Text

License Amendment Request to Revise Technical Specification Section 3.7.8, Nuclear Service Water System, to Add a New Condition to Allow Single Pond Return Header Operation of the Nsws with a 30-Day Completion Time
	ML17261B255
Person / Time
Site:	Catawba
Issue date:	09/14/2017
From:	Simril T; Duke Energy Carolinas
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	CNS-17-014
	Download: ML17261B255 (153)
	v • d • e

Tom Simril

(~ DUKE Vice President ENERGY~ Catawba Nuclear Station Duke Energy CN01 VP I 4800 Concord Road York , SC 29745 o: 803. 701.3340 f: 803.701 .3221 tom.simril@duke-energy .com CNS-17-014 10 CFR 50.90 September 14, 2017 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

Subject:

Duke Energy Carolinas, LLC (Duke Energy)

Catawba Nuclear Station (CNS), Units 1 and 2 Facility Operating License Numbers NPF-35 and NPF-52 Docket Numbers 50-413 and 50-414 License Amendment Request to Revise Technical Specification Section 3. 7.8, "Nuclear Service Water System" to add a new condition to allow Single Pond Return Header Operation of the NSWS with a 30-Day Completion Time Pursuant to 10 CFR 50.90, Duke Energy requests a license amendment to revise the CNS Unit 1 and Unit 2 Technical Specifications (TSs). The proposed change will revise TS Section 3.7.8, "Nuclear Service Water System," to add a new condition to allow Single Pond Return Header Operation of the NSWS with a 30-Day Completion Time.

The Enclosure provides the description and assessment of the proposed change. Attachment 1 contains the List of Regulatory Commitments associated with the requested change.

Attachment 2 provides the existing TS pages marked-up to show the proposed changes.

Attachment 3 provides the existing TS Bases pages marked-up to show the proposed changes, for information only. Changes to the existing TS Bases will be implemented under the Technical Specification Bases Control Program. Attachment 4 provides the Probablisitic Risk Assessment (PRA) Peer Review Findings and Resolutions.

A meeting was held with the NRC staff on January 25, 2017 to discuss the proposed change described above.

In accordance with Duke Energy administrative procedures and the Quality Assurance Program Topical Report, this amendment request has been previously reviewed and approved by the Catawba Plant Operations Review Committee.

The proposed change has been evaluated in accordance with 10 CFR 50.91 (a)(1) using criteria in 10 CFR 50.92(c), and it has been determined that the proposed change involves no significant hazards consideration. The bases for these determinations are included in the Enclosure and Attachments.

www.duke-energy.com

U.S. Nuclear Regulatory Commission CNS-17-014 September 14, 2017 Page 12 This submittal contains three regulatory commitments; see Attachment 1 to the Enclosure.

Duke Energy is requesting that the NRC review and approve this LAR within one year from the date of submittal. Duke Energy is also requesting the standard 120-day implementation period in conjunction with this LAR.

In accordance with 10 CFR 50.91, Duke Energy is notifying the State of South Carolina of this request by transmitting a copy of this letter and enclosure to the designated State Official.

Please direct any questions or concerns to Carrie Wilson, Senior Engineer, Catawba Regulatory Affairs, at (803) 701-3014.

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

Executed on September 14, 2017.

Sincerely,

-IOYY"\~

Tom Simril Vice President, Catawba Nuclear Station

Enclosure:

Description and Assessment of the Proposed Change Attachments:

1. List of Regulatory Commitments

2. Technical Specification Pages (Mark-up)

3. Technical Specification Bases Pages (Mark-up, For Information Only)

4. PRA Peer Review Findings and Resolutions

U.S. Nuclear Regulatory Commission CNS-17-014 September 14, 2017 Page 13 xc (with enclosure and attachments):

C. Haney, Region II Administrator U.S. Nuclear Regulatory Commission Marquis One Tower 245 Peachtree Center Avenue NE, Suite 1200 Atlanta, GA 30303-1257 J. D. Austin, Senior Resident Inspector U.S. Nuclear Regulatory Commission Catawba Nuclear Station M. Mahoney, Project Manager U.S. Nuclear Regulatory Commission 11555 Rockville Pike Mailstop 0-8H4A Rockville, MD 20852 Susan E. Jenkins, Manager S.C. DEHEC Radioactive & Infectious Waste Management jenkinse@dhec.sc.gov

U.S. Nuclear Regulatory Commission CNS-17-014 Enclosure, Page 1 ENCLOSURE Description and Assessment of the Proposed Change

Subject:

License Amendment Request to Revise Technical Specification Section 3.7.8, "Nuclear Service Water System" to add a condition to allow the Single Pond Return Header Operation of the Nuclear Service Water System (NSWS) with a 30-Day Completion Time

1.

SUMMARY

DESCRIPTION

2. DETAILED DESCRIPTION 2.1 System Design and Operation 2.2 Current Technical Specifications Requirements 2.3 Reason for the Proposed Change 2.4 Description of the Proposed Change

3. TECHNICAL EVALUATION 3.1 Traditional Engineering Considerations 3.2 Deterministic and Risk Assessment (with References)

4. REGULATORY EVALUATION 4.1 Applicable Regulatory Requirements / Criteria 4.2 Precedent 4.3 No Significant Hazards Consideration Analysis 4.4 Conclusions

5. ENVIRONMENTAL CONSIDERATION

6. ENCLOSURE ATTACHMENTS:

1. List of Regulatory Commitments

2. Technical Specification Pages (Mark-up)

3. Technical Specification Bases Pages (Mark-up, For Information Only)

4. PRA Peer Review Findings and Resolutions (with References)

U.S. Nuclear Regulatory Commission CNS-17-014 Enclosure, Page 2

1.

SUMMARY

DESCRIPTION In accordance with 10 CFR 50.90, Duke Energy Carolinas, LLC (Duke Energy) requests an amendment to the CNS Unit 1 Renewed Facility Operating License (NPF-35) and the CNS Unit 2 Renewed Facility Operating License (NPF-52), by incorporating the attached proposed change into the Unit 1 and Unit 2 Technical Specifications (TSs). Specifically, the proposed amendment is a request to revise TS Section 3.7.8, "Nuclear Service Water System" to add a new condition to allow Single Pond Return Header Operation of the NSWS with a 30-Day Completion Time (CT). The NSWS has Duke System Designation RN.

The proposed change would revise the TS to include a Single Pond Return Header Operation to the Nuclear Service Water System (NSWS) that involves isolating one train of the NSWS Pond Return piping at the Auxiliary Building wall and maintaining the discharge crossover lines open between trains in the Auxiliary Building and Emergency Diesel Generator (EDG) Buildings. This provides a common safety related discharge path through the single remaining in-service Pond Return line. This alignment, Single Pond Return Header Operation, allows a Pond Return Header to be removed from service while a flow path is maintained through both trains of NSWS supplied equipment to the Standby Nuclear Service Water Pond (SNSWP).

The NSWS Single Pond Return Header Operation is necessary to allow internal inspections and modifications of the NSWS Pond Return buried piping between the Auxiliary Building and the discharge to the SNSWP. The amount of time required to perform current internal inspections and planned modifications will take in excess of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , which is the current TS CT for the NSWS described in TS 3.7.8. In addition, this requested TS amendment will support future inspections in conformance with 10 CFR 50 Appendix A, General Design Criterion (GDC) 45, Inspection of Cooling Water Systems, and NRC Generic Letter 89-13, Service Water System Problems Affecting Safety-Related Equipment.

The timeframe for these inspections and modifications and subsequent periodic follow-up inspections of the NSWS Pond Return Headers will be bounded by the requested 30-day CT for NSWS Single Pond Return Header Operation. This timeframe is supported by the PRA analysis of Single Pond Return Header Operation.

2. DETAILED DESCRIPTION 2.1. System Design and Operation The NSWS, including Lake Wylie and the Standby Nuclear Service Water Pond (SNSWP), is the ultimate heat sink for various QA Condition 1 heat loads during normal operation, design basis events (Condition II, III, and IV Events per ANSI N18.2-1973), and other design events as dictated by Catawba licensing criteria.

During normal operation, the NSWS supplies cooling water to various safety related components. While in normal operation, the maximum heat load and flow requirements on the NSWS are encountered with the Unit in Mode 5 due to decay heat removal requirements.

During ANSI N18.2-1973 initiated events, the NSWS is required to support Emergency Core Cooling System (ECCS) operation by providing cooling water to various safety related components along with emergency makeup to selected QA Condition 1 Systems. The ANSI N18.2-1973 event that imposes the most stringent design basis requirement is the Condition IV Initiator of Loss of Coolant Accident (LOCA). In accordance with 10 CFR 50 Appendix A, GDC 2, Natural Phenomena, Catawba must withstand the effects of a Safe Shutdown Earthquake (SSE) without affecting the ability of the safety systems to shut down the plant. As such, the

U.S. Nuclear Regulatory Commission CNS-17-014 Enclosure, Page 3 ANSI N18.2-1973 design basis events are considered after the occurrence of an earthquake.

This means that a Loss of Lake Wylie and a dual unit Loss of Offsite Power (LOOP) are assumed.

Additional licensing criteria design events include Loss of Control Room, Fire Protection, and Security Events. Each of these events imposes specific requirements on the functions of the NSWS. A review of the UFSAR, Design Basis Specifications, and station procedures shows that the NSWS Single Pond Return Header Operation will not adversely affect the plant response to these events.

2.2. Current Technical Specifications Requirements The current TS CT applicable to the NSWS Single Pond Return Header Operation allows 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to restore an inoperable NSWS train, as specified in TS 3.7.8 Condition A.

2.3. Reason for the Proposed Change The NSWS Single Pond Return Header Operation is necessary to allow planned maintenance activities, such as internal inspections and modifications, of the NSWS Pond return buried piping between the Auxiliary Building and the discharge to the SNSWP. The amount of time required to perform current internal inspections and planned modifications will take in excess of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , which is the current TS CT for the NSWS described in TS 3.7.8. In addition, this requested TS amendment will support future inspections in conformance with 10 CFR 50 Appendix A, GDC 45, Inspection of Cooling Water Systems, and NRC Generic Letter 89-13, Service Water System Problems Affecting Safety-Related Equipment.

The current issues that require the additional completion time are the implementation of modifications addressing the abandonment of carbon steel piping from each units diesel generators and the resolution of an Operable But Degraded/Nonconforming (OBDN) condition related to the flow distribution to the long and short leg discharges to the SNSWP.

The original NSWS piping to and from the EDGs has been replaced with an alternate material utilizing a new route. During the implementation of these modifications, the original carbon steel piping was cut and capped inside the respective EDG buildings. The connection to the pond return headers was not terminated, resulting in four piping "dead legs." These dead legs result in vulnerability for future leaks that could divert return flow to the SNSWP. Additionally, a through wall leak is not in compliance with Selected Licensee Commitment (SLC) 16.5-5 (Structural Integrity). The identification, characterization, and remediation of a leak in this piping could not be achieved within the immediate completion time note in SLC 16.5-5. The other required action for a leak in an ASME Code Class 3 component is to immediately isolate the affected component. This action would result in the entry into Condition A of TS 3.7.8 which has a 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months CT. Since the structural integrity of the piping could not be verified or restored within the 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months CT, both units would be required to be in Mode 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months because the SNSWP return headers are shared between units.

Installation of ASME Section III piping isolation boundaries will remove the potential source for leaks, which challenge the reliability and availability of the SNSWP return headers.

U.S. Nuclear Regulatory Commission CNS-17-014 Enclosure, Page 4 The return headers to the SNSWP split and discharge flow to separate discharge structures on opposite ends of the SNSWP to assure that surface cooling will occur in all areas of the pond.

An orifice was installed to create a pressure drop in the shorter of the two discharge lines to divert flow to the longer of the discharge lines and assure surface cooling over the entire SNSWP. During the original design of NSWS, this orifice was sized based on higher design flow rates for all of the components cooled by the system. The resulting orifice size does not provide sufficient backpressure at the flow rates that occur normally, or during the response to a design basis event. This results in all of the flow to the SNSWP returning via the shorter of the discharge lines. Per the Ultimate Heat Sink safety evaluation in UFSAR Section 9.2.5.4, the flow split analyzed is 75 percent to the shorter discharge and 25 percent to the longer discharge. After evaluation, the NSWS was determined to be OBDN due to the discrepancy with the UFSAR. The evaluation determined that the dissipation of heat with all return flow through the shorter discharge following a design basis event was not impacted, resulting in the SNSWP being operable.

Implementation of the modification to resize the orifice will return the flow distribution to agree with the UFSAR description. The improved flow distribution will also increase heat dissipation margin from the SNSWP following a design basis event.

During the execution of the orifice replacement and piping abandonment-modifications utilizing the requested 30-day completion time, multiple inspections will be performed at various locations in each return header. The inspections will include cleaning of sections of piping (approximately 20 foot sections) to allow base metal inspections and ultrasonic testing at header manway accesses to establish a firm baseline of piping condition to meet aging management requirements per Chapter 18 of the UFSAR, to credit GL 89-13 inspections and to aid in the determination of potential future refurbishment actions to maintain the reliability of the SNSWP return headers. Additional robotic inspections will be performed of piping that is not as easily accessible.

The addition of the requested condition will allow the implementation of the orifice replacement and piping abandonment modifications as well as allow future inspections and implementation of future modifications. The changes and inspections will ensure the long-term reliability of the NSWS and permit the recovery of operating margin associated with temperature dispersion in the SNSWP and piping integrity in later years of plant life.

2.4. Description of the Proposed Change TS 3.7.8 Nuclear Service Water System (NSWS)

Duke Energy proposes to modify the Catawba NSWS TS 3.7.8 to add a new Condition D and to rename the existing Condition D to E. Minor clarifications of the existing Conditions A, B, and C will also be included due to the new condition.

The proposed changes to TS Section 3.7.8, Nuclear Service Water System (NSWS) are contained in Attachments 2 and 3.

The new Condition D will state One NSWS Pond return header inoperable due to NSWS being aligned for single Pond return header operation.

The existing Condition D will be revised to be Condition E and will state Required Action and associated Completion Time of Condition A, B, C, or D not met.

U.S. Nuclear Regulatory Commission CNS-17-014 Enclosure, Page 5 The required action for the new Condition D will state: Restore NSWS Pond return header to OPERABLE status. The CT for this Condition is proposed to be 30 days. The required actions with the respective completion time for the new Condition E will remain as Be in MODE 3 in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months AND be in Mode 5 in 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

The new Condition D text below will include clarification noted which is very similar to the notes provided for the existing Condition B (single supply header operation) and Condition C (single NSWS Auxiliary Building discharge header).

1. "Entry into this Condition shall only be allowed for pre-planned activities as described in the Bases of this Specification."

2. "Immediately enter Condition A of this LCO if one or more NSWS components become inoperable while in this Condition and one NSWS train remains OPERABLE."

3. "Immediately enter LCO 3.0.3 if one or more NSWS components become inoperable while in this Condition and no NSWS train remains OPERABLE."

Limiting the new Condition D to one NSWS Pond return header being inoperable (vs. an entire NSWS train inoperable) is necessary to prevent cascading inoperability to NSWS cooled components that have a less than 30 day CT.

For example, an inoperable "A" train of the NSWS would cascade, resulting in the inoperability of the "A" train EDG and Component Cooling Water (CCW) systems, which have 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months CTs.

Therefore, the new Condition D specifies one NSWS Pond return header as being inoperable as opposed to the entire affected train. Since both trains of the NSWS are fully capable of meeting the required flow requirements in the proposed alignment, the operable return header effectively meets the functional requirements of the isolated header; thereby, avoiding the impact of cascading to other systems, structures and components.

The proposed NSWS Single Pond Return Header Operation can be described as follows:

The proposed NSWS Single Pond Return Header Operation will involve isolating one train of the NSWS Pond Return piping at the Auxiliary Building wall and maintaining the discharge crossover lines open between trains in the Auxiliary Building and EDG Buildings. This will provide a common safety related discharge path through the single shared in-service Pond Return line. This proposed alignment will allow a Pond Return Header to be removed from service while a shared discharge flow path is maintained for all essential NSWS Supplied equipment to the SNSWP. While in this alignment, the NSWS will be capable of supplying required essential equipment (both trains of both units) with the design basis cooling water flows to support accident mitigation on one unit and the cool down loads of the other unit.

U.S. Nuclear Regulatory Commission CNS-17-014 Enclosure, Page 6 Additionally, in this proposed alignment, the NSWS will be pre-aligned to the SNSWP, which is the assured source of cooling water for the NSWS. This removes the risk of an active failure where an automatic valve fails to reposition on a loss of Lake Wylie. Specifically:

1. One of the two "in series" Auxiliary Building Lake Return Isolation valves (1RN57A or 1RN843B) will be open to increase the reliability of swapping discharge to Lake Wylie, if the alternate discharge path is needed.

2. The NSWS Return Header Crossover Valves 1RN53B and 1RN54A are open with power removed, and therefore will not auto-close on low-low NSWS suction pit level or Transfer to the Auxiliary Shutdown Panel.

3. The automatic valves in the safety flow path for alignment to the SNSWP are open with power removed. For work and inspections on the "A" Train Pond Return Header, valves 1RN58B, 2RN848B, and 1RN848B are opened with power removed. Similarly, for work and inspections on the "B" Train Pond Return Header, valves 1RN63A, 2RN846A, and 1RN846A are opened with power removed.

4. The NSWS suction supply is aligned to the SNSWP.

5. The Unit 1 and Unit 2 NSWS Non-Essential headers are isolated. Isolation valves are closed with power removed.

6. Power remains on the closed Motor Operated valves, which isolate the Lake Wylie Return flow paths. This allows a rapid re-establishment of discharge flow if the alternate path is needed. These are valves 1(2)RN849B, 1(2)RN847A, and either 1RN843B or 1RN57A.

7. The four manually operated NSWS Return Header Crossover Valves 1(2)RNP08 and 1(2)RNP09, in both units Diesel Generator Buildings, will be locked open.

General Comments:

1. The NSWS cannot be aligned in Single Pond Return Header Operation if the NSWS is already in the NSWS Single Auxiliary Building Discharge Header alignment or the NSWS Single Supply Header alignment. These configurations are described in CNS TSs 3.7.8 and the associated TS Bases. The combination of any two of these alignments has not been analyzed.

2. While the NSWS is aligned in Single Pond Return Header Operation, Unit 1 and Unit 2 are in a "TS Action Statement" for the affected NSWS Pond return header (the isolated Pond Return Header).

3. It is intended that NSWS Single Pond Return Header Operation be utilized with both Units in Mode 1. There is no specific mode requirement for use of this alignment.

4. This requested condition will be entered for preplanned maintenance and inspections only.

It is anticipated that entry into the condition should not be required more often than once per year, per train.

U.S. Nuclear Regulatory Commission CNS-17-014 Enclosure, Page 7 RN Sing le Pond Return Header Long Leg Discharge SNSWP A Train Isolated Isolated Piping ~

Single Pond Return Header Flowpath 2RN8498 2RN8488 :iRNS JA to Lake Wylie Note3 4 Note4

N~~~ --------------------------I 2RNP08 : 2RNP09 28 NS Hx Note 5 1 Note 5 1RN588' Sp-0 I riBOiGl:fiADiGl

~I~

r 2RN2298 Note3 ;

I 28 KD : 2AKD Hx , Hx '

1RN63A:

Notes: 28 I Train

1. 1 RN57A is open to increase the reliability of swapping to Lake Wylie if this action is needed.

2. RN AJB Header crossover valves are open with power I removed and therefore will not auto-close on RN Low-Low '

suction Pit level or transfer to the Aux Shutdown Panel. ~ :c::i

3. RN Return isolation valve to the SNSW P is open (power ~ *~ I I

removed) with RN suction aligned to the SNSW P. I A 8 Train Train

4. Power remains on these valves to allow repositioning from the control room if the alternate discharge path is needed.

5. RN D/G Header crossover valves will be locked open.

18KD I 1AKD 18 NS Hx Hx I Hx 1RN8438 1RN57A (i'Bi5iG"\1rAOJG) Note4 Note1

~ 1 Room I

1RNP08 : 1RNP09 Note 5 1 Note 5 SharedReturnf-~~~~~-+~~~~~~~~~~~--1"7--~~~~~~~~~~~~~~~~

to Lake Wylie Figure 2-1 Proposed Slngle Pond Return Header Operation for the NSWS

U.S. Nuclear Regulatory Commission CNS-17-014 Enclosure, Page 8 RN Single Pond Return Header Long Leg Discharge SNSWP B Train Isolated A Train 8 Train Short Leg Orifices Isolated Piping ~

Single Pond Return Header Flowpath Shared Return

~RN 2RN8498 Note4 :

847A _________________________

N <?_t~ ~

to Lake Wylie I

RNP08 : 2RNP09 28NSHx Note 5 , Note 5

'~I O.~ )

2iiiiXi):(2ARoom 2RN2298 1RN588:

28KD : 2AKD Hx , Hx 1RN63A:

28 Notes: Train Sp.() I

1. 1 RN8438 is open to increase the reliability of swapping to Note3:

Lake Wylie if this action is needed.

2. RN Header crossover valves are open with power I removed and therefore will not auto-close on RN Low-Low I suction Pit level or transfer to the Aux Shutdown Panel. (AiiXl:rv:;;;"'I

~1 c:J

3. RN Return isolation valve to the SNSWP is open (power removed) with RN suction aligned to the SNSWP. A 8 Train Train

4. Power remains on these valves to allow repositioning from the control room if the alternate d ischarge path is needed.

5. RN D/G Header crossover valves will be locked open.

18KD : 1AKD 18 NS Hx Hx (1 00.~1 *~ Hx 1RN8438 1RN57A Room 1lB22!!l.) Note 1 NoJe 4 I

1RNP08 : 1RNP09 ____________________________ !

