling towers with 3.7.9 A.1 3.7.4.a Yes Waterford 3 utilizes plant specific TS for the Ultimate Heat one cooling tower fan inoperable Sink. Train operability is dependent dry bulb temperature and a combination of OPERABLE wet cooling tower and dry cooling tower fans as provided in TS Table 3.7-3. This is an acceptable plant specific application of the type presented in TSTF-505.

Essential Chilled Water (ECW) 3.7.10 3.7.12 One ECW train inoperable 3.7.10 A.1 3.7.12 Yes TSTF-505 changes incorporated AC Sources - Operating 3.8.1 3.8.1.1 One required offsite circuit 3.8.1 A.3 3.8.1.1 a Yes TSTF-505 changes incorporated inoperable One required DG inoperable 3.8.1 B.4 3.8.1.1 b(2) Yes TSTF-505 changes incorporated Two required offsite circuit 3.8.1 C.2 3.8.1.1 e Yes TSTF-505 changes incorporated inoperable One required offsite circuit 3.8.1 D.1 3.8.1.1 c Yes TSTF-505 changes incorporated inoperable AND One required DG 3.8.1.D2 inoperable One required automatic load 3.8.1 F NA No Waterford 3 TS does not have specific condition for the sequencer inoperable load sequencer. Operability of the DG requires an OPERABLE load sequencer DC Sources - Operating 3.8.4 3.8.2.1 3.8.3.1 One or two battery chargers on 3.8.4 A.3 NA No Waterford 3 TS does not have specific condition for one train inoperable restoring an inoperable battery charger One battery on one train 3.8.4 B.1 3.8.2.1 a Yes TSTF-505 changes incorporated inoperable

W3F1-2021-0003 3 Page 6 of 6 Technical Specification TSTF-505 Waterford Apply Comments Description Rev 2 3 RICT?

TS TS One DC electrical power 3.8.4 C.1 3.8.3.1 c Yes Waterford 3 TS 3.8.3.1 specifies connection issue for subsystem inoperable for reasons cause to be rectified within the allowed outage time. This is other than A or B an acceptable plant specific application of the type presented in TSTF-505 Inverters - Operating 3.8.7 3.8.3.1 One required inverter inoperable 3.8.7 A.1 3.8.3.1 b Yes TSTF-505 changes incorporated Distribution Systems - 3.8.9 3.8.3.1 Operating One or more AC electrical power 3.8.9 A.1 3.8.3.1 a Yes TSTF-505 changes incorporated distribution subsystems inoperable One or more AC vital 3.8.9 B.1 3.8.3.1 b Yes TSTF-505 changes incorporated buses inoperable.

One or more DC electrical power 3.8.9 C.1 3.8.3.1 c Yes TSTF-505 changes incorporated distribution subsystems inoperable Programs and Manuals 5.5 6.5 RICT Program 5.5.18 6.5.19 Yes TSTF-505 changes incorporated

v t e Letter Sequence Request
EPID:L-2020-LLA-0279, TSTF-505, TSTF-425, TSTF-505 (Approved, Closed)
Initiation Request, Request, Request Acceptance Supplement, Supplement Administration Meeting Questions 26 July 2021 Answers 16 May 2022 19 August 2022 Results Approval, Approval, Approval, Approval Other: ML22209A106
MONTHYEAR2021-06-24 [Table View]

W3F1-2021-0003, Application for Technical Specification Change to Adopt Risk-Informed Extended Completion Times - RITSTF Initiative 4B

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W3F1-2021-0003, Application for Technical Specification Change to Adopt Risk-Informed Extended Completion Times - RITSTF Initiative 4B

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