Note 5 1 Note 5 1RN8498 Note 4 1 -

Shared Return ~~~~-'-~~~~~~~~~~~~-'-~+-~~~~~~~~~~~~~~~..../

to Lake Wylie Figure 2-2 Proposed Single Pond Return Header Operation for the NSWS

U.S. Nuclear Regulatory Commission CNS-17-014 Enclosure, Page 9

3. TECHNICAL EVALUATION The justification for the NSWS Single Pond Return Header Operation supports the position that while each Unit will be in an action statement for one Pond return header of NSWS, both trains of NSWS are capable of providing adequate flow to essential loads on both trains of both units to support design basis requirements. Since there is a reduction in redundancy by sharing of a single return header, PRA input is used along with a conventional evaluation of Single Failure and Pipe Rupture Criteria.

Duke calculation for NSWS Single Pond Return Header Design Basis analyzes the reliability of the NSWS while in Single Pond Return Header Operation against the Single Failure Criteria of GDC 44 and Pipe Rupture criteria as specified in the Standard Review plan. While in Single Pond Return Header operation, the isolated redundant return header to the SNSWP is not available, so a pipe rupture or a flow blockage cannot be mitigated by simply isolating crossover valves and switching to the redundant header. This calculation lays out a detailed alignment for NSWS Single Pond Return Header Operation and identifies possible failure mechanisms and locations. The evaluation in this calculation considers industry guidance and identifies a number of measures that can be taken to improve NSWS reliability while in this alignment. Flow modeling is included in this calculation and directs NSWS flow balancing such that adequate flow will be obtained to support all design heat load requirements. The calculation concludes that NSWS Single Pond Return Header Operation presents a configuration that can readily supply adequate flow to all equipment, and a NSWS that meets Single Failure and Pipe Rupture Criteria, with the following exceptions:

The removal of one Pond Return Header removes redundancy for NSWS Return piping to the SNSWP.

A failure of a single valve in the in-service flow path of NSWS (1RN58B or 1RN63A) can result in the temporary loss of the NSWS essential headers on both units, until the NSWS is re-aligned to Lake Wylie. However, the likelihood, consequences, and Operations response to this failure is very similar to the current accident response for the NSWS. The NSWS is aligned to Lake Wylie for normal operation and all design basis events, unless there is a loss of Lake Wylie. Note that valves 1RN-58B and 1RN-63A open on Sp signal, so events involving Hi-Hi Containment pressure (i.e. Large Break LOCA, Feed line Break in Containment, Steam line Break in Containment) would automatically open a discharge flow path to the SNSWP. For normal operation and events that do not generate an Sp signal (which automatically opens a flow path to the SNSWP), the NSWS discharges through the in series Lake Isolation Valves 1RN-57A and 1RN-843B. Either of these valves could fail closed and result in a loss of the single in-service NSWS flow path, which would require a realignment to the SNSWP to reestablish a discharge flow path for the NSWS Similarly, the response to a spurious failure to the closed position of the in-service SNSWP isolation valve 1RN-58B or 1RN-63A is to realign to Lake Wylie to reestablish a discharge flow-path for the NSWS.

A failure of a single valve, 1(2)RN846A or 1(2)RN848B, in the in-service NSWS pond return flow path from the EDGs, can result in the temporary loss of NSWS cooling water to both trains of EDGs for the affected unit. EDG cooling water flow alarms in the Control Room and EDG Annunciator panel alarms will alert the Operators to the failure. Procedures will direct mitigating actions, including re-alignment to Lake Wylie, to reestablish a discharge flow path.

U.S. Nuclear Regulatory Commission CNS-17-014 Enclosure, Page 10 Reg Guide 1.174 (An Approach for Using Probabilistic Risk Assessments in Risk-Informed Decisions on Plant Specific Changes to the Licensing Basis) allows the use of PRA risk analysis methodology in considering the acceptability of specific LAR submittals. The PRA analysis is documented in Duke calculations concludes that for the planned condition that a CT of 30 days presents a minimal and acceptable risk, and the intent of the conventional regulatory standards such as single failure criteria are met.

3.1 Traditional Engineering Considerations 3.1.1 Catawba UFSAR - Licensing Bases 3.1.1.1 UFSAR Section 3.6 Protection Against Dynamic Effects Associated With the Postulated Rupture of Piping Catawba's position in response to 10 CFR 50 App A GDC 4 and the Standard Review Plan BTP ASB 3-1 is documented in section 3.6 of the Catawba UFSAR. The following are applicable statements:

"At the time of the postulated pipe break, the plant is assumed to be in normal mode of plant operation in which the piping under investigation experiences the maximum conditions of pressure and temperature.

Consideration is given to the potential for a random single failure of an active component subsequent to the postulated pipe rupture.

3.6.2.1.2. General Design Criteria for Postulated Piping Breaks Other Than Reactor Coolant System

5. Consideration is given to the potential for a random single failure of an active component subsequent to the postulated pipe rupture. Where the postulated piping break is assumed to occur in one of two or more redundant trains of a dual-purpose moderate-energy essential system (i.e., one required to operate during normal plant conditions as well as to shut down the reactor and mitigate the consequences of the piping rupture), single failures of components in the other train or trains of that system only are not assumed, provided the system is designed to seismic Category I standards, is powered from both offsite and onsite sources, and is constructed, operated, and inspected to quality assurance, testing, and in-service inspection standards appropriate for nuclear safety systems.

3.6.2.1.2.2 Postulated Piping Break Locations For Moderate-Energy Piping Systems Systems identified as containing moderate-energy piping are examined by detailed drawing review for postulated through-wall cracks as defined below...

1. Cracks in Duke Class B, C and F piping are postulated at the following locations:

a. The terminal ends of the pressurized portions of the run.

b. At intermediate pipe-to-fitting weld locations of potential high stress or fatigue (e.g. pipe fittings, valves, flanges and welded attachments) that result in the maximum effects from fluid spraying, flooding or environmental conditions except in portions of piping where the maximum stress range is less than 0.4 (1.2 Sh + SA) as defined in items 2b2 and 3b2 of Section 3.6.2.1.2.1."

The above criterion, 3.6.2.1.2 Item 5, means that a single active failure need not be applied to equipment on the opposite train of the postulated break.

U.S. Nuclear Regulatory Commission CNS-17-014 Enclosure, Page 11 The above criteria, 3.6.2.1.2.2 Item 1.b, excludes the requirement to postulate moderate energy pipe breaks if the predicted stress level of the piping in that area is below a certain point. This precludes considering a pipe rupture in a well-supported system where maximum predicted material stress levels are low. This consideration will be applied in the Single Pond Return Header Operation to portions of the NSWS where pipe ruptures cannot be mitigated by isolation.

3.1.1.2 UFSAR Section 6.3.2.5 ECCS System Reliability Section 6.3.2.5 of the Catawba UFSAR addresses single failure considerations of ECCS Systems. Much of this is also applicable to NSWS as passive failures are addressed and their impact flooding in the Auxiliary Building. The UFSAR states under 6.3.2.5:

"...Subsequent Leakage From Components in Safeguards Systems With respect to piping and mechanical equipment outside the containment, considering the provisions for visual inspection and leak detection, leaks will be detected before they propagate to major proportions. A review of the equipment in the system indicates that the largest sudden leak potential would be the sudden failure of a pump shaft seal. Evaluation of leak rate assuming only the presence of a seal retention ring around the pump shaft showed flows less than 50 GPM would result. Piping leaks, valve packing leaks, or flange gasket leaks have been of a nature to build up slowly with time and are considered less severe than the pump seal failure.

... Assuming none of the RHRS and Containment Spray System room sump pumps are operating, the operator has at least 30 minutes from receipt of the high-level alarm to isolate the passive failure and prevent the sump from overflowing. However, with only one of the four Nuclear Safety Related sump pumps operating, the pump down rate exceeds the leakage rate.

Larger leaks in the ECCS are prevented by the following:

a. The piping is classified in accordance with ANS Safety Class 2 and receives the ASME Class 2 quality assurance program associated with this safety class.

b. The piping, equipment, and supports are designed to ANS Safety Class 2 seismic classification permitting no loss of function for the design basis earthquake.

c. The system piping is located within a controlled area of the plant.

d. The piping system receives periodic pressure tests and is accessible for periodic visual inspection.

e. The piping is austenitic stainless steel that, due to its ductility, can withstand severe distortion without failure."

3.1.1.3 UFSAR Section 9.2.1 Nuclear Service Water Section 9.2.1 describes relevant NSWS information contained in the Catawba UFSAR and represents the current licensing bases.

"9.2.1.1 Design Bases

... Sufficient redundancy of piping and components is provided to ensure that cooling is maintained to essential loads at all times.

U.S. Nuclear Regulatory Commission CNS-17-014 Enclosure, Page 12 9.2.1.2.4 Main Discharge Section There are two main discharge headers, extending the width of the Auxiliary Building with channel 1A and 2A components returning flow to the A header, and channel 1B and 2B components returning flow to the B header. During normal station operation when RN pumps are taking suction from Lake Wylie, discharge crossover valves are open, and all heat exchangers in operation discharge through the channel A return to Lake Wylie via the Low Pressure Service Water discharge. Automatically upon emergency low pump house pit level (as in loss of Lake Wylie), double isolation valves close on the return line to Lake Wylie, double isolation valves close on the discharge header crossover, and single isolation valves open on each channel return to the SNSWP. This sequence, along with isolation of the non-essential header and supply header crossover valves ensures two independent, redundant supplies and returns, satisfying the single failure criteria.

9.2.1.3 Safety Evaluation The NSWS is designed to withstand a safe shutdown earthquake and to prevent any single failure from limiting the ability for the engineered safety features to perform their safety functions.

...The RN System is designed to supply the cooling water requirements of a simultaneous LOCA on one unit and cooldown on the other unit assuming a single failure anywhere on the system, loss of offsite power and loss of Lake Wylie. Upon complete channel separation, both units are assured of having a source of water, at least one pump capable of supplying required flow on its associated channel, and at least one essential header to provide cooling water to components served by RN. Channels A and B are connected together only at six places: five between the RN supply headers and one between the RN discharge headers.

Redundant motor operated isolation valves are provided on the normally open crossover lines, and manual isolation valves are used on normally closed, rarely used crossover lines.

Table 9-4. Nuclear Service Water System Failure Analysis Component: Main SNSWP return valves 1RN63A, 1RN58B Malfunction: Failure to open on Loss of Lake Comment & Consequence: Each valve serves one shared train of RN System return to SNSWP, so failure of one valve to open when Lake return valves close results in failure of only one channel in both units. The remaining channel in each unit is sufficient to shut down both units safely.

If a Unit 1 diesel is known to be out of service, these valves are aligned to the Unit 2 diesel of corresponding channel.

Component: Channel A shared return line to SNSWP Malfunction: Rupture or plug Comment & Consequence: Isolate affected return line A and utilize backup train return line B until train A is repaired.

Component: Crossover valves 1RN53B or 1RN54A Malfunction: Failure to close on Loss of Lake.

Comment & Consequence: A and B valves are in series, so failure of either valve will not prevent channel separation when required."

U.S. Nuclear Regulatory Commission CNS-17-014 Enclosure, Page 13 In reference to the above failure analysis, the failures occurring during a swap to the SNSWP on a Loss of Lake Wylie will not be applicable in Single Pond Return Header Operation, as NSWS will be pre-aligned to the SNSWP. For the passive failures on the return line of "rupture or plug," mitigation cannot be a matter of swapping to the alternate train. A discussion of the credibility and response to these failures follows.

3.1.2 Catawba Documentation This section summarizes applicable Duke Calculations, Design Basis Documents, and Specifications.

3.1.2.1 Catawba Design Basis Document (DBD) on Plant Systems Single Failure Catawba Nuclear Station has documented its position on Single Failure Criteria in a Design Basis Document. Catawba has primarily adopted the guidance under the ANSI/ANS documents on Single Failure Criteria. The term "Single Failure Criterion" is applied to single failure rules based on the ANSI /ANS documents in conjunction with an ANSI N18.2-1973...Condition II, III, or IV event. The following is stated in the Single Failure DBD:

"3.2 Single Failure Criterion 3.2.1 Mechanical and Electrical systems shall be designed such that the unit can mitigate the consequences of a Design Basis Event (Condition II, III and IV) in proceeding to a safe shutdown assuming a single failure.

For Mechanical systems, the single failure shall be either:

1. An active failure during the short term

2. An active or passive failure in the long term, assuming no prior failure during the short term.

3.1 Definitions 3.1.2 Active Failure: The failure of a powered component such as a piece of mechanical equipment, component of the electrical supply system or instrumentation and control equipment to act on command to perform its design function. Examples include the failure of a valve to move to its correct position, the failure of an electrical breaker or relay to respond, the failure of a pump, fan or diesel generator to start, etc.

Consideration of equipment moving spuriously from the proper safeguards position, such as a motor operated valve inadvertently shutting, is specifically excluded.

3.1.3 Passive Failure: The structural failure of a static component that limits that components effectiveness in carrying out its design function. When applied to a Westinghouse designed fluid system, this means a break in the pressure boundary resulting in abnormal leakage not exceeding 50 GPM for 30 minutes. Such leak rates are consistent with limited cracks in pipes, sprung flanges, valve-packing leaks, pump seal failures, check valve failures, etc. In the absence of specific safety commitments, this value should be assumed when evaluating the effect of such passive failures on ECCS performance."

U.S. Nuclear Regulatory Commission CNS-17-014 Enclosure, Page 14 3.1.2.2 Catawba Calculation, Single Failure of the Nuclear Service Water System The most comprehensive evaluation of single failure in the NSWS is contained in calculation, A Flow Distribution Model of the RN System. This calculation elaborates on the Nuclear Service Water analysis in the Catawba UFSAR Table 9-4 and considers multiple equipment losses on failures of relays in addition to component failures. This calculation is predominantly a review of active failures and has been reviewed with respect to Single Pond Return Header Operation.

Since the NSWS will be pre-aligned to the SNSWP, an active failure involving a valve failing to operate during a swap to the SNSWP is not an issue. As a result, the most limiting single failure is the loss of one EDG and its associated NSWS Pump.

3.1.2.3 Catawba Specification, Catawba Pipe Rupture Analysis Criteria Specification The specification describes the methods and criteria used to perform the pipe rupture analysis for the design of Catawba Nuclear Station. The following statements are applicable to this consideration of Single Pond Return Header Operation:

"3.2 DEFINITIONS Active Component - A component that requires mechanical, electrical, pneumatic, or hydraulic actions to activate or deactivate in order to perform its safety function.

Main Run - Piping interconnecting terminal ends. All branch lines from the main run are considered branch runs, except that all branch lines that are included within the main run piping in the stress analysis computer mathematical model and are shown to have a significant effect on the main run behavior may be considered part of the main run.

Piping Run - A main or branch run that is bounded by terminal ends.

7.1 PROTECTION REQUIREMENTS 7.1.3 PROTECTION FROM RUPTURE OF MODERATE-ENERGY PIPING SYSTEMS

...through-wall cracks shall be postulated in moderate energy systems outside containment in accordance with Paragraph 7.2; and it shall be demonstrated by analysis that environmental conditions, flooding associated with the escape of contained fluids and water spray cannot cause unacceptable levels of damage to essential components and systems:

7.1.4 PROTECTION CRITERIA AND ASSUMPTIONS

b. Consideration is given to the potential for random single failure of an active component subsequent to the postulated pipe rupture...

p. When evaluating the consequences of pipe breaks for both high and moderate energy piping system, a minimum of thirty minutes shall be allowed for operator actions unless lesser times are fully justified considering manual action, alarms, system information available, and specific break location.

7.2.2 MODERATE ENERGY PIPING SYSTEMS 7.2.2.1 BREAK LOCATIONS Through-wall leakage cracks shall be postulated to occur at locations determined from the following criteria:

a. Cracks in Duke Class B, C and F piping are postulated at the following locations:

U.S. Nuclear Regulatory Commission CNS-17-014 Enclosure, Page 15

1) The terminal ends of the pressurized portions of the run.

2) At intermediate pipe to fitting weld locations of potential high stress or fatigue (e.g. pipe fittings, valves, flanges and welded attachments) that result in the maximum effects from fluid spraying, flooding or environmental conditions, or as defined in Paragraph 7.2.1.2 or 7.2.1.3. Cracks are not postulated at any intermediate location where the maximum stress, S, is less that 0.4 (1.2Sh+Sa)."

The above criteria presented in the Catawba Pipe Rupture Analysis Criteria Specification duplicates the criteria listed previously from Section 3.6 of the Catawba UFSAR and is consistent with that in the Standard Review Plan (NUREG 0800) section 3.6.2 Branch Technical Position 3-1 (MEB) paragraph B.2.c.

3.1.2.4 Discussion of the NSWS Single Pond Return Header Operation Licensing Bases 3.1.2.4.1 Nuclear Service Water System Design Basis The NSWS, including Lake Wylie and the Standby Nuclear Service Water Pond, is the ultimate heat sink for various QA Condition 1 heat loads during normal operation, design basis events (Condition II, III, and IV Events per ANSI N18.2-1973), and other design events (as dictated by Catawba licensing criteria).

"During normal operation, the NSWS supplies cooling water to various safety related components. While in normal operation, the maximum heat load and flow requirements on the NSWS are encountered with the Unit in Mode 5 due to decay heat removal requirements.

During ANSI N18.2-1973 initiated events, the NSWS is required to support ECCS operation by providing cooling water to various safety related components along with emergency makeup to selected QA Condition 1 Systems. The ANSI N18.2-1973 event that imposes the most stringent design basis requirement is the Condition IV Initiator of Loss of Coolant Accident (LOCA). In accordance with 10 CFR 50 Appendix A, GDC 2, Natural Phenomena, Catawba must withstand the effects of an SSE earthquake without affecting the ability of the safety systems to shut down the plant. As such, the ANSI N18.2-1973 design basis events are considered after the occurrence of an earthquake. This means that a Loss of Lake Wylie and a dual unit Loss of Offsite Power (LOOP) are assumed.

Additional licensing criteria design events include Loss of Control Room, Fire Protection, and Security Events. Each of these events imposes specific requirements on the functions of the NSWS. A review of the UFSAR, Design Basis Specifications, and station procedures shows that NSWS Single Pond Return Header Operation will not adversely affect the plant response to these events.

Note that in a Loss of Control Room event, there is a transfer of control to the Aux Shutdown Panel (ASP) Complex. As part of the NSWS response to the transfer (in dual header return operation), the NSWS discharge header crossover valves 1RN53B and 1RN54A would get a signal to close. This action ensures that, with the failures that could be encountered during a swap to the SNWSP that NSWS Pump runout does not occur if insufficient NSWS Pumps were available. In NSWS Single Pond Return Header Operation, the crossover valves are blocked from closing to ensure both trains of NSWS have a discharge flow path. However, in Single Pond Return Header Operation, the NSWS is pre-aligned to the SNSWP, and loss of an entire NSWS Pit (during a swap to the SNSWP) is not assumed. This means sufficient NSWS Pumps

U.S. Nuclear Regulatory Commission CNS-17-014 Enclosure, Page 16 are available, including following a transfer to the ASP, and the disabling of this function is acceptable."

3.1.2.5 Catawba Nuclear Station Design Basis Event Response during NSWS Single Pond Return Header Operation 3.1.2.5.1 Technical Specification Completion Time From a conventional regulatory perspective, with both units in a short term TS LCO it is not a requirement to postulate a failure. This is based on the limited amount of time that the plant operates in the action statement and the overall low risk. This perspective is described in ANSI 58.9-1981 and NRC GL 80-30.

ANSI 58.9-1981: "If one train of a redundant safety-related fluid system or its safety-related supporting systems is temporarily rendered inoperable due to short-term maintenance as allowed by the unit technical specifications, a single failure need not be assumed in the other train.

NRC GL 80-30: "The specified time to take action, usually called the equipment out-of-service time, is a temporary relaxation of the single failure criterion, which, consistent with overall system reliability considerations, provides a limited time to fix equipment or otherwise make it OPERABLE."

NRC Reg Guide 1.174 allows PRA considerations to be used in decision making for licensee submittals for TS Amendments. This is applied to the NSWS Single Pond Return Header proposal, which extends the existing CT from 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to 30 days, and provides a NSWS alignment that meets Single Failure and Pipe Rupture Criteria, with the following exceptions:

The removal of one Pond Return Header removes redundancy for NSWS Return piping to the SNSWP.

A failure of a single valve in the in-service flow path of NSWS (1RN58B or 1RN63A) can result in the temporary loss of both NSWS essential headers of NSWS on both units, until the NSWS is realigned to Lake Wylie. However, NSWS Single Pond Discharge Header Operation and the current normal alignment to Lake Wylie can also result in a temporary loss of both trains of NSWS on both units due to one valve (1RN-57A or 1RN-843B) in the flow path failing closed (due to disc/stem failure). Mitigation of this failure involves re-alignment of the NSWS system to the alternate heat sink.

A failure of a single valve, 1(2)RN846A or 1(2)RN848B, in the in-service NSWS pond return flow path from the EDGs, can result in the temporary loss of NSWS cooling water to both trains of EDGs for the affected unit. EDG cooling water flow alarms in the Control Room and EDG Annunciator panel alarms will alert the Operators to the failure. Procedures will direct mitigating actions, including re-alignment to Lake Wylie, to reestablish a discharge flow path.

The characteristics of NSWS Single Pond Return Header Operation are discussed here as related to the Single Failure Criterion and Pipe Rupture. This was done to identify measures that can mitigate risks and provide input to the PRA analysis. The PRA analysis subsequently determined that the proposed CT of 30 days represents a minimal and acceptable risk.

U.S. Nuclear Regulatory Commission CNS-17-014 Enclosure, Page 17 3.1.2.6 Consideration of the Single Failure Criterion in NSWS Single Pond Return Header Operation The NSWS is designed to meet Single Failure Criterion of Appendix A GDC 44 that states, "Sufficient redundancy of piping and components is provided to ensure that cooling is maintained to essential loads at all times." Two trains of piping and components are provided such that if a failure were to occur on one train that the trains could be separated and the remaining train provide the design functions.

In Single Pond Return Header Operation, one of the return headers to the SNSWP is isolated at the Auxiliary Building wall, and the discharge crossover line in the Auxiliary Building is maintained open such that both trains can utilize the remaining single discharge path. This challenges compliance with the Single Failure Criterion in that a failure on one train cannot be mitigated by simply isolating crossover valves, and switching to the redundant header. For design basis events, the failure that must be considered is a single active failure or a single passive failure. The NSWS Single Pond Return Header Operation of allowing return headers to remain connected following a LOCA (Sp signal) is acceptable if design basis functions can still be met assuming a single failure in conjunction with a design basis event as described above.

The following points supports the NSWS Single Pond Return Header Operation as a low risk condition with respect to single failures:

3.2.1.7 Active Failures Active failures are evaluated in a Catawba Calculation. The most limiting active failure is the loss of a NSWS Pit due to the failure of 1RN3A or 1RN4B to open following the loss of Lake Wylie. This failure takes out the pit, and thus both NSWS pumps (of that train) are not available.

Other active failures, such as a failure of an EDG to start, result in only a loss of one NSWS Pump and are of lesser consequence. No single active failure can be postulated that would result in a leak or a significant diversion of flow that would affect the ability of the essential headers to provide required flow to essential components.

As part of the requirements for entry into NSWS Single Pond Return Header Operation, the NSWS will be pre-aligned to the SNSWP. This configuration removes the possibility of an active failure that could prevent one of the NSWS pump pits from being aligned to the SNSWP. This therefore prevents the complete loss of one pit on a single active failure. The next most limiting active failure would be failure of a EDG to start which would include loss of the associated NSWS Pump.

In Single Pond Return Header Operation, following a Loss of one NSWS Pump in a pit (due to failure of one EDG) during a design basis event (LOCA-Sp) on one unit, three remaining NSWS pumps will provide flow to all four essential headers and to the three remaining EDGs.

Adequate flow must be provided in this condition to support the LOCA loads of one unit and the shutdown loads of the other unit. Flow modeling, performed under the NSWS flow model calculation, along with cases modeled in Appendix A of calculation CNC-1223.24-00-0072 verify adequate flow and pressure to essential components while the system is aligned per Single Pond Return Header operation, assuming the worst-case single active failure of a loss of one EDG. This will be confirmed by NSWS flow balance testing prior to entry into Single Pond Return Header Operation. This supports the ability of the NSWS to meet its design basis requirements during Single Pond Return Header Operation coincident with an active failure.

U.S. Nuclear Regulatory Commission CNS-17-014 Enclosure, Page 18 3.112.8 Passive Failures Passive Failures are described in the ANS 58.9, Single Failure Criteria for Light Water Reactors, in the Catawba UFSAR, and in the Catawba Design Basis Document for Single Failure.

Passive failures take the form of external piping leakage or internal failures such as a valve stem/disc separation that may restrict flow.

3.1.2.9 External Leakage Passive Failures External Leakage passive failures are limited in flow rate by consideration of only credible failures such as flange or packing leaks. ANS 58.9 states, The design flow for a passive failure shall be defined by analysis of realistic passive failure mechanisms in the system, considering conditions of operation and possible failure or leakage modes, as appropriate. ... As an example...a review ... may result in the definition of a design leak rate for passive-failure evaluation based on maximum flow through a failed valve packing or pump mechanical seal.

The Catawba UFSAR in Section 6.3.2.5 (ECCS Reliability) uses this approach to limit the flow rate assumption of a passive failure leak of the ECCS piping outside of containment to 50 GPM.

This is based on (1) provisions for visual inspection and leak detection to detect leaks before they propagate to major proportions, (2) an evaluation indicating the largest potential sudden leak is a 50 GPM failure of a pump shaft seal, and (3) larger leaks in the ECCS are deemed non credible due to ECCS piping QA classification, seismic design, location, testing, inspection, and material.

Similar arguments can be made concerning the maximum passive failure leak rate from NSWS Piping. A review of components on the NSWS Discharge Header between the last isolation valve from essential components and the discharge point to the SNSWP was performed. This section of piping must remain in service and pressurized to provide a discharge flow path from all four essential headers (1A, 1B, 2A, and 2B). The assumed value for maximum credible external passive failure leakage is 50 GPM. This is conservatively accepted based on the calculated full packing loss leak of 18 GPM through the 30 Fisher butterfly valves just inside the Auxiliary Building wall.

With this amount of continuous leakage, the NSWS can still provide adequate flow to all essential components. In addition, the long-term diversion of flow from returning to the SNSWP can be tolerated without a significant loss of level of the SNSWP. The SNSWP Thermal Analysis calculation allows a 50 GPM leak for the 30-day mission of the SNSWP. This consideration is required, since following a loss of Lake Wylie, the NSWS will be aligned to the SNSWP on a long-term basis. Note that the SNSWP Thermal Analysis calculation assumes that the SNSWP level is at the minimum allowed elevation of 571 ft. per CNS TSs. Prior to aligning the NSWS for Single Pond Return Header Operation, the SNSWP makeup will occur as required to ensure that the SNSWP is at its overflow elevation of 574 ft., so additional volume will be available if any passive failure leaks were to occur. The additional volume between the minimum TS elevation of 571 ft. and the overflow elevation of 574 ft. is approximately 29 million gallons.

From a flooding standpoint, the Auxiliary Building is provided with four Nuclear Safety Related sump pumps that can each handle a continuous leak of greater than 50 GPM. The UFSAR in Section 6.3.2.5 (ECCS Reliability) in response to a 50 GPM leak states that ... Assuming none of the RHRS and Containment Spray System room sump pumps are operating, the operator has at least 30 minutes from receipt of the high-level alarm to isolate the passive failure and prevent the sump from overflowing. However, with only one of the four Nuclear Safety Related sump

U.S. Nuclear Regulatory Commission CNS-17-014 Enclosure, Page 19 pumps operating, the pump down rate exceeds the leakage rate. These pumps can be credited for handling the 50 GPM maximum credible passive failure NSWS external leakage in the Auxiliary Building.

For the external leak passive failure flooding consideration of the EDG Rooms, each EDG is provided with a sump and two Nuclear Safety Related Sump Pumps (WN System) each with a 50 GPM capacity. These have adequate capacity to provide long-term handling of an external leak passive failure in a EDG room. Similar to the Auxiliary Building piping the maximum assumed value for credible external passive failure leakage in the EDG Rooms is 50 GPM. This is conservative compared to the Auxiliary Building leakage as the EDG Room credible source is a packing leak through a smaller 10 Fisher butterfly valve.

In summary, for a passive failure external leak in the Auxiliary Building while in Single Pond Return Header Operation, the NSWS is considered to still be capable of meeting its design requirements even though train separation does not occur. In the existing dual discharge header design, a leak on one train can be isolated such that flow to essential components on the other train is not affected. It also allows the leak to be subsequently stopped by shutdown of pumps in the faulted train. In Single Pond Return Header Operation as discussed above, passive failure leaks will not divert enough flow to starve essential equipment of their needed flow. In addition, the amount of leakage postulated can be tolerated on a long-term basis without affecting the function of the NSWS or the ultimate heat sink.

3.1.2.10 Internal Blockage Passive Failures Internal blockage passive failures involve the structural failure of a component. ANS 58.9 states that, ... A passive failure is a failure of a component to maintain its structural integrity or the blockage of a process flow path. Blockage of a process flow path could occur, for example, due to the separation of a valve disc from its stem.

The instances of valves being in the shared discharge flow path is considered for the failure of inadvertent repositioning or blockage from a stem to disc failure. There is one such valve on each unit's shared EDG Discharge and two such valves on the shared Auxiliary Discharge.

Whereas these failures cannot be ruled out, they may be considered an acceptably low risk by the PRA based on consideration of valve reliability, and the availability of re-alignment to Lake Wylie. This is discussed as follows:

1. To prevent any potential obstruction to flow, the manual valves are physically locked in their safety position (open). Those with powered actuators (motor operated valves) will have power removed.

2. From an internal integrity standpoint these valves are considered highly reliable, with a low risk of failure. These are QA-1, Fisher Stainless Steel Posi-seal valves, maintained under a preventive Maintenance program, and have been evaluated for reliability by the station's Engineering Department.

3. In the unlikely case that a blockage does occur, the NSWS can be realigned to temporarily discharge to Lake Wylie until one of the essential discharge headers is restored. Abnormal Procedures will direct Operators to diagnose issues with the NSWS during Single Pond Return Header operation such as leaks or blockage and re-align the NSWS to Lake Wylie to restore flow. Note that Lake Wylie is assumed to be lost due to a seismic event in a design basis accident. However, this is a low probability event and

U.S. Nuclear Regulatory Commission CNS-17-014 Enclosure, Page 20 will contribute to the overall reliability of the NSWS in this situation. If Lake Wylie is available as a heat sink, the NSWS could be realigned (suction and discharge) to the Lake to restore a flow path for the NSWS. If Lake Wylie is not intact, only the NSWS discharge could be aligned to the Lake and would be discharging to a "dry" Lake until the out-of-service NSWS header was restored. In the meantime, this would decrease SNSWP level and inventory. With approximately 29.3 million gallons of inventory in the SNSWP between the overflow (574.0 ft.) and the TS minimum elevation (571 ft.), there would be approximately 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to restore the out of service NSWS header (assuming a flow rate of 20,300 GPM) before the TS minimum elevation was reached (neglecting any SNSWP inventory losses due to seepage, evaporation or leakage, or any SNSWP inventory gains due to rain or incoming flow from springs or yard catch basins). At elevation 573.0 ft., approximately 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months would be allowed to restore the out of service NSWS header.

4. As a feature that supports the ability to rapidly establish discharge flow to Lake Wylie, the system alignment is such that only one discharge valve would need to be opened from the affected area (U1 EDG, U2 EDG, or the Auxiliary Building) to initiate flow. In addition, these valves are motor operated valves that will have power on them and can be opened from the Control Room.

Note that a passive failure of a single valve in the in-service flow path of NSWS (1RN58B or 1RN63A) can result in the temporary loss of both trains of NSWS on both units, until the NSWS is realigned to Lake Wylie. However, the likelihood, consequences, and Operations response to this failure is very similar to the current accident response for the NSWS. The NSWS is aligned to Lake Wylie for normal operation and all design basis events, unless there is a loss of Lake Wylie. Note that valves 1RN-58B and 1RN-63A open on Sp signal, so events involving Hi-Hi Containment pressure (i.e. Large Break LOCA, Feed line Break in Containment, Steam line Break in Containment) would automatically open a discharge flow path to the SNSWP. For normal operation and events that do not generate an Sp signal (which automatically opens a flow path to the SNSWP), the NSWS discharges through the in series Lake Isolation Valves 1RN-57A and 1RN-843B. Either of these valves could fail closed and result in a loss of the single in-service NSWS flow path, which would require a realignment to the SNSWP to reestablish a discharge flow path for the NSWS. Similarly, the response to a spurious failure to the closed position of the in-service SNSWP isolation valve 1RN-58B or 1RN-63A is to realign to Lake Wylie to reestablish a discharge flow-path for the NSWS.

In summary, NSWS Single Pond Discharge Header Operation and the current normal alignment to Lake Wylie can result in a temporary loss of both trains of NSWS on both units due to one valve in the flow path failing closed (due to disc/stem failure). Mitigation of this failure involves re-alignment of the NSWS system to the alternate heat sink.

U.S. Nuclear Regulatory Commission CNS-17-014 Enclosure, Page 21 RN Single Pond Return Header Long Leg Flow Blockage Response Discharge SNSWP A Train Isolated

'

RN Intake A Train B Train

~~

Isolated Piping - Short Leg ~ ~

RN Discharge Discharge V Flowpath Postulated Flow X Blockage 2RN849B 2RN847A 28 NS Hx '

~:(2AD/G

~ 1 Room

) I 2RN229B 1RN588 :

I 28 KO 2AKD I Hx Hx 1RN63A:

28 I Train Actions:

1. Immediately Open 2RN849B to re-establish discharge flowpath to Unit 2 D/Gs.

2. Swap RN Suction to Lake Wylie: I I

a. Open 1 RN1 A , 1 RN2B, 1 RN5A, 1 RN6B. (not shown) (AiiX):C"l

b. Close 1 RN3A, 1 RN4B. (not Shown) ~ II~

I

3. Comple te Swap of RN Discharge to Lake W ylie: I A B

a. Open 1 RN849B, 1(2)RN847A , and 1 RN843B. Train Train

b. Restore power and Close 1 RN58B and 1(2)RN848B.

1BKD 1AKD 18 NS Hx Hx , Hx l!!22!!!.J IIrA l'iiiiiiG\ 1 D/GJ Room I

I I

1RN849B Shared Return *k----~----}----------'=====ttJ----------------~

to Lake Wylie Figure 3-1 NSWS Diagrams of Re-alignment in Response to Internal Blockage Passive Failures

U.S. Nuclear Regulatory Commission CNS-17-014 Enclosure, Page 22 RN Single Pond Return Header Long Leg Flow Blockage Response Discharge A Train Isolated Isolated Piping """'""""""""

RN Discharge Flowpath Postulated Flow X Blockage I

2RN8498 2RN847A 28 NS Hx 'I

~:(2ADIG

~ 1 Room

) I 2RN2298 1RN588:

I 2BKD 2AKD I Hx Hx 1RN63A:

I Actions:

1. Immediately Open 1 RN849B to re- establish discharge flow pa th to Unit 1 D/Gs.

2A I 2.Swap RN Suction to Lake Wylie: Train I

a. Open 1 RN1A , 1 RN2B, 1 RN5A, 1 RN6B. (not shown) (AiiXl :c::"'l

~ II~

b. Close 1 RN3A , 1 RN4B. (not Shown)

I I

3.Complete Sw ap of RN Discharge to Lake Wylie: A B Train Train a.Open 2RN849B, 1(2)RN847A, and 1RN843B.

b. Restore power and Close 1 RN58B and 1 (2)RN8488.

1B KD 1AKD 18 NS Hx Hx 1 Hx 1RN8438 1RN57A l'iB5iGl , ADIG

~ 1 Room I

11 1 I

I 1RNP08 : 1RNP09 ____________________________ !

I 1RN8498 1RN8488

_____________ I L 1RN847A

___________ i 1RN846A Shared Return *'i-----~--+------------'===i-+l---------------~

to Lake Wylie Figure 3-2 NSWS Diagrams of Re-allgnment In Response to Internal Blockage Passive Fallures

U.S. Nuclear Regulatory Commission CNS-17-014 Enclosure, Page 23 RN Single Pond Return Header Long Leg Flow Blockage Response Discharge SNSWP A Train Isolated

'- RN Intake A Train B Train

~~

Isolated Piping ..........., Short Leg ~ "'"

RN Discharge Discharge V Flowpath Postulated Flow X Blockage 2RN849B I

28 NS Hx '

~:(2AD/G ) I

~ 1 Room 2RN229B 1RN58B:

I 2B KD 2AKD I Hx Hx 1 1RN63A

~:

1. Immediately Open 1 RN843B to re-establish d is charge flowpath from the Aux Bldg.

2.Swap RN Suction to Lake Wylie:

(AiiXI c:"l

a. Open 1 RN1A, 1 RN2B, 1 RNSA, 1 RN6B. (not shown) ~~

b . Close 1 RN3A, 1 RN4B. (not Shown)

A B

3. Complete Sw a p of RN Discharge to Lake Wylie : Train Train a.Open 1(2)RN849B and 1(2)RN847A.

b . Restore power and Close 1 RN58B and 1(2)RN848B.

1B KD 1AKD Hx 1 Hx l.!!22!!!.J II rRoom

!IBiim 1 ADIG) 1RN849B Shared Return f--- - - _ , _ _ _ - + - - - - - - - - - - - - ' --+-H---- - - - - - - - - - - - - -- '

to Lake Wylie Figure 3-3 NSWS Diagrams of Re-alignment in Response to Internal Blockage Passive Failures

U.S. Nuclear Regulatory Commission CNS-17-014 Enclosure, Page 24 RN Single Pond Return Header Long Leg Flow Blockage Response Discharge f SNSWP B Train Isolated

"- RN Intake Isolated Piping .............

RN Discharge Flowpath Postulated Flow X Blockage 2RN8498 2RN847A I

2RNP08 : 2RNP09 28 NS Hx 'I

~:(2ADIG

~1 Room

) 1RN588:

I 2RN2298 I

2BKD 2AKD I Hx Hx 1 1RN63A 28 Train

~:

1. Immediately Open 2RN84 7A to re-establish flow to Unit 2 D/Gs.

2. Swap RN Suction to Lake Wylie:

('AuXl 0

a. Open 1 RN1A, 1 RN2B, 1 RN5A, 1 RN6B. (not shown) ~c::J

b. Close 1 RN3A, 1 RN4B. (not Shown)

A B

3. Complete the swap of RN Discharge to Lake Wylie: Train Train a.Open 1(2)RN849B, 1RN847A, and 1RN57A.

b. Restore power a nd C lose 1 RN63A and 1(2)RN846A.

1BKD 1AKD 18 NS Hx Hx , Hx t1BiiiG\ I(1ADIG) l!!22!!!J I Room 1

I 1RN8498 Shared Return f-----~--+rr------------~--+-----------------~

to Lake Wylie Figure 3-4 NSWS Diagrams of Re-alignment in Response to Internal Blockage Passive Failures

U.S. Nuclear Regulatory Commission CNS-17-014 Enclosure, Page 25 RN Single Pond Return Header Long Leg Flow Bockage Response Discharge (

SNSWP B Train Isolated

'- RN Intake A Train B Train Short Leg Orifices Isolated Piping -

RN Discharge Flowpath Postulated Flow X Blockage 2RN849B I

28 NS Hx '

~:[2A

~ 1 Room D~ 1 1RN588:

I 2RN229B I

2BKD 2AKD I

Hx Hx 1 1RN63A

~:

1. Immediately Open 1 RN847A to re-establish flow to Unit 1 D/Gs.

2 . Swap RN Suction to Lake Wylie:

(AiiXl c::1

a. Open 1 RN1 A, 1 RN2B, 1 RN5A, 1 RN6B. (not shown) L!!!!!J ~

b. Close 1 RN3A, 1 RN4B. (not Shown)

A B 3 . Complete the swap of RN Discharge to Lake Wylie: Train Train

a. Open 1(2)RN849B , 2RN847A, and 1 RN57A.

b . Restore power and Close 1RN63A and 1(2)RN846A 1B KD 1AKD 18 NS Hx Hx 1 Hx 1RN8438 1RN57A (iii5iG'l '

~ 1 Room 11A0~1 I

1RN8498 ~N846A Shared Return ~----'=~;t------------'=-7----------------~

to Lake Wylie Figure 3-5 NSWS Diagrams of Re-alignment in Response to Internal Blockage Passive Failures

U.S. Nuclear Regulatory Commission CNS-17-014 Enclosure, Page 26 RN Single Pond Return Header Long Leg Flow Blockage Response Discharge f SNSWP B Train Isolated A Train B Train Short Leg Orifices Isolated Piping Discharge RN Discharge Flowpath Postulated Flow Blockage x

2RN8498 I

~RN847A 28 NS Hx '

~:(2AD/G

~1 Room

) I 2RN2298 1RN588:

I 28 KO 2AKD Hx Hx 28 Train Actio ns:

1. Immediate ly O pen 1 RN57A to re-establis h flow from the Aux Bldg.

2. Swap RN Suctio n to Lake Wylie:

(AiiX) :c:"'l

a. Open 1 RN1A , 1 RN28, 1 RN5A, 1 RN68. (not shown) ~le.. I b . C lose 1 RN3A, 1 RN48. (not Shown) I I

A B

3. Complete the swap of RN Discharge to Lake Wylie: Train Train

a. Open 1(2)RN8498 and 1(2)RN847A.

b. Restore power and Close 1 RN63A and 1(2)RN846A.

1BKD 1AKD 18 NS Hx Hx , Hx 1RN8438 1RN57A (iiiiiiG'\ I1 r AD IG)

~ Room I

1RN846A 1RN8498 I Shared Return f----~-hl-----------~-+-----------------'

to Lake Wylie Figure 3-6 NSWS Diagrams of Re-alignment in Response to Internal Blockage Passive Failures

U.S. Nuclear Regulatory Commission CNS-17-014 Enclosure, Page 27 3.1.2.11 Passive Failure of the Isolation Boundary It is unclear if industry standards suggest that a passive failure includes the internal failure of a butterfly valve that could result in the sudden release of flow. It would seem that the sudden failure of a stem from disc of a closed butterfly valve would be much less likely than this failure occurring with flow through the valve. These isolation valves are high quality Fisher PosiSeal butterfly valves that have been evaluated for reliability against repositioning from a closed position, so it is therefore conservative to consider a spurious opening of an isolation valve.

In the NSWS Single Pond Return Header Operation, the section of return header piping that is isolated for inspections, maintenance, or modifications consists almost completely of buried piping outside of the boundary of any buildings (the isolation valves are inside the Auxiliary Building and EDG Buildings, so small sections of isolated piping are between the isolation valves and walls, are inside). Personnel access points to this piping will only be through open manways located in the yard areas.

While the NSWS is aligned per Single Pond Return Header Operation, there are no openings or access points in the isolated NSWS piping inside the Auxiliary Building or EDG Buildings, so any external flooding due to isolation valve failure or any other source will not flood into the Auxiliary Building or EDG Buildings.

In the unlikely event of a failure of an upstream isolation valve where the valve suddenly goes to the open position, NSWS discharge flow would exit through open manways to the yard. The two possibilities to consider, (1) flooding in the yard and (2) diversion of return flow to the SNSWP, are determined to be acceptable as follows:

1. Ability of the yard drainage system to handle released flow:

NSWS flow modeling shows that all flow released through open manways would flow out of the manways closest to the powerblock (manways M-5-1, M-6-1, and M-7-1). The resulting maximum predicted outflow of 18,591 GPM is based on the maximum flow setting of the throttle valves and four-pump operation. This flow is well within the station's yard drainage system flood handling capacity as described in Catawba calculation CNC 1114.00-00-0040, so no flooding of the Auxiliary Building will occur, and as a result, it will not affect the ability to safely shutdown the plant by normal procedures.

2. Diversion of return flow to the SNSWP:

NSWS discharge flow from open manways would not return to the SNSWP and therefore would reduce pond inventory. However, adequate margin for emergent leakage from the SNSWP is available in the SNSWP Thermal Analysis calculation to cover the volume of water that would be diverted with a leak of this nature and a 116 minute response time, and the additional 29.3 million gallons of inventory between SNSWP elevation 571.0 ft. and 574.0 ft. adds approximately 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to this response time. Approximately 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months of response time is available at a pond elevation of 573.0 ft. Isolation valves are available in the Aux Bldg and EDG buildings that can isolate flow to the out of service header and still retain flow paths to at least one complete return header of essential equipment per unit. A response time of 30 minutes is reasonable and is assumed by the Catawba Pipe Rupture Specification during similar responses to postulated pipe ruptures.

U.S. Nuclear Regulatory Commission CNS-17-014 Enclosure, Page 28 3.1.2.12 Consideration of Postulated Pipe Rupture Response in NSWS Single Pond Return Header Operation The Catawba licensing basis complies with the methods described by the Standard Review Plan (NUREG 0800) with Branch Technical Positions (ASB 3-1 and MEC 3-1) for postulating leaks on moderate energy piping such as the NSWS (RN). As directed by these documents, the initial condition is assumed to be "normal plant conditions." The pipe leak is evaluated for the ability to cooldown the plant with all equipment available except that equipment lost as a result of the pipe rupture (flooding or spray), and one active failure during mitigation of the leak. In addition, it is assumed that one active failure occurs while mitigating the leak - but not to occur on the opposite train of the affected piping.

The Branch Technical Position (ASB 3-1) for pipe ruptures states that pipe leaks will be postulated for moderate energy Service Water piping such as the NSWS. Specific crack sizes, based on the diameter of the pipe, and the locations that they are required to be postulated, are specified. Catawba specification Catawba Pipe Rupture Analysis Criteria Specification, is used to calculate the leak flow of 1898 gpm for a postulated crack on the 42" NSWS discharge pipe in the Auxiliary Building in Catawba calculation Flood Levels for Structures Outside of the Reactor Building.

Since both Units will be in TS Action Statements for the affected NSWS Pond return header, a failure is not postulated for the limited period of time of the action statement allowed by the TS.

This is conveyed by ANSI 58.9-1981 and NRC GL 80-30. However, for the purposes of defense in depth, the methods and ability to mitigate a pipe rupture during Single Pond Return Header Operation are discussed here and considered in the PRA evaluation.

In Single Pond Return Header Operation, the redundant return header to the SNSWP is not available, and a pipe rupture cannot be mitigated by simply isolating crossover valves and switching to the redundant header. However, the NSWS piping in the Auxiliary Building is robustly designed and supported, and much of this is below the stress threshold requirement to postulate a leak. For all locations where leaks are required to be postulated, leaks can be mitigated by isolation of affected piping and re-alignment of suction and discharge sources to Lake Wylie (if needed) with continued operation of the NSWS.

For all NSWS leaks, Station Abnormal Procedures will direct Operators to troubleshoot, determine the leak location, and perform re-alignment to isolate the leak and ensure adequate NSWS flow. NSWS Single Pond Return Header Operation for each train is such that motor operated valves will be used to perform the majority of the isolations and re-alignment to the SNSWP. A few manual isolation valves are part of the isolation and are in accessible areas of the Auxiliary Building or EDG Buildings. In addition, all identified motor operated valves have local hand wheels on the actuators and can be locally (manually) repositioned in the event of a loss of actuator power.

While the response to pipe leaks while in Single Pond Return Header Operation is not as simple as isolation of crossovers and switching to the redundant train, it is reasonable to expect resolution in the assumed 30 minutes per the Catawba Pipe Rupture Specification. Postulated leaks while in NSWS Single Pond Return are bounded by the existing case of an 1898 GPM leak on the 42" NSWS Header.

U.S. Nuclear Regulatory Commission CNS-17-014 Enclosure, Page 29 RN Single Pond Return Header Long Leg Pipe Rupture Response Discharge SN SWP A Train Isolated

'- RN Intake A Train 8 Train Short Leg Orifices Isolated Piping ~

RN Discharge Flowpath Postulated Leak in this Section -

of Piping I

2RN8498 2RN847A I

28 NS Hx '

~:(2ADH>

~ 1 Room

) I 2RN2298 1RN588:

I 28 KD 2AKD I Hx Hx 1RN63A:

Actions:

1. Swap RN Discharge to Lake Wylie :

a. Open 1(2)RN849B, 1(2)RN847A, and 1 RN843B.

I

b. Restore power and Close 1 RN58B and 1(2)RN848B. (AiiX) :c::"l

~ le..

2. Swap RN Suction to Lake Wylie:

a. Open 1 RN1 A, 1 RN2B, 1 RN5A, 1 RN6B. (not shown) A 8 b . Close 1 RN3A , 1 RN4B. (not Shown) Train Train 18 KD 1AKD 18 NS Hx Hx 1 Hx ffii'DiGl

~ : Room

' 11ADH>J I

I I I 1RN8498 1RN846A I

Shared Return ~---~-----11r-----------'=====t:cr----------------j to Lake Wylie Figure 3-7 NSWS Diagrams of Re-alignment in Response to Postulated Pipe Ruptures

U.S. Nuclear Regulatory Commission CNS-17-014 Enclosure, Page 30 RN Single Pond Return Header Pipe Rupture Response Long Leg Discharge SNSWP A Train Isolated Isolated Piping RN Discharge --- I Flowpath Post ulated Leak in this Section -

of Piping 2RN8498 I

28 NS Hx 'I f;):[2A0~

~1 Room 1 I 2RN2298 1RN588:

I 2BKD 2A KD Hx Hx 1 1RN63A Actions :

1. Resto re Power and Close 1 RN53B.

2A Train

('AuX) 0

~c::J A 8 Train Train 18KD 1AKD 18 NS Hx Hx , Hx l'iiiiiiGl

~ 1 Room IrA 0~1 I

1RN8498 IRN8488 : 1RN847A 1~N846A

'- -- ~ 1 ----------------------

Shared Return <----~----+-----------~__,.H-----------------~

to Lake Wylie Figure 3-8 NSWS Diagrams of Re-alignment in Response to Postulated Pipe Ruptures

U.S. Nuclear Regulatory Commission CNS-17-014 Enclosure, Page 31 RN Single Pond Return Header Long Leg Pipe Rupture Response Discharge SNSWP A Train Isolated

'- RN Intake A Train 8 Train Short Leg ~ v-Isolated Piping ............, Discharge V RN Discharge Flowpath Postulated Leak in this Section -

of Piping 2RN8498 I

~RN847A 28 NS Hx '

~:12AD~

~1 Room 1 2RN2298 1RN588:

I I

28KD 2AKD I

Hx Hx 1 1RN63A 28 Actions: Train

1. Swap RN Discharge to Lake Wylie:

a.Open 1(2)RN849B, 1(2)RN847A, and 1 RN843B.

b. Restore power and Close 1 RN58B and 1(2)RN848B.

(AiiX) ,C"'l

2. Restore power and Close 1 RN53B or 1 RN54A ~l~

3.Swap RN Suction to Lake Wylie.

A 8

a. Open 1 RN1A, 1 RN2B, 1 RNSA, 1 RNSB. (not shown) Train Train

b. Close 1 RN3A, 1 RN4B. (not Shown) 18KD 1AKD 18 NS Hx Hx 1 Hx (iiiDiG'llrA D~J l!!22!!!.J 'I Room 1~N846A 1RN8498 Shared Return f-----'---'7-----------'=~'!t-----------------_/

to Lake Wylie Figure 3-9 NSWS Diagrams of Re-alignment in Response to Postulated Pipe Ruptures

U.S. Nuclear Regulatory Commission CNS-17-014 Enclosure, Page 32 RN Single Pond Return Header Long Leg Pipe Rupture Response Discharge SNSWP A Train Isolated Isolated Piping -

RN Discharge Flowpath Postulated Leak in this Section -

of Piping 2RN8498 I

~RN847A 28 NS Hx '

~:12AD~

~1 Room 1 2RN2298 1RN588:

I I

28KD 2AKD I

Hx Hx 1 1RN63A 28 Train Actions:

1. Locally Close 1RNP19

('iiiiX) c:-a

~~

A 8 Train Train 18KD 1AKD 18 NS Hx Hx , Hx 1RN8438 1RN57A (iii'iiiGl

~ 1 Room IrAD~1 I

I 1RNP08 : 1RNP09 1RN846A 1RN8498 I Shared Re turn<'-----~--+-----------~-+\!-----------------~

to Lake Wylie Figura 3-10 NSWS Diagrams of Re-alignment in Response to Postulated Pipe Ruptures

U.S. Nuclear Regulatory Commission CNS-17-014 Enclosure, Page 33 RN Single Pond Return Header Long Leg Pipe Rupture Response Discharge SNSWP A Train Isolated Isolated Piping RN Discharge --- 1 Flowpath Postulated Leak in this Section -

of Piping 2RN8498 I

2RNP08 : 2RNP09 28 NS Hx '

~:~ I

~I~ 2RN2298 1RN588:

I 28KD 2AKD Hx Hx 1RN63A' 28 Train A ctions:

1. Swap RN Discharge to Lake Wylie (except 2B D/G)

a. Open 1 RN849B, 1(2 )RN847A, and 1 RN843B.

b. Restore power and C lose 1 RN58B. ~C"'\

2. Locally Close 2RNP09 and 1(2)RN848B.

~b 3 . Swap RN Suction to Lake Wylie. A 8 Train Train

a. Open 1 RN1A, 1 RN2B, 1 RN5A, 1 RNSB. (not shown)

b. Close 1 RN3A, 1 RN4B. (not Shown) 18KD 1AKD 18 NS Hx Hx , Hx 1RN8438 1RN57A (1'8"5Kil l1

~

rAD/G)

Room I

I 1RN8498 Shared Return '=~---'-----+----------~~rr---------------_,,

to Lake Wylie Figure 3-11 NSWS Diagrams of Re-alignment in Response to Postulated Pipe Ruptures

U.S. Nuclear Regulatory Commission CNS-17-014 Enclosure, Page 34 RN Single Pond Return Header Long Leg Pipe Rupture Response Discharge SNSWP A Train Isolated Isolated Piping ~

RN Discharge Flow path Postulated Leak in this Section -

of Piping 2RN8498 2RN847A I

28 NS Hx 'I

~:(2ADKi

~ 1 Room

) I 2RN2298 1RN588 :

I 28KD 2AKD I Hx Hx 1RN63A:

28 I Train Actions:

1. Locally Close 2RNP08.

(AuX) rv;;;-i

~c::J A 8 Train Train 18 KO 1AKD 18 NS Hx Hx , Hx (1ii'iiiGl IrADKiJ l!!.22!:!!) 1 Room I

1RN8498 IRN8488

_____________ I L 1RN847A

___________ i 1RN846A Shared Return E------~---t-----------~--+ft----------------~

to Lake Wylie Figura 3-12 NSWS Diagrams of Re-alignment in Response to Postulated Pipe Ruptures

U.S. Nuclear Regulatory Commission CNS-17-014 Enclosure, Page 35 RN Single Pond Return Header Long Leg Pipe Rupture Response Discharge SNSWP A train Isolated

'- RN Intake A Train 8 Train v

Isolated Piping -

RN Discharge Short Leg ~

Discharge v Flowpath Postulated Leak in this Section -

of Piping 2RN8498 2RN847A I

2RNP08 : 2RNP09 28 NS Hx '

~:[2Ao~

~1 Room

) I 2RN2298 1RN588 :

I 28KD 2AKD I Hx Hx 1RN63A:

28 I Train Actions:

1. Locally Close 1 RNP08 .

I I

(iiiiX) :C"'l

~le_,

I I

I A 8 Train Train 18KD I 1AKD 18 NS Hx Hx I Hx l'iiiiim' 11A l!!22!!!.J 1 0~1 Room I

1RN8498 Shared Return<------~--+----------~--.----------------~

to Lake Wylie Figure 3-13 NSWS Diagrams of Re-alignment in Response to Postulated Pipe Ruptures

U.S. Nuclear Regulatory Commission CNS-17-014 Enclosure, Page 36 RN Single Pond Return Header Long Leg Pipe Rupture Response Discharge SNSWP A Train Isolated A Train 8 Train Short Leg Orifices Isolated Piping -

RN Discharge Flowpath Postulated Leak in this Section -

of Piping 2RN8498 2RN847A I

28 NS Hx

~:(2A

~ 1 Room D/G ) I 2RN2298 1RN588 :

I 28 KD 2AKD I Hx Hx 1 1RN63A

~:

1. Swap RN Discharge t o Lake Wylie (except 1 B D/G) a.Open 2RN8498, 1(2)RN847A, and 1RN8438.

b . Restore power and Close 1RN588 and 1(2 )RN8488. (AiiX\ c::"'l

2. Locally Close 1RNP09. ~t::.J

3. Swap RN Suction to lake W ylie. A 8 Train Train

a. Open 1 RN1 A , 1 RN28, 1 RNSA, 1 RN68. (not shown) b . Close 1 RN3A , 1 RN48. (not Shown) 18KD 1AKD 18 NS Hx Hx , Hx 1RN8438 1RN57A IrRoom (iiiiiiGl 1

~

ADIG)

I I

1RN8498 1RN8488

_________ L I 1RN847A

___________ i 1RN846A l

Shared Return f-----~-+-----------~--+tt---------------~

to Lake Wylie Figure 3-14 NSWS Diagrams of Re-alignment in Response to Postulated Pipe Ruptures

U.S. Nuclear Regulatory Commission CNS-17-014 Enclosure, Page 37 RN Single Pond Return Header Long Leg Pipe Rupture Response Discharge SNSWP B Train Isolated

'- RN Intake A Train 8 Train Short Leg Orifices Isolated Piping -

RN Discharge Flowpath Postulated Leak in this Section -

of Piping 2RN8498 2RN846A 2RN847A I

2RNP08 : 2RNP09 28 NS Hx '

I r;):(2AD/G

~* Room

) 2RN2298 1RN588:

I I

28KD 2AKD I

Hx Hx 1 1RN63A 28 Train Actions:

1. Swap RN Discharge to Lake Wylie :

a. Open 1(2)RN849B, 1(2)RN847A, and 1RN57A.

b. Restore power and Close 1 RN63A and 1(2)RN846A. ('AuX) C"l

2. Swap RN Suction to Lake Wylie:

~~

a. Open 1 RN1A, 1 RN2B, 1 RN5A, 1 RN6B. (not shown) A 8 Train Train b . Close 1 RN3A, 1 RN4B. (not Shown) 18KD 1AKD 18 NS Hx Hx 1 Hx v 11BO/G)

Room 1

1 rAD/GJ Room I

I 1RNP08 : 1RNP09 1~N846A 1RN8498 Shared Return f------'----+~----------'----;i1t----------------'

to Lake Wylie Figura 3-15 NSWS Diagrams of Re-alignment in Response to Postulated Pipe Ruptures

U.S. Nuclear Regulatory Commission CNS-17-014 Enclosure, Page 38 RN Single Pond Return Header Long Leg Pipe Rupture Response Discharge SNSWP B Train Isolated

'- RN Intake A Train 8Train Short Leg Orifices Short Leg Isolated Piping - Discharge RN Discharge --- I Flowpath Postulated Leak in this Section -

of Piping 2RN8498 2RN847A I

28 NS Hx

~:(2ADIG

~ 1 Room

)

2RN2298 I

28 KD 2AKD I Hx Hx 1 1RN63A 28 Tr ain Actions:

1. None. Piping in this area is below the threshold for assuming a postulated leak.

(AiiXl C"'l

~e_,

A 8 Train iTrain 18KD 1AKD 18 NS Hx Hx 1 Hx 1RN8438 1RN57A l1ii"DiG\

~ Room IrADIG) 1 I

I 1RNP08 : 1RNP09 I

1RN8498 t RN8488 : 1RN847A l ~N846A

'- -- 11 ----------------------

Shared Return <i----~-ttt-----------~-+----------------j to Lake Wylie Figure 3-16 NSWS Diagrams of Re-alignment in Response to Postulated Pipe Ruptures

U.S. Nuclear Regulatory Commission CNS-17-014 Enclosure, Page 39 RN Single Pond Return Header Long Leg Pipe Rupture Response Discharge (

SNSWP B Train Isolated

'- RN Intake A Train 8 Train Short Leg Orifices

~

Short Leg Isolated Piping - Discharge RN Discharge --- I Flowpath Postulated Leak in this Section -

of Piping 2RN8498 2RN847A I

28 NS Hx '

~:(2AO~

~ 1 Room

] I 1RN588:

I 28 KD 2AKD Hx Hx 1 1RN63A

~:

1. Restore power and Close 1RN53B or 1 RN54A .

(AiiXl c.:"'\

~c:J A 8 Train rain 18 KD 1AKD 18 NS Hx Hx , Hx 1RN8438 1RN57A (1iiiiiG\

~ 1 Room IrA 0~1 I

I 1RNP08 : 1RNP09 1RN8488 : 1RN847A l ~N846A 1RN8498

'- 1 ----------------------

Shared Return <----~-.,.,,._----------~-+----------------~

to Lake Wylie Figura 3-17 NSWS Diagrams of Re-alignment in Response to Postulated Pipe Ruptures

U.S. Nuclear Regulatory Commission CNS-17-014 Enclosure, Page 40 RN Single Pond Return Header Long Leg Pipe Rupture Response Discharge SNSWP B Train Isolated

'- RN Intake A Train B Train Short Leg Orifices Isolated Piping """"""

RN Discharge Flowpath Postulated Leak in this Section -

of Piping 2RN849B 2RN847A I I 2RNP08 : 2RNP09 28 NS Hx '

~:[2AD/G

~ 1 Room

) I 2RN2298 1RN58B:

I 28 KD 2AKD I Hx Hx 1 1RN63A

1. Swap RN Discharge to Lake Wylie :

a. Open 1(2)RN849B, 1(2)RN847A , and 1 RN57A.

b. Restore powe r and Close 1 RN63A and 1(2)RN848B. f'TuXl C"l

~o

2. Swap RN Suction to Lake Wylie:

Tr~inl

a. Ope n 1 RN1A, 1 RN2B, 1 RN5A, 1 RN6B. (not shown) B Train

b. Close 1 RN3A, 1 RN4B. (not Shown)

3. Locally C lose 1RNP19.

18 KD 1AKD 18 NS Hx Hx , Hx 1RN8438 1RN57A ffiiiiiGl IrRoom l!!22!!!.J 1 ADIG)

I 1~N846A 1RN849B Shared Return f----~--+rr------------~--+----------------~

to Lake Wylie Figure 3-18 NSWS Diagrams of Re-alignment in Response to Postulated Pipe Ruptures

U.S. Nuclear Regulatory Commission CNS-17-014 Enclosure, Page 41 RN Single Pond Return Header Long Leg Pipe Rupture Response Discharge SNSWP Blain Isolated

'- RN Intake A Train 8 Train Short Leg Orifices Isolated Piping ~

RN Discharge Flowpath Postulated Leak in this Section -

of Piping 2RN8498 2RN847A I

28 NS Hx '

f;l:(2ADHi)

~1 Room 2RN2298 1RN588:

I I

28 KO 2AKD I Hx Hx 1 1RN63A 28 Train Actions:

1. Locally Close 2RNP09. (AiiX) r::::"'I

~~

A 8 Train rain 18KD 1AKD 18 NS Hx Hx 1 Hx 1RN8438 1RN57A (iB'6iGl 1r AD IG)

~ Room 1 I

I 1RNP08 : 1RNP09 I

1RN8498 ~1~~~~8~- ___ ~ __1~~~~7~_____ 1,RN846A Shared Return<-----~-=+-------------+----------------~

to Lake Wylie Figure 3-19 NSWS Diagrams of Re-alignment in Response to Postulated Pipe Ruptures

U.S. Nuclear Regulatory Commission CNS-17-014 Enclosure, Page 42 RN Single Pond Return Header Long Leg Pipe Rupture Response Discharge B Train Isolated Isolated Piping -

RN Discharge Flowpath Postulated Leak in this Section -

of Piping 2RN8498 I

2RN846A ~RN847A 2RNP08 : 2RNP09 28 NS Hx '

~:(2AD/G

~1 Room

) I 2RN2298 1RN588:

I 28KD 2AKD I Hx Hx 1RN63A:

28 I Train Actio ns :

1. Locally Close 2RNP08.

I (AiiX):C'l

~1e:J A 8 Train Train 18KD 1AKD 18 NS Hx Hx , Hx 1RN8438 1RN57A (iii'iiiGl I1 r ADIG)

~ Room I

I 1RN8498 1~N846A Shared Return <-----~--+-----------~-H-----------------J to Lake Wylie Figure 3-20 NSWS Diagrams of Re-alignment in Response to Postulated Pipe Ruptures

U.S. Nuclear Regulatory Commission CNS-17-014 Enclosure, Page 43 RN Single Pond Return Header Long Leg Pipe Rupture Response Discharge f SN SWP B Train Isolated

'- RN Intake A Train B Train

~~

Short Leg Orifices Short Leg*!,,.,.

Isolated Piping ............, Discharge RN Discharge Flowpath Postulated Leak in this Section -

of Piping 2RN849B ---, 2RNs4s8----: - 2RNs4sP.-- --- 2RN847A I ------------------------- - ---

2RNP08 : 2RNP09 28 NS Hx

~:(2ADIG

~ 1 Room I 1RN588 :

I 2RN2298 I

28KD 2AKD I Hx Hx 1 1RN63A Actions:

1 . Swap RN Discharge to Lake Wylie :

a.Open 1(2)RN849B, 2RN847A and 1 RN57A.

b. Restore power and Close 1 RN63A and 1(2)RN846A. (AiiX) C"'l

~~

2. Locally Close 1 RNP08.

3. Swap RN Suction to Lake Wylie: A 8 Train rain

a. Open 1 RN1A, 1 RN2B, 1 RNSA, 1 RN6B. (not show n)

b. Close 1 RN3A, 1 RN4B. (not Shown) 18KD 1AKD 18 NS Hx Hx , Hx 1RN8438 1RN57A (iii"i5iGl I l.!!22!!!) 1 rA DIG)

Room I

I I

I 1

___ 1 ~~~~8~- ___ ~ __1~~~~7~_____ 1 ~N846A 1RN8498 Shared Return +-----...l,....~~----~-------1---.::i----~==========="

to Lake Wylie Figure 3-21 NSWS Diagrams of Re-allgnment In Response to Postulated Pipe Ruptures

U.S. Nuclear Regulatory Commission CNS-17-014 Enclosure, Page 44 RN Single Pond Return Header Long Leg Pipe Rupture Response Discharge r SNSWP B Train Isolated

'- RN Intake ATrain 8 Train Short Leg Orifices Isolated Piping ~

RN Discharge Flowpath Postulated Leak in this Section -

of Piping 2RN8498 2RN847A I

RNPOS : 2RNP09 28 NS Hx '

f;):[2A

~1 Room DIG ) 1RN588:

I 2RN2298 I

~

28KD 2AKD I Hx Hx 1RN63A:

28 I Train Actions:

1. Swap RN Discharge to Lake Wylie (except 1 B D/G) a.Open 2RN849B, 1(2 )RN847A, and 1RN843B.

I b . Restore power and C lose 1 RN58B and 1(2)RN848B. f'AiiXI :c:;"'l

2. Locally Close 1RNP09. ~1eJ I I

I

3. Swap RN S uctio n to Lake Wylie. A 8 Train rain

a. O pen 1 RN1A, 1 RN2B, 1 RN5A , 1 RN6B. (not shown)

b. Close 1 RN3A, 1 RN4B. (not Shown) 18 KD 1AKD Hx , Hx IrA (iii'iiiGl 1 U!22!!!J I DIG)

Room 1RN8498 1RN847A Shared Return f--------b===~----=------L__:1-----------===========--/

to Lake Wylie Figure 3-22 NSWS Diagrams of Re-alignment in Response to Postulated Pipe Ruptures

U.S. Nuclear Regulatory Commission CNS-17-014 Enclosure, Page 45 3.1.2.13 Summary of Measures to Enhance Reliability Several measures have been described in this evaluation to enhance the reliability of the NSWS while aligned in Single Pond Return Header Operation. These are summarized here:

1. The NSWS is pre-aligned to the SNSWP prior to entering Single Pond Return Header Operation. The resulting benefit is not having to assume an active failure that would take out one complete NSWS pit (with two NSWS Pumps and two EDGs. The next most limiting active failure is loss of one EDG and its corresponding NSWS Pump.

2. Valves in the shared (single) flow path will be prevented from repositioning. Manual valves will be physically locked in the open position. Motor operated valves will be open with power removed and tagged to prevent local operation. Removing power from motor operated valves reduces risk of the valve changing state. This is supported by the PRA calculation results for the Single Pond Return Header Operation.

3. The discharge crossover motor operated valves will be open with power removed and tagged to prevent local operation. This reduces the risk of losing one train (from both units) flow path due to inadvertent closure. Removing power reduces the risk of the valve changing state.

4. The motor operated valves that isolate the NSWS discharge from discharging to Lake Wylie will remain powered such that they can be quickly opened from the Control Room to establish an alternate discharge path if needed. In addition, where the two valves are in series for the Auxiliary Building return to Lake Wylie, only one valve will remain closed to increase the reliability of establishing this flow path.

5. The support system for the NSWS Discharge piping associated with Train 1A in the Auxiliary Building will be maintained such that stress levels are below the threshold for considering a pipe leak under the Pipe Rupture program. This ensures that for all sections where pipe ruptures are postulated that leaks can be isolated with the NSWS continuing to operate with adequate equipment to support shutdown of both units. Catawba NSWS piping stress calculations RNG, RNH, and RNE have been completed to indicate the requirement to maintain this low stress level. To reduce stress at the 1A Component Cooling (KC) Heat Exchanger piping return nozzle location, a 1/4" thick reinforcing pad will be added to the existing reinforcing pad per a plant modification. The 1/4" reinforcing pad must be installed prior to entering NSWS Single Pond Return Header Operation.

6. NSWS Flow Balance testing will take place prior to entering Single Pond Return Header Operation. This will ensure the NSWS is capable of providing adequate cooling water flow to support LOCA loads on one unit, concurrent with the shutdown loads of the other unit -

while assuming the most limiting single failure that is loss of one EDG and its associated NSWS Pump.

7. Unit 1 and Unit 2 NSWS Non-Essential Headers will be isolated as part of the alignment for NSWS Single Pond Return Header Operation. This will ensure that the NSWS Essential Headers and EDGs receive their required flow and is consistent with the flow modeling that was performed. The specific valves that will be closed with power removed are:

1RN49A or 1RN50B and 1RN51A or 1RN52B (U1 NSWS Non-Essential Header Isolation Valves)

2RN49A or 2RN50B and 2RN51A or 2RN52B (U2 NSWS Non-Essential Header Isolation Valves)

U.S. Nuclear Regulatory Commission CNS-17-014 Enclosure, Page 46

8. Prior to entering Single Pond Return Header Operation, makeup to the SNSWP may be required to ensure that the SNSWP level is at elevation of 573.0 ft. or greater at the time when Single Pond Return Header Operation is entered.

3.1.2.14 NSWS Flow Analysis during Design Basis Events in Single Pond Return Header Operation Entry into NSWS Single Pond Return Header operation is designed for conditions where all four trains of NSWS and EDGs are operable. These conditions enhance the reliability of the NSWS while it is in the 30 day CT and are assumed in the PRA model that qualified the TS CT based on the calculated risk.

Flow modeling for the NSWS Single Pond Return Header Operation is done under the most limiting conditions for the NSWS, which is a LOCA on one unit with the other unit being concurrently shutdown and a loss of one EDG and its associated NSWS Pump. Although it is not required to postulate a failure while in a TS Action statement, this is conservative and supports the argument that Single Pond Return Header Operation can supply adequate flow to required essential equipment and a NSWS that meets Single Failure and Pipe Rupture Criteria, with the following exceptions:

The removal of one Pond Return Header removes redundancy for NSWS Return piping to the SNSWP.

A failure of a single valve in the in-service flow path of NSWS (1RN58B or 1RN63A) can result in the temporary loss of both NSWS essential headers on both units, until the NSWS is realigned to Lake Wylie. However, the likelihood, consequences, and Operations response to this failure is very similar to the current accident response for the NSWS. The NSWS is aligned to Lake Wylie for normal operation and all design basis events, unless there is a loss of Lake Wylie. Note that valves 1RN-58B and 1RN-63A open on Sp signal, so events involving Hi-Hi Containment pressure (i.e. Large Break LOCA, Feed line Break in Containment, Steam line Break in Containment) would automatically open a discharge flow path to the SNSWP. For normal operation and events that do not generate an Sp signal (which automatically opens a flow path to the SNSWP), the NSWS discharges through the in series Lake Isolation Valves 1RN-57A and 1RN-843B. Either of these valves could fail closed and result in a loss of the single in-service NSWS flow path, which would require a realignment to the SNSWP to reestablish a discharge flow path for the NSWS. Similarly, the response to a spurious failure to the closed position of the in-service SNSWP isolation valve 1RN-58B or 1RN-63A is to realign to Lake Wylie to reestablish a discharge flow-path for the NSWS.

A failure of a single valve, 1(2)RN846A or 1(2)RN848B, in the in-service NSWS pond return flow path from the EDGs, can result in the temporary loss of NSWS cooling water to both trains of EDGs for the affected unit. EDG cooling water flow alarms in the Control Room and EDG Annunciator panel alarms will alert the Operators to the failure. Procedures will direct mitigating actions, including re-alignment to Lake Wylie, to reestablish a discharge flow path.

Note that the conditions for supporting the LOCA loads on one unit while a concurrent orderly shutdown on the other unit support GDC 5 requirements for sharing of systems between units.

System conditions for flow modeling consists of 16 cases listed in the table below. These cases are run using the NSWS flow model with equipment throttle valves adjusted to provide minimum

U.S. Nuclear Regulatory Commission CNS-17-014 Enclosure, Page 47 flow (with margin) to supplied equipment. Results indicate the system is capable of providing adequate flow and pressure to all design loads. These predicted flow rates are documented in Catawba calculation CNC-1223.24-00-0072 (RN Single Pond Return Header Design Basis).

Flow balance testing will be performed prior to entry into Single Pond Return Header Operation.

This ensures that the system will operate as predicted while in Single Pond Return Header Operation.

Conditions for the NSWS flow models are as follows:

Single Case Initial Condition(s) Event Comments Failure 1-4 1. Single Pond Unit 1 LOCA Case 1-4: 1. Unit 2 Concurrent Cooldown to Mode 5.

Return Header - (Sp) Each EDG 2. Unit 1 NSWS Loads:

NSWS Pond Return LOOP Both with assoc - 1A, 1B KC Hxs Header A OOS Units NSWS - 1A, 1B NS Hxs

2. Unit 1 - Mode 1 Pump - 1A, 1B KD Hxs

3. Unit 2 - Mode 1 - 1A, 1B CA Assured Makeup

4. NSWS suction and - 1A or 1B YC Chiller discharge are aligned 3. Unit 2 NSWS Loads:

to the SNSWP - 2A, 2B KC Hxs

- 2A, 2B KD Hxs

- 2A, 2B CA Assured Makeup

4. Backwash flow on 3 NSWS pumps

5. Cases:

Case 1: 1A EDG & NSWS Pump Fail Case 2: 1B EDG & NSWS Pump Fail Case 3: 2A EDG & NSWS Pump Fail Case 4: 2B EDG & NSWS Pump Fail 5-8 1. Single Pond Unit 1 LOCA Case 5-8: 1. Unit 2 Concurrent Cooldown to Mode 5.

Return Header - (Sp) Each EDG 2. Unit 1 NSWS Loads:

NSWS Pond Return LOOP Both with assoc - 1A, 1B KC Hxs Header B OOS Units NSWS - 1A, 1B NS Hxs

2. Unit 1 - Mode 1 Pump - 1A, 1B KD Hxs

3. Unit 2 - Mode 1 - 1A, 1B CA Assured Makeup

4. NSWS suction and - 1A or 1B YC Chiller discharge are aligned 3. Unit 2 NSWS Loads:

to the SNSWP - 2A, 2B KC Hxs

- 2A, 2B KD Hxs

- 2A, 2B CA Assured Makeup

4. Backwash flow on 3 NSWS pumps

5. Cases:

Case 5: 1A EDG & NSWS Pump Fail Case 6: 1B EDG & NSWS Pump Fail Case 7: 2A EDG & NSWS Pump Fail Case 8: 2B EDG & NSWS Pump Fail

U.S. Nuclear Regulatory Commission CNS-17-014 Enclosure, Page 48 Single Case Initial Condition(s) Event Comments Failure 9-12 1. Single Pond Unit 2 LOCA Case 9-12: 1. Unit 2 Concurrent Cooldown to Mode 5.

Return Header - (Sp) Each EDG 2. Unit 1 NSWS Loads:

NSWS Pond Return LOOP Both with assoc - 1A, 1B KC Hxs Header A OOS Units NSWS - 1A, 1B NS Hxs

2. Unit 1 - Mode 1 Pump - 1A, 1B KD Hxs

3. Unit 2 - Mode 1 - 1A, 1B CA Assured Makeup

4. NSWS suction and - 1A or 1B YC Chiller discharge are aligned 3. Unit 2 NSWS Loads:

to the SNSWP - 2A, 2B KC Hxs

- 2A, 2B KD Hxs

- 2A, 2B CA Assured Makeup

4. Backwash flow on 3 NSWS pumps

5. Cases:

Case 9: 1A EDG & NSWS Pump Fail Case 10: 1B EDG & NSWS Pump Fail Case 11: 2A EDG & NSWS Pump Fail Case 12: 2B EDG & NSWS Pump Fail 13-16 1. Single Pond Unit 2 LOCA Case 13-16: 1. Unit 2 Concurrent Cooldown to Mode 5.

Return Header - (Sp) Each EDG 2. Unit 1 NSWS Loads:

NSWS Pond Return LOOP Both with assoc - 1A, 1B KC Hxs Header B OOS Units NSWS - 1A, 1B NS Hxs

2. Unit 1 - Mode 1 Pump - 1A, 1B KD Hxs

3. Unit 2 - Mode 1 - 1A, 1B CA Assured Makeup

4. NSWS suction and - 1A or 1B YC Chiller discharge are aligned 3. Unit 2 NSWS Loads:

to the SNSWP - 2A, 2B KC Hxs

- 2A, 2B KD Hxs

- 2A, 2B CA Assured Makeup

4. Backwash flow on 3 NSWS pumps

5. Cases:

Case 13: 1A EDG & NSWS Pump Fail Case 14: 1B EDG & NSWS Pump Fail Case 15: 2A EDG & NSWS Pump Fail Case 16: 2B EDG & NSWS Pump Fail Notes

1. In the table above:

OOS is Out of Service

KC is the Component Cooling System

KD is the Emergency Diesel Generator Jacket Water Cooling System

NS is the Containment Spray System

CA is the Auxiliary Feedwater System

YC is the Control Room Chilled Water System

2. Entry in the NSWS Single Pond Return Header Operation is restricted to having both trains of EDGs, NSWS Pumps, and NSWS supplied equipment on both units operable. If there is subsequent inoperability of NSWS and entry into TS 3.7.8 Condition A, the procedural option to align the NSWS in the "One Pump Flow Balance" shall not be used. Procedurally, this option requires closing NSWS Supply Header Crossover Valves, which is not modeled in the NSWS flow model for Single Pond Return Header Operation.

U.S. Nuclear Regulatory Commission CNS-17-014 Enclosure, Page 49

3. Performing scheduled, planned or maintenance that renders EDGs, NSWS pumps, or NSWS supplied equipment both inoperable and unavailable on either train of NSWS of either unit is prohibited while NSWS is aligned for Single Pond Return Header Operation with one exception. For the EDGs, a monthly periodic test is performed to confirm operability. Prior to starting the EDG, a "bar and roll" of the EDG is performed. This renders the EDG inoperable but available, and is allowed while the NSWS is aligned for Single Pond Return Header Operation.

4. It should be noted that when NSWS is aligned for Single Pond Return Header operation that the NSWS return header crossovers will be open with power removed and therefore will not auto-close on a NSWS low-low pit level signal or transfer to Aux Shutdown Panel. This is allowable since the three remaining NSWS Pumps have adequate capacity to supply the remaining cooling water demands of the three NSWS essential headers and three EDGs.

3.2.1.15 Conclusion The justification provided for the NSWS Single Pond Return Header Operation supports the position that while each Unit will be in an action statement for one NSWS pond return header, either train of NSWS is capable of providing adequate flow to essential loads to support design basis requirements. Measures have been identified and described to increase reliability of the NSWS in response to single failure and pipe rupture. With respect to single failure and pipe rupture response the NSWS has a high level of reliability that in consideration of the PRA meets the intent of GDC 2, 4, 5, 44, and 45 for the 30 day CT.

3.2 Deterministic and Risk Assessment of Proposed NSWS 30-Day CT 3.2.1 PRA Scope The change in risk associated with the requested 30-day CT for a single Standby Nuclear Service Water Pond (SNSWP) return header out of service has been evaluated for both Catawba units in accordance with the guidance of RGs 1.174 and 1.177 (Refs. 3.2.8.1 and 3.2.8.2). Hazard groups were evaluated to determine which sources of risk could affect the decision, and the risk from such hazards was assessed quantitatively and qualitatively using a PRA that has been assessed against the Capability Category II Supporting Requirements (SRs) in the existing PRA standards as well as Reg. Guide 1.200 (see Attachment 4).

The Catawba PRAs currently model internal events for Core Damage Frequency (CDF) and LERF, internal flooding, high winds and fire. These models have been peer reviewed and the impact of the open findings has been evaluated.

Section 2.3.2 of RG 1.177 identifies the NRCs regulatory position on PRA scope, and states, in part:

in some cases, a PRA of sufficient scope may not be available. This will have to be compensated for by qualitative arguments, bounding analyses, or compensatory measures.

This section further states, in part:

The scope of the analysis should include all hazard groupsunless it can be shown that the contribution from specific hazard groups does not affect the decision.

U.S. Nuclear Regulatory Commission CNS-17-014 Enclosure, Page 50 RG 1.174 Section 2.3.1 further clarifies this concept:

A qualitative treatment of the missing modes and hazard groups may be sufficient when the licensee can demonstrate that those risk contributions would not affect the decision; that is, they do not alter the results of the comparison with the acceptance guidelines 3.2.2 PRA Quality / Technical Adequacy The Catawba internal events model (excluding LERF) received a peer review against the requirements of the ASME/ANS RA-Sa-2009 PRA Standard with consideration for identified changes from Addendum A to Addendum B of the PRA Standard (ASME/ANS RA-Sb-2013) and Regulatory Guide 1.200, using the process defined in Nuclear Energy Institute (NEI) 05-04 (Ref. 3.2.8.3) in December 2015. The internal flood PRA and LERF models received focused scope peer reviews against the ASME/ANS RA-Sa-2009 PRA Standard (Ref. 3.2.8.4) and Regulatory Guide 1.200 (Ref. 3.2.8.5) in 2012. Additionally, an F&O closure effort was completed in July 2017, for internal flood and LERF, to validate the F&O closure process met the Appendix X requirements (Ref. 3.2.8.10).

The Catawba Fire PRA model received a peer review against the requirements of the ASME/ANS RA-Sa-2009 PRA Standard and Regulatory Guide 1.200, using NEI 07-12 (Ref.

3.2.8.6) in July 2010. The high winds PRA model received a peer review against the requirements of the ASME/ANS RA-Sb-2013 PRA Standard (Ref. 3.2.8.7) and Regulatory Guide 1.200, in August 2013, using NEI 05-04, after a comparison between the ASME/ANS RA-Sa-2009 and ASME/ANS RA-Sb-2013 standards showed no substantive differences for high winds.

3.2.2.1 Internal Events, CDF and LERF The scope of this December 2015 Peer Review included all internal events PRA requirements except LERF and Internal Flooding models. There were 8 internal events PRA Findings which were considered to be open after the December 2015 Peer Review (Table 1 of Ref. 3.2.8.8) and the 2017 independent F&O closure technical review. This included F&Os for which the SRs were met at Capability Category II or rated at Capability Category I. Reference 3.2.8.9 provides a resolution to each of these F&O Findings. These are discussed along with their impact to the 30-day CT LAR in Section 4.1 of Attachment 4.

Five LERF SRs have been assessed as meeting only Capability Category I (Appendix D of Reference 3.2.8.8). Each of these SRs, along with their impact to the 30-day CT LAR, is discussed in Section 4.2 of Attachment 4.

3.2.2.2 Internal Flooding All SRs associated with the F&Os were judged to have been met at CC I/II/III. There are no open findings associated with the internal flooding PRA model as noted in Section 4.3 of .

U.S. Nuclear Regulatory Commission CNS-17-014 Enclosure, Page 51 3.2.2.3 High Winds The High Winds PRA model received a peer review against the requirements of the ASME/ANS RA-Sb- 2009 PRA Standard and Reg. Guide 1.200, using the process outlined NEI 05-04 in August 2013. A comparison between the ASME/ANS RA-Sa-2009 and ASME/ANS RA-Sb-2013 standards (Ref. 3.2.3.11) showed no substantive differences for High Winds. There were no SRs rated as Capability Category I or Not Met; however, the Peer Review identified 5 finding level F&Os. As shown in Section 4.4 of Attachment 4, the High Winds PRA has been updated to incorporate resolutions to these F&Os. The resolution of these F&Os have not been independently reviewed.

3.2.2.4 Fire The Catawba Fire PRA (FPRA) model received a peer review against the requirements in Section 4 of the ASME/ANS RA-Sa-2009 PRA Standard, using NEI 07-12 (Ref. 3.2.8.11) in July 2010. 20 finding level F&Os were resolved via updates to the FPRA and are documented in the FPRA calculations. Additionally, multiple RAIs were generated by the NRC during the review of the NFPA 805 LAR. A number of these RAIs referenced the F&Os from the FPRA Peer Review.

These RAI responses have been incorporated into the FPRA and are documented through various RAI responses, as well as the RAI 03 response which involved quantification of the FPRA after all RAI responses were incorporated. No new methods were introduced during the FPRA changes; therefore, no additional Peer Review is needed. Each of the open finding-level F&Os, along with their impact on the 30-day CT LAR, are discussed in Section 4.5 of .

3.2.3 Fault Tree Model Changes to Evaluate 30-day NSWS TS CT Catawba is submitting this LAR for a permanent change to TS 3.7.8; Nuclear Service Water System (NSWS), to add a new condition for increased completion time for LCO 3.7.8 when the system is in the SNSWP single pond return header alignment. While the NSWS is operating in this configuration, one of the shared train-related return headers is removed from service to allow internal inspections and modifications of the NSWS Pond Return buried piping between the Aux Bldg. and the discharge to the SNSWP. The required NSWS flow to all shared, train-related safety equipment is discharged through the remaining OPERABLE SNSWP return header via train-related crossover valves.

It is intended to utilize the NSWS Single Pond Return alignment with both units in Mode 1.

There is no specific mode requirement for use of this alignment; however, the proposed T.S.

3.7.8 wording states that the spec is applicable to Modes, 1, 2, 3 and 4. Therefore, this assessment was performed considering at-power operation only.

In order to assess the changes in risk when only one SNSWP return header is available, several modifications to the internal events, internal flooding, fire and high winds models were required.

The NSWS will be pre-aligned to the SNSWP prior to entering the single pond return header condition. This means the normal supply and discharge flow paths from / to Lake Wylie will be isolated and normal flow will be provided via the SNSWP supply and return headers. Placing the NSWS in this configuration creates a more passive condition since potential failures due to an automatic swap from Lake Wylie to the SNSWP on low lake level are eliminated.

U.S. Nuclear Regulatory Commission CNS-17-014 Enclosure, Page 52 3.2.3.1 Model Change Overview The Catawba Internal Events model consists of separate models for each unit and accounts for multiple trains. For each unit model, similar changes were made to account for the single SNSWP return header configuration. For the High Winds, Fire and Internal Flooding models, the PRA model is a single unit model that generally assumes A-Train equipment is running with B-Train equipment in standby. By practice, maintenance is modeled as affecting the B-Train.

3.2.3.2 General Assumptions and Model Changes The following assumptions and model changes were applied globally to the PRA analysis:

For this analysis, the 1A NSWS pump was assumed to be running with the A SNSWP header in service and the B SNSWP return header in the 30-day CT. The NSWS is designed such that any one NSWS pump can supply flow to any of the four essential headers (Unit 1 A and B; Unit 2 A and B). Accordingly, the internal events, internal flooding, fire and high winds fault tree models were reconfigured to emulate operating conditions while in the 30-day CT.

The analysis assumes one entry per header train during a year (i.e., two entries into the Tech. Spec. per year).

As a condition for the CT, performing scheduled, planned or discretionary maintenance that renders EDGs, NSWS pumps or NSWS supplied equipment both inoperable and unavailable on either train of NSWS of either unit is prohibited while the NSWS is aligned for Single Pond Return Header Operation with one exception. For the EDGs, a monthly periodic test is performed to confirm operability. Prior to starting the EDG, a bar and roll of the EDG is performed. The renders the EDG inoperable but available, and is allowed while the NSWS is aligned for Single Pond Return Header Operation. For this reason, the corresponding maintenance events for these SSCs were set to 0 for the duration of this configuration.

Valves in the shared (single) flowpath will be prevented from repositioning. Manual valves will be physically locked in the open position. Motor-operated valves (MOVs) in the affected flow path(s) will be open with power removed and tagged to prevent local operation. Any MOVs with power removed were modeled as manual valves (and, for fire PRA modeling purposes, their corresponding ignition sources were removed).

MOVs that isolate the NSWS discharge to Lake Wylie will remain powered such that they can be quickly opened from the Control Room to establish an alternate discharge path if needed. One of these two "in-series" isolation valves will be open to increase the reliability of swapping discharge to Lake Wylie if the alternate discharge path is needed.

NSWS Return Header crossover valves will be open with power removed (and their corresponding ignition sources removed from the fire PRA), and therefore will not auto-close (on low-low NSWS suction pit level or transfer to the Aux. Shutdown Panel).

The normally-open supply valves from Lake Wylie to the NSWS pumps are isolated and the normally-closed supply valves from the SNSWP are open.

For HRA considerations, recovery actions are planned to address events such as swapping the NSWS suction and discharge back to Lake Wylie, restoring power to key MOVs for isolation and local operation of manual valves for isolation. The internal events, internal flooding and high winds analyses used a bounding analysis in which

U.S. Nuclear Regulatory Commission CNS-17-014 Enclosure, Page 53 such potential accident sequence recoveries were added to the model but were not credited (i.e., HRA events set to 1.0). This is a conservative assumption given information in the CT design basis document (Ref 3.2.8.14) that indicates significant inventory is available in the SNSWP such that draining down to the minimum pond level generally would take longer than 24 hrs. (This was the case for the internal events, flooding and high winds analyses. However, in assessing the fire CT risk results, it became necessary to quantify these values to address the spurious operation of MOVs.)

The use of FLEX equipment was NOT credited in the quantitative assessments.

3.2.3.3 Common Cause Failure (CCF) Evaluation The logic models for the 30-day CT configuration did not require any new common cause events because no new failure modes were added that required a CCF assessment.

3.2.4 Risk Results The internal events, internal flooding, high winds and fire models were quantified for CDF and LERF to determine the CDF, LERF, CCDP and CLERP that would result from the approval of the 30-day CT change. Three cases were evaluated for each hazard. The base case represents the current model of record (MOR) configuration for at-power operation. The CT case assesses the risk with the NSWS aligned in the single SNSWP return header configuration. Finally, the Non-CT case is similar to the base case except that it assesses the risk for normal at-power operation with the nominal component unavailability values applied over the time during the year when the NSWS is not in the CT configuration (for this analysis, the nominal 12-month unavailability of the NSWS and the EDGs is applied over a 10-month period). Sections 3.2.4.1 -

3.2.4.4 contain the CDF and LERF values for the base case, CT case and non-CT case.

Section 3.2.4.8 contains the ICCDP / ICLERP results. Section 3.2.4.9 contains the delta CDF and delta LERF results.

3.2.4.1 Internal Events Analysis The Internal Event CDF and LERF results for Unit 1, for the three cases described in Section 3.2.4 above, are shown in Table 3.2.4.1-1. The Unit 2 results are shown in Table 3.2.4.1-2.

Table 3.2.4.1 Unit 1 Internal Events Results Case CDF ( / yr) LERF ( / yr)

Base Case 7.23E-06 3.42E-07 CT Configuration 7.50E-06 3.39E-07 Non-CT 7.28E-06 3.50E-07 Table 3.2.4.1 Unit 2 Internal Events Results Case CDF ( / yr) LERF ( / yr)

Base Case 7.33E-06 4.01E-07 CT Configuration 7.39E-06 3.68E-07 Non-CT 7.38E-06 4.14E-07 The increase in CDF is primarily due to new accident sequences in whichNSWSisolation valves spuriously transfer position without operator recovery (a conservative assumption as listed above). This risk increase is somewhat abated by no activities involving maintenance

U.S. Nuclear Regulatory Commission CNS-17-014 Enclosure, Page 54 unavailability being performed on the NSWS and the EDGs during the CT. The decrease in LERF is due solely to no activities involving maintenance unavailability being performed on the NSWS and the EDGs during the CT.

Importance measures were obtained for the 30-day CT case as a way of determining Tier 2 measures. Results were captured for basic events with a Fussell-Vesely (F-V) value greater than or equal to 0.005 and a Risk Achievement Worth (RAW) greater than or equal to 2. Tier 2 measures are presented in Section 3.2.6.

3.2.4.2 Internal Flooding Analysis The Internal Flooding CDF and LERF results for Unit 1, for the three cases described in Section 3.2.4 above, are shown in Table 3.2.4.2-1. The Unit 2 results are shown in Table 3.2.4.2-2.

Table 3.2.4.2 Unit 1 Internal Flooding Results Case CDF ( / yr) LERF ( / yr)

Base Case 4.08E-05 7.64E-07 CT Configuration 4.05E-05 7.32E-07 Non-CT 4.08E-05 7.71E-07 Table 3.2.4.2 Unit 2 Internal Flooding Results Case CDF ( / yr) LERF ( / yr)

Base Case 4.08E-05 7.64E-07 CT Configuration 4.05E-05 7.32E-07 Non-CT 4.08E-05 7.71E-07 The decrease in CDF and LERF stems mainly from the lack of maintenance unavailability on the NSWS and the EDGs during the CT.

Importance measures were obtained for the 30-day CT case as a way of determining Tier 2 measures. Results were captured for basic events with a Fussell-Vesely (F-V) value greater than or equal to 0.005 and a Risk Achievement Worth (RAW) greater than or equal to 2. Tier 2 measures are presented in Section 3.2.6.

3.2.4.3 Fire Analysis The Fire CDF and LERF results for Unit 1, for the three cases described in Section 3.2.4 above, are shown in Table 3.2.4.3-1. The Unit 2 results are shown in Table 3.2.4.3-2.

Table 3.2.4.3 Unit 1 Fire Results Case CDF ( / yr) LERF ( / yr)

Base Case 4.28E-05 4.51E-06 CT Configuration 3.88E-05 4.02E-06 Non-CT 4.41E-05 4.63E-06 Table 3.2.4.3 Unit 2 Fire Results

U.S. Nuclear Regulatory Commission CNS-17-014 Enclosure, Page 55 Case CDF ( / yr) LERF ( / yr)

Base Case 4.38E-05 4.37E-06 CT Configuration 4.01E-05 4.01E-06 Non-CT 4.52E-05 4.48E-06 The decrease in CDF and LERF stems mainly from the lack of maintenance unavailability on the NSWS and the EDGs during the CT as well as a reduction in the number of ignition sources when power is removed from certain MOVs.

Importance measures were obtained for the 30-day CT case as a way of determining Tier 2 measures. Results were captured for basic events with a Fussell-Vesely (F-V) value greater than or equal to 0.005 and for a Risk Achievement Worth (RAW) greater than or equal to 2. Tier 2 measures are presented in Section 3.2.6.

3.2.4.4 High Winds Analysis The High Winds CDF and LERF results for Unit 1, for the three cases described in Section 3.2.4 above, are shown in Table 3.2.4.4-1. The Unit 2 results are shown in Table 3.2.4.4-2.

Table 3.2.4.4 Unit 1 High Winds Results Case CDF ( / yr) LERF ( / yr)

Base Case 8.13E-06 2.16E-06 CT Configuration 6.29E-06 1.53E-06 Non-CT 8.58E-06 2.30E-06 Table 3.2.4.4 Unit 2 High Winds Results Case CDF ( / yr) LERF ( / yr)

Base Case 7.43E-06 2.19E-06 CT Configuration 5.94E-06 1.54E-06 Non-CT 7.81E-06 2.34E-06 The decrease in CDF and LERF is attributed primarily to the lack of maintenance unavailability on the NSWS and the EDGs during the CT.

Importance measures were obtained for the 30-day CT case as a way of determining Tier 2 measures. Results were captured for basic events with a Fussell-Vesely (F-V) value greater than or equal to 0.005 and for a Risk Achievement Worth (RAW) greater than or equal to 2. Tier 2 measures are presented in Section 3.2.6.

3.2.4.5 External Flooding Risk Analyses have been performed that show the Key Safety Functions for Catawba, including the NSWS and EDGs, are not adversely affected by an external flood event. The Local Intense Precipitation (LIP) event is the controlling flood event (Ref. 3.2.8.24). Curbing has been installed around the EDG Bldg. stairwell access doors to address the impact of the LIP event. The NSWS pump structure floor is situated at Elev. 600, which is above the maximum PMP elevation, and the NSWS pump motors are mounted on top of the pump discharge head (> Elev. 600). The results of the LIP analyses, the installation of the EDG Bldg. stairwell access door curbing and

U.S. Nuclear Regulatory Commission CNS-17-014 Enclosure, Page 56 the elevation of the NSWS pump structure support the conclusion that the risk of external event flooding from a PMP event is negligible.

Specifically in regards to the NSWS Single Pond Return Header alignment, the section of return header piping that is isolated for inspections, maintenance, or modifications consists almost completely of buried piping outside of the boundary of any buildings. Personnel access points to this piping will only be through open manways located in the yard areas (per Ref. 3.2.8.14). In the unlikely event of a failure of an upstream isolation valve where the valve suddenly goes to the open position, most of the NSWS discharge flow would exit the open manways to the yard.

Per Ref. 3.2.8.14, the two possibilities to consider, flooding in the yard and diversion of return flow to the SNSWP, are determined to be acceptable as follows:

NSWS flow modeling shows that all flow released through open manways would flow out of the manways closest to the powerblock (M-5-1, M-6-1, and M-7-1). Appendix A of Ref. 3.2.8.14 documents application of the NSWS Flow model with the addition of open manways discharging to the yard. The resulting maximum predicted outflow of 18,591 gpm is based on the maximum flow setting of the throttle valves and 4 pump operation. This flow is well within the station's yard drainage system flood handling capacity as described in Appendix F of Ref. 3.2.8.14 and will not affect the ability to safely shutdown the plant by normal procedures.

NSWS discharge flow from open manways would not return to the SNSWP and therefore would reduce pond inventory. However, adequate margin for emergent leakage from the SNSWP is available to cover the volume of water that would be diverted with a leak of this nature and a 116 minute response time (per Appendix B of Ref. 3.2.8.14). Isolation valves are available in the Aux Bldg and EDG buildings that can isolate flow to the out of service header and still retain flowpaths to at least one complete return header of essential equipment per unit. A response time of 30 minutes is reasonable and is assumed by the Catawba Pipe Rupture Specification (Ref. 3.2.8.15) during similar responses to postulated pipe ruptures.

Therefore, per the above analysis, consideration of external flooding is not a factor for this assessment.

3.2.4.6 Seismic Risk In this section, consideration is given to a qualitative assessment of the impact of the plant with the SNSWP single return header configuration on the seismic CDF. As indicated in the NSWS Design Basis Document (Ref. 3.2.8.16), Lake Wylie is the non-safety source of cooling water and the SNSWP is the assured source. Thus, the seismic PRA (SPRA) analysis performed for the Catawba IPEEE submittal (Ref. 3.2.8.17) assumes a loss of Lake Wylie following a seismic event. In this analysis, structures with a median seismic capacity greater than 2.5g are screened out of the SPRA logic model due to their extremely low probability of failure. This includes the NSWS pump structure as well as the SNSW intake and discharge structure. Similarly, components with a median capacity greater than 2.0g were eliminated from the SPRA logic model. Among those screened components were the NSWS pumps and valves as they have relatively high seismic capacities due to the conservatism in the qualification process coupled with the inherent ruggedness of the process and electrical equipment. The NSWS piping is also screened out of the SPRA logic model as it is one of Category I piping systems that are generally considered seismically rugged and assigned a high median capacity in excess of 2.0g.

U.S. Nuclear Regulatllly Commlsalon CNS-17-014 EndOBura, Paga 57 Fer Ille rec&lned lllnldlll'M and COITlllOnenls In Ille !ogle model, the 1'8&\11111111 SCDF eallmates range frDm 1.7E-5iyaar to 2.8E~ using the ahpla llWll&ga method (Raf. 3.2.8.18).

In l'Bllponm to Iha NRC 10 CFR 60.54(1) lattm (Raf. 3.2.8.19) follat.1ng Iha accklant al the Fi*ushima Dai-ichi nudas pll!l'lllr plant 11111ulting fnlm tha March 11, 2011 Great Tohoku Earthquake, Duke EnetQy submitted the updated aeiamic hazards and new Ground Motion Re&ponge Speclrum (GMRS) ID tlle NRC for~ (Ref, 3.2.8.20). 1be llgur. beklw choWI a comparlaon of Iha , _ GMRS to SSE accelaratlon rea~ spectra. From that llgura, Iha dnlgn baala SSE Im BB liB the GMRS below 6.6 Hz. and the GMRS bagl1111 to Im BB d Iha CabrMJa SSE abova 5.5 Hz. Tha peak accalandion al the naw G~ ia 0.75g at 30 Hz. Tha 11/geet ratio of the GMR~to-SSE speclral ecceleration in Ille 1-10 Hz: frequency rarce i8 1.91.

Grwnd mollons at leYels -.. ID two !hes Ille SSE are upec:Ced to ~uce o~ a 11111111 probabllty al fallure for safety.nllaled SSCa due to CC111aervathle design prac!lcea. In the high fraquanc:y range graatar lhlm 10 Hz, atructund clapl-11111111 In this fraquanc:y range ara amal and ara conaldarad non'41amaglng.

Catawba Nuclear Station SSE vs. GMRS o.t

_ o.a

.!! '~

c 0

0.7 I ' '

I f! 0.6 I I

I

~ 0.5 I I

I

\ - - <:>Mlt5

<i: o.A I I - s>C

~i.I O.l I VI 0.2 I I

I I

0. 1 l --- --- --,, '

0 - -

0.1 LO 10.0 100.0 Frequency (Hz)

In fact, with Ille same eimple a.eiage melllod used along with the new eeilmic hazards (i.e.,

GMRS), the eetlmated SCDF la 2.8E-5/yr (Rel. 3.2.8.18). Thia la within Iha prevtoua range of eatlllllllBB d88Cllbad above. Thia Indicates that Catawba has msgln to withstand potential aar1hquakaa e>> BBdlng b allglnal daalgn baals and no mnoam axlabs l"llgSdlng adaquala prvlaction agaiml the rav aaismic haz:&Rla.

In addition, Catawba has Implemented FLEX Ordar EA-12-049 that raqulrae Iha llcanll89 ID dllVlllop, Implement, and maintain guidance and abatagl1111 ID maintain or restara core coollrci (Raf. 3.2.8.21). The R.EX mitigalirc abateoille utilize a th1'9&-pheead appioach. Upon tranlition to Pll... 2, portable equipment II l"8d ID maintain and/or eatabllsh required FLEX IUndlonl.

W*'1 supply fnlm 1he SNSWP la pn>lllded via portable dleffl drtwn pumps, whit eledl1cll

U.S. Nuclear Regulatory Commission CNS-17-014 Enclosure, Page 58 power is supplied to various components via portable diesel generators. This assures increased defense-in-depth and diversity for beyond-design-basis scenarios including seismic, that involve an extended loss of AC power and loss of normal access to the ultimate heat sink occurring simultaneously at all units on a site. Catawba plans to complete the seismic mitigating strategies assessment (MSA) in accordance with Ref. 3.2.8.22 (Appendix H), Path 4, by August 31, 2017.

Successful completion of the MSA Path 4 for Catawba will verify successful implementation of FLEX against the reevaluated seismic hazard. Since selected plant components were evaluated up to the GMRS-to-SSE ratio during ESEP (Ref. 3.2.8.23), there is high confidence in successful completion of the Seismic MSA Path 4. Therefore, consideration of the seismic risk impact while the plant is in the SNSWP single return header configuration is not a significant factor for this assessment.

3.2.4.7 Shutdown risk It is intended that the NSWS Single Pond Return header alignment will be utilized with both units in Mode 1. There is no specific mode requirement for use of this alignment; however, the proposed T.S. 3.7.8 wording states that the spec. is applicable to Modes, 1, 2, 3 and 4. The impact of the 30-day CT while online will thus have no negative impact on shutdown risk.

3.2.4.8 ICCDP / ICLERP for 30-Day CT The ICCDP and ICLERP for one entry into the T.S. are now computed. First, the delta CDF and LERF computed in the Sections 3.2.4.1 thru 3.2.4.4 above are tabulated below:

Table 3.2.4.8 Base Case Risk, All Hazards Internal Internal High Fire Total Risk Metric Events Flood Winds

( / yr.) ( / yr.)

( / yr.) ( / yr.) ( / yr.)

U1 CDF 7.23E-06 4.08E-05 4.28E-05 8.13E-06 9.90E-05 U1 LERF 3.42E-07 7.64E-07 4.51E-06 2.16E-06 7.78E-06 U2 CDF 7.33E-06 4.08E-05 4.38E-05 7.43E-06 9.94E-05 U2 LERF 4.01E-07 7.64E-07 4.37E-06 2.19E-06 7.73E-06 Table 3.2.4.8 30-Day CT Risk, All Hazards Internal Internal High Fire Total Risk Metric Events Flood Winds

( / yr.) ( / yr.)

( / yr.) ( / yr.) ( / yr.)

U1 CDF 7.50E-06 4.05E-05 3.88E-05 6.29E-06 9.31E-05 U1 LERF 3.39E-07 7.32E-07 4.02E-06 1.53E-06 6.62E-06 U2 CDF 7.39E-06 4.05E-05 4.01E-05 5.94E-06 9.39E-05 U2 LERF 3.68E-07 7.32E-07 4.01E-06 1.54E-06 6.65E-06

U.S. Nuclear Regulatory Commission CNS-17-014 Enclosure, Page 59 Table 3.2.4.8 Delta Risk, All Hazards Internal Internal High Fire Total Risk Metric Events Flood Winds

( / yr.) ( / yr.)

( / yr.) ( / yr.) ( / yr.)

U1 CDF 2.70E-07 -3.00E-07 -4.00E-06 -1.84E-06 -5.87E-06 U1 LERF -3.00E-09 -3.20E-08 -4.90E-07 -6.30E-07 -1.16E-06 U2 CDF 6.00E-08 -3.00E-07 -3.70E-06 -1.49E-06 -5.43E-06 U2 LERF -3.30E-08 -3.20E-08 -3.60E-07 -6.50E-07 -1.08E-06 Thus, for a 30-day CT, using the definitions for ICCDP and ICLERP presented earlier, U1 ICCDP = [(CDFCT Config. - CDFBaseline) x (30 days) / 365 days/yr]

= (-5.43E-06) x 30 / 365

= -4.82E-07 U1 ICLERP = [(LERFCT Config. - LERFBaseline) x (30 days) / 365 days/yr]

= (-1.16E-06) x 30 / 365

= -9.53E-08 U2 ICCDP = [(CDFCT Config. - CDFBaseline) x (30 days) / 365 days/yr]

= (-5.43E-06) x 30 / 365

= -4.46E-07 U2 ICLERP = [(LERFCT Config. - LERFBaseline) x (30 days) / 365 days/yr]

= (-1.08E-06) x 30 / 365

= -8.88E-08 Therefore, for both units, the ICCDP is less than 1E-6 and the ICLERP is less than 1E-07; therefore, these risk metrics meet the acceptance guidelines of Reg. Guide 1.177.

3.2.4.9 Delta CDF and LERF The change in CDF and LERF during the 10 month non-CT period can now be calculated by solving the base line risk models using the adjusted (Non-CT) values for NSWS and EDG unavailability. For Unit 1, the results are:

U.S. Nuclear Regulatory Commission CNS-17-014 Enclosure, Page 60 Table 3.2.4.9 Unit 1 Delta Risk, With Adjusted Maintenance Nominal Adjusted Nominal Adjusted Risk Baseline Baseline Delta CDF Baseline Baseline Delta LERF Component CDF CDF ( / yr.) LERF LERF ( / yr.)

( / yr.) ( / yr.) ( / yr.) ( / yr.)

U1 Internal 7.23E-06 7.28E-06 5.00E-08 3.42E-07 3.50E-07 8.00E-09 Events U1 Internal 4.08E-05 4.08E-05 0.00 7.64E-07 7.71E-07 7.00E-09 Flooding U1 Fire 4.28E-05 4.41E-05 1.30E-06 4.51E-06 4.63E-06 1.20E-07 U1 High 8.13E-06 8.58E-06 4.50E-07 2.16E-06 2.30E-06 1.40E-07 Winds U1 TOTAL 1.80E-06 U1 LERF 2.75E-07 CDF Similarly, for Unit 2, the results are:

Table 3.2.4.9 Unit 2 Delta Risk, With Adjusted Maintenance Nominal Adjusted Nominal Adjusted Risk Baseline Baseline Delta CDF Baseline Baseline Delta LERF Component CDF CDF ( / yr.) LERF LERF ( / yr.)

( / yr.) ( / yr.) ( / yr.) ( / yr.)

U2 Internal 7.33E-06 7.38E-06 5.00E-08 4.01E-07 4.14E-07 1.30E-08 Events U2 Internal 4.08E-05 4.08E-05 0.00 7.64E-07 7.71E-07 7.00E-09 Flooding U2 Fire 4.38E-05 4.52E-05 1.40E-06 4.37E-06 4.48E-06 1.10E-07 U2 High 7.43E-06 7.81E-06 3.80E-07 2.19E-06 2.34E-06 1.50E-07 Winds U2 TOTAL 1.83E-06 U2 LERF 2.80E-07 CDF

U.S. Nuclear Regulatory Commission CNS-17-014 Enclosure, Page 61 To estimate the actual change in CDF and LERF, the delta risk values for the CT and non-CT configurations must be added together proportionally:

U1 CDF = [(2/12 x U1 CDFCT ) + (10/12 x U1 CDFnon-CT )]

= [(2/12 x -5.87E-06 / yr ) + (10/12 x 1.80E-06 / yr)]

= 5.22E-07 / yr.

U1 LERF = [(2/12 x U1 LERFCT ) + (10/12 x U1 LERFnon-CT )]

= [(2/12 x -1.16E-06 / yr ) + (10/12 x 2.75E-07 / yr)]

= 3.58E-08 / yr.

U2 CDF = [(2/12 x U2 CDFCT ) + (10/12 x U2 CDFnon-CT )]

= [(2/12 x -5.43E-06 / yr ) + (10/12 x 1.83E-06 / yr)]

= 6.20E-07 / yr.

U2 LERF = [(2/12 x U2 LERFCT ) + (10/12 x U2 LERFnon-CT )]

= [(2/12 x -1.08E-06 / yr ) + (10/12 x 2.80E-07 / yr)]

= 5.33E-08 / yr.

Thus, for both units, the delta CDF is less than 1E-06 / yr. and the delta LERF is less than 1E-07 / yr. and thus meet the acceptance guidelines of Region III of RG 1.174 for a very small risk increase. Per Figures 4 and 5 of Reg. Guide 1.174, these changes in risk are in Region III which, per the Reg. Guide, will be considered regardless of whether there is a calculation of the total CDF and LERF (Region III).

3.2.5 PRA Model Configuration and Control Program The PRA is maintained and updated such that its representation of the as-built, as-operated plant is sufficient to support the applications for which it is used. Duke Energy maintains procedures that evaluate and prioritize changes in PRA inputs as well as address discovery of new information that could affect the PRA.

The PRA model is reviewed whenever plant accident response characteristics are changed.

Any identifiable plant change is analyzed for its risk significance. This includes plant physical modifications, changes to Emergency or Abnormal Procedures, as well as Technical Specifications and Selected Licensee Commitment changes. Additionally, all open PRA unincorporated change items are reviewed prior to the start of an application for their impact on that application. There are no open items that have any impact on the NSWS 30-day day CT application as shown below:

U.S. Nuclear Regulatory Commission CNS-17-014 Enclosure, Page 62 Tracking Impact Brief Disposition (If Risk Impact ID Description Description N/A)

C-07-0016 Medium Add alternate This would be a risk reduction because of adding a new This issue would feedwater way to get feedwater to the steam generators. The actual be a modeling makeup line to risk reduction is dependent on the procedure steps and enhancement /

each S/G. operator actions required to supply this new source of risk reduction alternate feedwater. To estimate it a new event (G003), and does not (Reference No feedwater flow from alternate SG Blowdown supply, affect any Letter to NRC was added under gates F100A, F101B, F100C, and applications.

2/26/07) F100D with a value of 1.0. The model was solved and recovered and the base case PRA values were obtained.

Then G003 was changed from 1.0 to 0.5 to estimate a failure probability of 50%. The delta CDF obtained was 3.9E-06/yr and the delta LERF obtained was 2.4E-07/yr.

This falls into the medium range.

C-14-0005 Medium Model ability to Changes have been made to the cr3b model, to allow the This issue would cross connect second MDEFWP to fill any given SG. Existing gates be a modeling Motor Driven F105A [No Flow MDP1A to SG1A] and F105C [No Flow enhancement/ris CA Pumps MDP1B to SG1C] were used as surrogates for flow from k reduction and (MDCAPs), per MDEFWP 1A and MDEFWP 1B, respectively, to any of does not affect PIP G the other SGs. A Human Error, FCAXCONDHE [CA any applications.

1294. Cross Connect Failure] was also added to estimate the probability that the operator fails to cross connect the EFW headers. The change in CDF in the Internal Events was approximately 5E-07. Additional benefits are expected in the Fire PRA model and the Seismic PRA model. Thus, it is expected to border on low to medium impact.

C-15-0003 Medium Fire PRA only - The exact CDF reduction has not been calculated, This modification Changes however, previous fire risk assessments had identified this is estimated to provided by failure as a significant risk contributor for fires in Fire Area result in a EC 111944 will 48. Therefore, this change form is conservatively decrease in CDF prevent categorized as MEDIUM. and LERF for the spurious fire PRA.

operation of *Note: This issue was identified and documented in PIP valve 2CA50A. C-13-01779 which addresses this and other fire risk Therefore, the issues including the development of appropriate risk assessment modification. A new PIP is not being written for this for the CT would "Medium" change because the issue specifically and not be impacted exclusively pertains to the fire risk analysis and would not in a negative affect any other PRA applications. manner.

C-15-0006 Medium Fire PRA Only: These cables were originally evaluated in Section 5.3 of This modification EC 112410 will the CNS Fire PRA Application Calc CNC 1535.00 is estimated to modify cables 0113. The evaluation modified the Zone Tag table in the result in a and cable FRANC to delete valve 1SA VA0145 from fire area 15. significant routing to This mod accomplishes this change which results in an decrease in CDF eliminate hot estimated CDF and LERF risk reduction of 5.7E 06/ rx yr and LERF for the short failures and 1.3E 07/ rx yr, respectively. Therefore, this model fire PRA.

of valve 1SA- change is characterized as MEDIUM.

145 for the Therefore, the TDCAP. This change was identified as part of the NFPA-805 risk assessment development process and is already addressed for the for the CT would supporting FRE calculation. No other PRA applications not be impacted outside of NFPA-805 are affected by this change. in a negative Therefore, there are no updates required to other existing manner.

PRA calculations or programs. Completion of the modification is being tracked separately for NFPA-805 implementation to meet regulatory commitments.

U.S. Nuclear Regulatory Commission CNS-17-014 Enclosure, Page 63 3.2.6 Tier 2 Component Evaluation Reg. Guide 1.177 defines Tier 2 of the NRC staffs three-tiered approach for evaluating the risk associated with proposed TS CT changes as the identification of potentially high-risk configurations that could exist if equipment, in addition to that associated with the change, were to be taken out of service simultaneously or other risk-significant operational factors, such as concurrent system or equipment testing, were also involved. The objective of this part of the evaluation is to ensure that appropriate restrictions on dominant risk-significant configurations associated with the change are in place. Based on SSCs assumed to be available in the PRA analysis, the NSWS system and the EDGs should be not have maintenance performed on them when the plant is in the 30-day CT that renders these SSCs as both inoperable and unavailable. Additionally, importance metrics were evaluated to determine any other components which the plant should avoid removing from service during the CT. These importance measures (Risk Achievement Worth and Fussell-Vesely) during the extended CT provide insights into what equipment should remain available during the extended CT. The SSCs whose unavailability should be avoided during the CT based on test and maintenance Fussell-Vesely and Risk Achievement Worth importance measures are given in Table 3.2.6-1:

Table 3.2.6 Tier 2 SSCs SSC Criteria Reason Protected in PRA risk NSWS Maintaining service water availability assessment Emergency Diesel Protected in PRA risk Maintaining ac power sources Generators (EDGs) assessment IE (CDF / LERF)

IF (CDF / LERF) Maintaining cooling to important accident Component Cooling System HW (CDF) mitigation systems Fire (CDF)

IE (CDF / LERF)

IF (CDF / LERF) Provide secondary side heat removal Auxiliary Feedwater System HW (CDF) capability Fire (CDF / LERF)

Instrument Air System IE (CDF / LERF) Support operation of AOVs / instruments IE (CDF / LERF)

Standby Shutdown Facility Provide alternate secondary side heat IF (CDF / LERF)

(SSF) removal capability and RCP seal injection Fire (CDF / LERF)

Residual Heat Removal IE (CDF)

Provide residual heat removal System Fire (CDF)

IE (CDF / LERF)

IF (CDF / LERF) Maintaining cooling to important accident 4160V ac Essential Power HW (CDF / LERF) mitigation systems Fire (CDF / LERF)

Engineered Safeguards IE (LERF) Provide containment isolation function Features Actuation System Switchyard BTP 8-8 Maintaining availability of off-site power 3.2.7 Compensatory Actions As mentioned earlier in this analysis, in addition to the planned operator recovery actions, planned or discretionary maintenance that renders one or more NSWS pumps and/or the

U.S. Nuclear Regulatory Commission CNS-17-014 Enclosure, Page 64 associated EDGs inoperable on either train of NSWS is prohibited while in the Single Pond Return Header alignment.

NSWS flow balance testing will take place prior to entering Single Pond Return Header operations. This will ensure the NSWS is capable of providing adequate cooling water flow to support LOCA loads on one unit concurrent with the shutdown loads of the other unit - while assuming the most limiting single failure which is loss of one EDG and its associated NSWS Pump.

3.2.8 References 3.2.8.1 USNRC 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.

3.2.8.2 USNRC Regulatory Guide 1.177, An Approach for Plant-Specific, Risk-Informed Decision Making: Technical Specifications, Revision 1.

3.2.8.3 Process for Performing Internal Events PRA Peer Reviews Using the ASME/ANS PRA Standard, NEI 05-04, Revision 2, Nuclear Energy Institute, November 2008.

[ML083430462]

3.2.8.4 ASME/ANS RA-Sa-2009, Addenda to ASME/ANS RA-S-2008 Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, ASME and the American Nuclear Society, February 2009.

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

3.2.8.6 Fire Probabilistic Risk Assessment (FPRA) Peer Review Process Guidelines, NEI 07-12, Revision 1, Nuclear Energy Institute, June 2010. [ML102230070]

3.2.8.7 ASME/ANS RA-Sb-2013, Addenda to ASME/ANS RA-S-2008 Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, American Society of Mechanical Engineers, New York, NY, September 2013.

3.2.8.8 CNC-1535.00-00-0200, Catawba Nuclear Station PRA Peer Review F&O Resolutions, Rev. 2.

3.2.8.9 CNC-1535.00-00-0220, Catawba Nuclear Station Probabilistic Risk Assessment Resolution of Peer Review Facts and Observation, Revision 1.

3.2.8.10 APC 17-13,

Subject:

NRC Acceptance of Industry Guidance on Closure of PRA Peer Review Findings, dated May 8th, 2017 with attachment Appendix X.

3.2.8.11 Fire Probabilistic Risk Assessment (FPRA) Peer Review Process Guidelines, NEI 07-12, Revision 1, Nuclear Energy Institute, June 2010. [ML102230070]

3.2.8.12 ML14077A054, Letter from Catawba Nuclear Station to NRC, Flood Hazard Reevaluation Report, Response to NRC 10 CFR 50.54(f) Request for Additional Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(f) regarding Recommendations 2.1, 2.3 and 9.3 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident, Dated March 12, 2014.

U.S. Nuclear Regulatory Commission CNS-17-014 Enclosure, Page 65 3.2.8.13 CNS-194292-010, R 002 Draft, External Flooding & Dam Failures 3.2.8.14 CNC-1223.24-00-0072; RN Single Pond Return Header Design Basis; Rev. 1 3.2.8.15 CNS-1206.03-00-0001, Catawba Pipe Rupture Analysis Criteria Specification, Rev 2 3.2.8.16 CNS-1574.RN-00-0001, Design Basis Document for the Nuclear Service Water System (RN), Rev. 63 3.2.8.17 Duke Power Co., Catawba Nuclear Station Response to Generic Letter 88-20, Supplement 4 (IPEEE), June 1994 3.2.8.18 Duke Energy letter, Supplement Information Regarding Reevaluated Seismic Hazard Screening and Prioritization Results - Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-Chi Accident, dated Oct 20, 2016 3.2.8.19 NRC Letter, Request for Information Pursuant to Title 10 of the Code of Federal Regulations 0.54(f) Regarding Recommendations 2.1, 2.3, and 9.3, of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident, dated March 12, 2012 3.2.8.20 Duke Energy letter, Seismic Hazard and Screening Report (CEUS Sites), Response to NRC 10 CFR 50.54(f) Request for Additional Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(f) regarding Recommendations 2.1, 2.3 and 9.3 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident, dated March 31, 2014.

3.2.8.21 Duke Energy letter, Final Notification of Full Compliance with Order EA-12-049, Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond Design Basis External Events and with Order EA-1 2-051, Order to Modify Licenses With Regard To Reliable Spent Fuel Pool Instrumentation for McGuire Nuclear Station, dated December 7, 2015 3.2.8.22 Nuclear Energy Institute (NEI) 12-06, Diverse and Flexible Coping Strategies (FLEX)

Implementation Guide, dated December 2015 3.2.8.23 Duke Energy letter, Expedited Seismic Evaluation Process (ESEP) Report (CEUS Sites), Response to NRC Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(f) Regarding Recommendations 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident, dated December 17, 2014 3.2.8.24 CNC-1114.00-00-0040; Yard Drainage, Results of PMP; Rev. 30; October 21, 2015

U.S. Nuclear Regulatory Commission CNS-17-014 Enclosure, Page 66

4. Regulatory Evaluation 4.1 Applicable Regulatory Requirements / Criteria 4.1.1 Applicable Regulatory Documents The primary regulatory documents that are relevant when considering NSWS Single Pond Return Header Operation are:

10 CFR 50 Appendix A General Design Criteria: GDC 4 (Environmental and Dynamic Effects Design Bases) along with NUREG 0800 (Standard Review Plan)

10 CFR 50 Appendix A General Design Criteria: GDC 5 (Sharing of Structures, Systems and Components)

10 CFR 50 Appendix A General Design Criteria: GDC 44 (Cooling Water)

10 CFR 50 Appendix A General Design Criteria: GDC 45 (Inspection of Cooling Water System) and

NRC Reg Guide 1.174 (An Approach for Using Probabilistic Risk Assessments in Risk-Informed Decisions on Plant Specific Changes to the Licensing Basis).

Whereas Reg Guide 1.174 provides guidance on determining an acceptable CT based on core damage risk, it also directs the application of the traditional methods outlined by 10 CFR 50 Appendix A GDCs (such as the Single Failure Criterion). This is a defense in depth process that is applied with the PRA analysis and is used to evaluate areas of risk and measures mitigation such that the intent of the GDCs may continue to be met.

The applicable portions of each of the above documents is described in this section. The evaluation with respect to acceptability of NSWS Single Pond Return Header Operation follows the discussion of these documents.

4.1.1.1 10 CFR 50 App A, GDC 4 (Environmental and Dynamic Effects Design Bases)

GDC 4 addresses flooding concerns at Nuclear Power Plants. GDC 4 requires the evaluation of postulated pipe ruptures during normal operation and the evaluation that structures, systems, and components important to safety can withstand the effects of these breaks. This requirement is further defined by the Standard Review Plan (NUREG 0800) Section 3.6.1 and Branch Technical Position ASB 3-1. Specific criteria for postulated cracks in a moderate energy piping such as Nuclear Service Water are provided along with assumptions for initial conditions.

4.1.1.2 Standard Review Plan (NUREG 0800) Pipe Rupture The Standard Review Plan provides requirements that meet the requirements 10 CFR 50 Appendix A GDC 4 (Environmental and Dynamic Effects). Sections 3.6.1 (Plant Design For Protection Against Postulated Piping Failures in Fluid Systems Outside of Containment) and 3.6.2 (Determination of Pipe Rupture Locations and Dynamic Affects Associated with the Postulated Rupture of Piping) contain requirements including the associated Branch Technical Positions ASB 3-1 and MEB 3-1. The guidance in BTP ASB 3-1 includes:

"...postulated piping failures in fluid systems should not cause a loss of function of essential safety-related systems and that nuclear plants should be able to withstand postulated failures of any fluid system piping outside containment, taking into account the direct results

U.S. Nuclear Regulatory Commission CNS-17-014 Enclosure, Page 67 of such failure and the further failure of any single active component, with acceptable offsite results.

A single active component failure should be assumed in systems used to mitigate consequences of the postulated piping failure and to shut down the reactor...

All available systems, including those actuated by operator actions may be employed to mitigate the consequences of a postulated piping failure.

The critical crack size is taken to be 1/2 the pipe diameter in length and 1/2 the wall thickness in width."

BTP MEB 3-1 states in B.2.c.(1):

"Through-wall leakage cracks should be postulated in fluid systems..., except...(2) where the maximum stress range in these section of ...Class 2 or 3... piping is less than 0.4(1.2Sh +

SA)."

These requirements are reflected in the Catawba UFSAR and Catawba Pipe Specification. The exclusion for low stress areas is applied to one section of piping for NSWS Single Pond Return Header Operation; on the 1A Component Cooling (KC) Heat Exchanger return header nozzle.

This exclusion is possible by the addition of a reinforcing pad over the existing pad per future plant modification EC406529. The 1/4" reinforcing pad must be installed prior to entering NSWS Single Pond Return Header Operation.

Note that the installation of the reinforcing pad alleviates a pipe rupture in this area, which cannot be mitigated by re-alignment during Single Pond Return Header Operation (refer to Appendix G, and the scenario involving a rupture in the 1A NSWS Essential Header area with "B" train isolated). Isolating this area would remove the discharge flow path for three of four NSWS Essential Headers (1A, 1B and 2B).

When the NSWS is aligned to Lake Wylie, a rupture in the 1A NSWS Essential Header area can be promptly mitigated by isolation of the 1A NSWS Essential Header, and by aligning the three remaining NSWS Essential Headers to the SNSWP.

When NSWS is aligned to the SNSWP in the "normal dual header" alignment, a rupture in the 1A NSWS Essential Header area can be promptly mitigated by isolation of the 1A NSWS Essential Header, with the three remaining NSWS Essential Headers already aligned to the SNSWP.

4.1.1.3 ANS Standard on Pipe Rupture Methods ANSI/ANS-58.2-1980 (Design Basis for Protection of Light Water Nuclear Reactor Power Plants Against Effects of Postulated Pipe Rupture)

This document provided much of the detailed description applied in the Catawba UFSAR Section 3.6 and the Catawba Pipe Rupture Analysis Criteria Specification, Ref. 2.3.7.1. For the pipe rupture issues associated with Single Pond Return Header Operation, Single Pond Return Header Operation does not introduce any unanalyzed conditions, and since the requirements of

U.S. Nuclear Regulatory Commission CNS-17-014 Enclosure, Page 68 ANS-58.2 is very similar to the Standard Review Plan and the Catawba UFSAR, the details of this standard will not be listed here.

4.1.1.4 10 CFR 50 App A, GDC 5 (Sharing of structures, systems, and components)

GDC 5 states, "Structures, systems and components important to safety shall not be shared among nuclear power units unless it can be shown that such sharing will not significantly impair their ability to perform their safety functions, including, in the event of an accident in one unit, an orderly shutdown and cooldown of the remaining units."

In evaluating Single Pond Return Header Operation, the considerations of GDC 4 and GDC 44 are applied to establish measures that will enhance the reliability of the discharge portion of the NSWS. The evaluation of acceptability of NSWS Single Pond Return Header Operation is based on meeting the intent of these GDCs with (1) a minimal and acceptable PRA risk, (2) not impairing the ability of the SSCs to perform their safety functions and (3) with the NSWS remaining capable of providing adequate cooling water flow to support the accident on one unit concurrently with the shutdown and cooldown of the other. The application of this criteria is evaluated in Catawba calculation CNC-1223.24-00-0072 (RN Flow Analysis during Design Basis Events in Single Pond Return Header Operation), where various scenarios are modeled and flow and pressure to essential components is determined while aligned per NSWS Single Pond Return Header Operation and verified to meet acceptability criteria.

4.1.1.5 10 CFR 50 App A, GDC 44 (Cooling Water)

GDC 44 addresses the design of cooling water systems to be capable of performing their safety system function assuming a single failure.

"Suitable redundancy in components and features, and suitable interconnections, leak detection, and isolation capabilities shall be provided to assure that for onsite electric power system operation (assuming offsite power is not available) and for offsite electric power system operation (assuming onsite power is not available) the system safety function can be accomplished, assuming a single failure."

In addition, Appendix A provides a definition for single failure:

"Single failure. A single failure means an occurrence, which results in the loss of capability of a component to perform its intended safety functions. Multiple failures resulting from a single occurrence are considered to be a single failure. Fluid and electric systems are considered to be designed against an assumed single failure if neither (1) a single failure of any active component (assuming passive components function properly) nor (2) a single failure of a passive component (assuming active components function properly), results in a loss of the capability of the system to perform its safety functions."

In this evaluation, the single failure criterion is applied to the NSWS System with the limiting Design Basis Accident during NSWS Single Pond Return Header Operation. Failures are considered including active failures such as: (1) automatic valves that fail to reposition and (2) passive failures such as piping leaks and valve stem to disc failures. Whereas dual train pond return header operation clearly provides an alternate redundant header to use in the event of a failure of one header, in Single Pond Return Header Operation measures must be taken to eliminate failures or to respond to failures with procedural guidance such that they can be

U.S. Nuclear Regulatory Commission CNS-17-014 Enclosure, Page 69 mitigated. Single failure pertaining to the NSWS Single Pond Return Header Operation is evaluated with consideration of Industry Standards and the existing Catawba Licensing Bases.

4.1.1.6 ANSI/ANS Standards on Single Failure Two ANSI/ANS documents are applicable to Single Failure Documentation. These are ANSI N658-1976/ANS-51.7 (Single Failure Criteria for PWR Fluid Systems) and a later version, ANSI/ANS-58.9-1981 (Single Failure Criteria for Light Water Reactor Safety-Related Fluid Systems). These were written to clarify the requirements made in 10 CFR 50 Appendix A.

These documents contain the definition of single failure, active failure, and passive failure, and describe the rules for application of single failure criteria. In summary these state:

"Single Failure... refers to a random failure and its consequential effects assume in addition to an initiating event and its consequential effects for the purpose of safety-related fluid system design and analysis.

Active Failure ... An active failure is a malfunction, excluding passive failures, of a component that relies on mechanical movement to complete its intended function on demand. Examples of active failures include the failure of a powered valve or check valve to move to its correct position or the failure of a pump, fan, or diesel generator to start.

Passive Failure ... A passive failure is a failure of a component to maintain its structural integrity or the blockage of a process flow path. Blockage of a process flow path could occur, for example, due to the separation of a valve disc from its stem.

3. Rules for Application of the Single Failure Criteria 3.1 The unit shall be designed such that , for any Condition III or IV initiating event, the safety functions of ... emergency core and containment heat removal ... can be performed, assuming a single failure in addition to the initiating event.

3.4 During the short term, the single failure considered may be limited to an active failure.

3.5 During the long term assuming no prior failure during the short term, the limiting single failure considered can be either active or passive.

3.6 The design flow for a passive failure shall be defined by analysis of realistic passive failure mechanisms in the system, considering conditions of operation and possible failure or leakage modes, as appropriate.

... As an example...a review ... may result in the definition of a design leak rate for passive-failure evaluation based on maximum flow through a failed valve packing or pump mechanical seal.

4.3 If one train of a redundant safety-related fluid system or its safety-related supporting systems is temporarily rendered inoperable due to short-term maintenance as allowed by the unit technical specifications, a single failure need not be assumed in the other train."

Paragraph 3.6 basically defines credible passive failures of QA-1 piping to consist of realistic failures such as leaks through valve packing or mechanical seals. As a result, a cooling water

U.S. Nuclear Regulatory Commission CNS-17-014 Enclosure, Page 70 system may be able continue to provide adequate flow to components as long as it has the margin to withstand the continued leakage flow from a passive failure.

Paragraph 4.3 explains from a conventional regulatory compliance perspective, with both units in a short term TS LCO it is not a requirement to postulate a failure. This is based on the limited amount of time that the plant operates in the action statement and the overall low risk. The risk acceptability of the revised CT in this submittal is qualified by the PRA evaluation in accordance with Reg Guide 1.174. However this does not preclude applying considerations of Single Failure criterion and Pipe Rupture guidance in the evaluation of risk.

4.1.1.7 10 CFR 50 App A, GDC 45 (Inspection of Cooling Water System)

GDC 45 states, "The cooling water system shall be designed to permit appropriate periodic inspection of important components, such as heat exchangers and piping, to assure the integrity and capability of the system."

NSWS Single Pond Return Header Operation will aid in performing inspections, as required by GDC 45, along with associated repairs. The buried piping on this header consists of internally uncoated carbon steel piping that experiences intermittent flow. Whereas its internal wall degradation due to corrosion has been slower than that of piping that sees continuous service, it will potentially need repairs and internal coating in future years to preserve adequate wall thickness for the remaining plant life. Access to this piping for internal inspections requires more time than the existing TS CT allows.

4.1.1.8 NRC Reg Guide 1.174 (An Approach for Using Probabilistic Risk Assessments in Risk-Informed Decisions on Plant Specific Changes to the Licensing Basis)

Reg Guide 1.174 provides guidance in the use of Probabilistic Risk Assessment (PRA) in the support of "licensee requests for changes to a plant's license bases, as in requests for license amendments and technical specification changes." The Reg Guide conveys the "NRC's desire to base its decisions on the results of traditional engineering evaluations, supported by insights (derived from the use of PRA methods) about the risk significance of the proposed changes." It goes on to promote a Defense in Depth approach, which includes (along with the PRA) the use of 10 CFR 50, Appendix A General Design Criteria, national standards, and engineering principals, such as the single-failure criterion.

In compliance with this Reg Guide, 10 CFR 50 Appendix A GDCs that are relevant to the NSWS Single Pond Return header Operation are identified and reviewed. This review evaluates critical areas and devises measures to mitigate the risk. Subsequently, the PRA analysis based on the Single Pond Return Header Operation and the identified measures was performed and qualifies a TS CT that meets the needs of the station and presents a minimal and acceptable risk.

Reg Guide 1.174 provides a set of seven criteria that judge the adequacy of the use of the Defense in Depth approach for this submittal. These criteria are discussed in the text of the LAR submittal.

Note also that Reg Guide 1.174 provides guidance to the quality of the PRA. The PRA is documented in a Duke Calculation CNC-1535.00-00-0219.

U.S. Nuclear Regulatory Commission CNS-17-014 Enclosure, Page 71 4.2 Precedent The NRC has previously approved changes similar to the proposed changes in this License Amendment Request for Catawba Nuclear Station.

4.2.1 The NRC approved a comparable license amendment request to allow single supply header operation of the NSWS for Catawba via Amendments 243/237 on July 30, 2008 (Catawba Nuclear Station, Units 1 and 2, Issuance of Amendments Regarding Single Supply Header Operation of the Buried Nuclear Service Water System (TAC Nos.

MD6275 AND MD6276))

ADAMS Accession Number ML081980769.

The approved amendments reference all of the Duke Energy correspondence associated with this license amendment request. This correspondence included the following:

Initial submittal dated July 30, 2007 (ADAMS Accession Number ML072640193)

Responses to request for additional information:

Dated May 27, 2008 (ADAMS Accession Number ML081510801)

Dated June 23, 2008 (ADAMS Accession Number ML081770060)

This license amendment allowed a change of the LCO CT for TS 3.7.8 from 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to 30 days for the operation of the NSWS with one trains supply header from the NSWS pump house to the Auxiliary Building isolated and out of service for preplanned maintenance and inspections.

4.2.2 The Nuclear Regulatory Commission has issued Amendment No. 271 to Renewed Facility Operating License NPF-35 and Amendment No. 267 to Renewed Facility Operating License NPF-52 on August 9, 2013 for the Catawba Nuclear Station, Units 1 and 2, respectively. (Catawba Nuclear Station, Units 1 and 2, Issuance of Amendments Regarding Technical Specifications Changes for Nuclear Service Water System (TAC NOS. ME7659 and ME7660))

ADAMS Accession Number ML13169A139 The approved amendments reference all of the Duke Energy correspondence associated with this license amendment request. This correspondence included the following:

Initial submittal dated November 22, 2011 (ADAMS Accession Number ML11327A149)

Responses to request for additional information:

Dated July 9, 2012 (ADAMS Accession Number ML12194A218)

Dated November 12, 2012 (ADAMS Accession Number ML12319A075)

Dated May 15, 2013 (ADAMS Accession Number ML13140A012)

Dated January 28, 2013 (ADAMS Accession Number ML13032A006)

This license amendment allowed a change of the LCO CT for TS 3.7.8 from 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to 14 days for the operation of the NSWS with one trains Unit 2 Standby Nuclear Service Water Pond return header, located in the Auxiliary Building, to be isolated and out of service for preplanned maintenance and inspections.

U.S. Nuclear Regulatory Commission CNS-17-014 Enclosure, Page 72 4.3 No Significant Hazards Consideration Analysis Duke Energy requests a license amendment to revise the CNS Unit 1 and Unit 2 Technical Specifications (TSs). The proposed change will revise TS Section 3.7.8, "Nuclear Service Water System," to add a new condition to allow Single Pond Return Header Operation of the NSWS with a 30-Day Completion Time.

The proposed change would revise the TS to include a Single Pond Return Header Operation to the Nuclear Service Water System (NSWS) that involves isolating one train of the NSWS Pond Return piping at the Auxiliary Building wall and maintaining the discharge crossover lines open between trains in the Auxiliary Building and Emergency Diesel Generator (EDG) Buildings. This provides a common safety related discharge path through the single remaining in-service Pond Return line. This alignment, Single Pond Return Header Operation, allows a Pond Return Header to be removed from service while a flow path is maintained through both trains of NSWS supplied equipment to the Standby Nuclear Service Water Pond (SNSWP).

The NSWS Single Pond Return Header Operation is necessary to allow internal inspections and modifications of the NSWS Pond Return buried piping between the Auxiliary Building and the discharge to the SNSWP.

Duke Energy has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, Issuance of amendment, as discussed below:

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

Response: No.

The proposed single pond return header operation configuration for NSWS operation and the associated proposed TS and TS Bases changes have been evaluated to assess their impact on plant operation and to ensure that the design basis safety functions of safety related systems are not adversely impacted. During single pond return header operation, the operating NSWS header will be able to discharge all required NSWS flow from safety related components. PRA has demonstrated that due to the limited proposed time in the single pond return header configuration, the resultant plant risk remains acceptable.

The purpose of this amendment request is to ultimately facilitate inspections and modifications of the NSWS Pond Return buried piping between the Auxiliary Building and the Discharge to the SNSWP. Therefore, NRC approval of this request will ultimately help to enhance the long-term structural integrity of the NSWS and will help to ensure the system's reliability for many years.

In general, the NSWS serves as an accident mitigation system and cannot by itself initiate an accident or transient situation. The only exception is that the NSWS piping can serve as a source of floodwater to safety related equipment in the Auxiliary Building or in the diesel generator buildings in the event of a leak or a break in the system piping. The probability of such an event is not significantly increased as a result of this proposed request. Safety related NSWS piping is tested and inspected in accordance with all applicable in-service testing and in-service inspection requirements. Given the negligible influence of flooding events on the NSWS for the submittal configuration (i.e., no dominant contribution from floods), it is judged that the

U.S. Nuclear Regulatory Commission CNS-17-014 Enclosure, Page 73 analyses assessing the influence of these events provide an acceptable evaluation of the contribution of the flood risk for the requested CT of 30 days.

The proposed 30 day TS Required Action CT has been evaluated for risk significance and the results of this evaluation have been found acceptable. The probabilities of occurrence of accidents presented in the UFSAR will not increase as a result of implementation of this change. Because the PRA analysis supporting the proposed change yielded acceptable results, the NSWS will maintain its required availability in response to accident situations. Since NSWS availability is maintained, the response of the plant to accident situations will remain acceptable and the consequences of accidents presented in the UFSAR will not increase.

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

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

Response: No.

Implementation of this amendment will not create the possibility of a new or different kind of accident from any accident previously evaluated. The proposed request does not affect the basic operation of the NSWS or any of the systems that it supports. These include the Emergency Core Cooling System, the Containment Spray System, the Containment Valve Injection Water System, the Auxiliary Feedwater System, the Component Cooling Water System, the Control Room Area Ventilation System, the Control Room Area Chilled Water System, the Auxiliary Building Filtered Ventilation Exhaust System, or the Diesel Generators.

During proposed single pond return header operation, the NSWS will remain capable of fulfilling all of its design basis requirements.

No new accident causal mechanisms are created as a result of NRC approval of this amendment request. No changes are being made to the plant, which will introduce any new type of accident outside those assumed in the UFSAR.

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

3. Does the proposed change involve a significant reduction in a margin of safety?

Response: No.

Implementation of this amendment will not involve a significant reduction in any margin of safety. Margin of safety is related to the confidence in the ability of the fission product barriers to perform their design functions during and following an accident situation. These barriers include the fuel cladding, the reactor coolant system, and the containment system. The performance of these fission product barriers will not be impacted by implementation of this proposed TS amendment. During single pond return header operation, the NSWS and its supported systems will remain capable of performing their required functions. No safety margins will be impacted.

The PRA analysis conducted for this proposed amendment demonstrated that the impact on overall plant risk remains acceptable during single pond return header operation. Therefore, there is not a significant reduction in the margin of safety.

U.S. Nuclear Regulatory Commission CNS-17-014 Enclosure, Page 74 Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, Duke Energy concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92, and, accordingly, a finding of no significant hazards consideration is justified.

4.4 Conclusion 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 Commissions 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.

5. ENVIRONMENTAL CONSIDERATION A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure.

Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

6. REFERENCES 6.1 10 CFR50 General Design Criteria for Nuclear Power Appendix A, GDC 2, Natural Phenomena Appendix A, GDC 4, Environmental and Dynamic Effects Design Bases Appendix A, GDC 5, Sharing of Structures, Systems and Components Appendix A, GDC 44, Cooling Water Appendix A, GDC 45, Inspection of Cooling Water System 6.2 Standard Review Plan (NUREG 0800)

Section 3.6.1, Plant Design for Protection Against Postulated Piping Failures in Fluid Systems Outside Containment, including BTP ASB 3-1 Section 3.6.2, Determination of Rupture Locations and Dynamic Effects Associated with the Postulated Rupture of Piping, including BTP MEB 3-1 Section 9.2.1, Station Service Water System Section 9.2.5, Ultimate Heat Sink 6.1 Regulatory Guide 1.174, An Approach for Using Probabilistic Risk Assessments in Risk-Informed Decisions on Plant Specific Changes to the Licensing Basis 6.2 Regulatory Guide 1.27, Ultimate Heat Sink for Nuclear Power Plants 6.3 Regulatory Guide 1.53, Application of the Single Failure Criterion to Safety Systems 6.4 ANSI N658-1976 ANS-51.7, Single Failure Criteria for PWR Fluid Systems 6.5 ANSI/ANS-58.9-1981, Single Failure Criteria for Light Water Reactor Safety -Related Fluid Systems

U.S. Nuclear Regulatory Commission CNS-17-014 Enclosure, Page 75 6.6 ANSI/ANS-58.2-1980, Design Basis for Protection of Light Water Nuclear Power Plants Against Effects of Postulated Pipe Rupture 6.7 CNC-1223.24-00-0072, RN Single Pond Return Header Operation Design Basis 6.8 CNC-1150.01-00-0001, SNSWP Thermal Analysis During One Unit LOCA and One Unit Shutdown 6.9 CNC-1223.24-00-0027, A Flow Distribution Model of the RN System 6.10 CNS-1206.03-00-0001, Catawba Pipe Rupture Analysis Criteria Specification 6.11 CNC-1535.00-00-0219, Risk Determination for Proposed Catawba RN LCOs to Implement Single Pond Return Header Operation 6.12 CATAWBA UFSAR Rev. 18

U.S. Nuclear Regulatory Commission , Page l 1 CNS-17-014 Attachment 1 List of Regulatory Commitments

U.S. Nuclear Regulatory Commission , Page l 2 CNS-17-014 List of Regulatory Commitments The following table identifies the regulatory commitments contained in this document by Duke Energy Carolinas, LLC (Duke Energy). Any other statements in this submittal represent intended or planned actions and are provided as information purposes and are not to be considered regulatory commitments.

Type Scheduled Commitment One- Continuing Completion Time Compliance The support system for the NSWS Discharge piping associated with Train 1A in the Auxiliary Building will be maintained such that stress levels are below the threshold for considering a pipe leak under the Pipe Rupture program. This ensures that for all sections where pipe ruptures are postulated that leaks can be X Complete isolated with the NSWS continuing to operate with adequate equipment to support shutdown of both units. Catawba NSWS piping stress calculations RNG, RNH, and RNE have been revised to indicate the requirement to maintain this low stress level.

To reduce stress at the 1A Component Cooling (KC)

Heat Exchanger piping return nozzle location, a 1/4" thick reinforcing pad will be added to the existing Prior to entry reinforcing pad per plant modification EC406529. X into TS The 1/4" reinforcing pad must be installed prior to Condition D entering NSWS Single Pond Return Header Operation.

NSWS Flow Balance testing will take place prior to entering Single Pond Return Header Operation. This will ensure the NSWS is capable of providing Immediately adequate cooling water flow to support LOCA loads prior to each X

on one unit, concurrent with the shutdown loads of entry into TS the other unit - while assuming the most limiting Condition D single failure, which is loss of one EDG and its associated NSWS Pump.

U.S. Nuclear Regulatory Commission CNS-17-014 Attachment 2 Technical Specification Pages (Mark-up)

NSWS 3.7.8 3.7 PLANT SYSTEMS 3.7.8 Nuclear Service Water System (NSWS)

LCO 3.7.8 Two NSWS trains shall be OPERABLE.

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

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. ------------NOTE------------ A.1 -------------NOTES------------

Not applicable while in 1. Enter applicable Condition B, C, or D of Conditions and this LCO unless entry is Required Actions of directed by Note 2 of LCO 3.8.1, "AC Condition B, C, or D. Sources

Operating," for emergency diesel One NSWS train generator made inoperable. inoperable by NSWS.

2. Enter applicable Conditions and Required Actions of LCO 3.4.6, "RCS LoopsMODE 4," for residual heat removal loops made inoperable by NSWS.

Restore NSWS train to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months OPERABLE status.

(continued)

Catawba Units 1 and 2 3.7.8-1 Amendment Nos. 271/267

FOR INFORMATION ONLY NSWS 3.7.8 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. -----------NOTES----------- B.1 Restore NSWS supply 30 days

1. Entry into this header to OPERABLE Condition shall only status.

be allowed for pre-planned activities as described in the Bases of this Specification.

2. Immediately enter Condition A of this LCO if one or more NSWS components become inoperable while in this Condition and one NSWS train remains OPERABLE.

3. Immediately enter LCO 3.0.3 if one or more NSWS components become inoperable while in this Condition and no NSWS train remains OPERABLE.

One NSWS supply header inoperable due to NSWS being aligned for single supply header operation.

(continued)

Catawba Units 1 and 2 3.7.8-2 Amendment Nos. 271/267

NSWS 3.7.8 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. -----------NOTES----------- C.1 Restore NSWS train to 14 days

1. Entry into this OPERABLE status.

Condition shall only be allowed for Unit 1 and for pre-planned activities as described in the Bases of this Specification. Entry into this Condition shall not be allowed while Unit 2 is in MODE 1, 2, 3, or 4.

2. Immediately enter Condition A of this LCO if one or more Unit 1 required NSWS components become inoperable while in this Condition and one NSWS train remains OPERABLE.

3. Immediately enter LCO 3.0.3 if one or more Unit 1 required NSWS components become inoperable while in this Condition and no NSWS train remains OPERABLE.

One NSWS train inoperable due to NSWS being aligned for single Auxiliary Building discharge header operation.

(continued)

INSERT 1 here Catawba Units 1 and 2 3.7.8-3 Amendment Nos. 271/267

INSERT 1 for TS 3.7.8 NSWS 3.7.8 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D. -----------NOTES----------- D.1 Restore NSWS Pond 30 days

1. Entry into this return header to Condition shall only OPERABLE status.

be allowed for pre-planned activities as described in the Bases of this Specification

2. Immediately enter Condition A of this LCO if one or more NSWS components become inoperable while in this Condition and one NSWS train remains OPERABLE.

3. Immediately enter LCO 3.0.3 if one or more NSWS components become inoperable while in this Condition and no NSWS train remains OPERABLE.

One NSWS Pond return header inoperable due to NSWS being aligned for single Pond return header operation.

(continued)

NSWS 3.7.8 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME ED. Required Action and ED.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition A, B, AND C, or D not met.

ED.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.8.1 --------------------------------------NOTE----------------------------

Isolation of NSWS flow to individual components does not render the NSWS inoperable.

Verify each NSWS manual, power operated, and In accordance with automatic valve in the flow path servicing safety related the Surveillance equipment, that is not locked, sealed, or otherwise Frequency Control secured in position, is in the correct position. Program SR 3.7.8.2 ----------------------------------NOTE---------------------------------

Not required to be met for valves that are maintained in position to support NSWS single supply, single Auxiliary Building discharge header operation, or single Pond return header operation.

In accordance with Verify each NSWS automatic valve in the flow path that is the Surveillance not locked, sealed, or otherwise secured in position, Frequency Control actuates to the correct position on an actual or simulated Program actuation signal.

SR 3.7.8.3 Verify each NSWS pump starts automatically on an In accordance with actual or simulated actuation signal. the Surveillance Frequency Control Program Catawba Units 1 and 2 3.7.8-4 Amendment Nos. 271/267

U.S. Nuclear Regulatory Commission CNS-17-014 Attachment 3 Technical Specification Bases (Mark-up, For Information Only)

NSWS B 3.7.8 B 3.7 PLANT SYSTEMS B 3.7.8 Nuclear Service Water System (NSWS)

BASES BACKGROUND The NSWS, including Lake Wylie and the Standby Nuclear Service Water Pond (SNSWP), provides a heat sink for the removal of process and operating heat from safety related components during a Design Basis Accident (DBA) or transient. During normal operation, and a normal shutdown, the NSWS also provides this function for various safety related and nonsafety related components. The safety related function is covered by this LCO.

The NSWS consists of two independent loops (A and B) of essential equipment, each of which is shared between units. Each loop contains two NSWS pumps, each of which is supplied from a separate emergency diesel generator. Each set of two pumps supplies two trains (1A and 2A, or 1B and 2B) of essential equipment through common discharge piping.

While the pumps are unit designated, i.e., 1A, 1B, 2A, 2B, all pumps receive automatic start signals from a safety injection or blackout signal from either unit. Therefore, a pump designated to one unit will supply post accident cooling to equipment in that loop on both units, provided its associated emergency diesel generator is available. For example, the 1A NSWS pump, supplied by emergency diesel 1A, will supply post accident cooling to NSWS trains 1A and 2A.

One NSWS loop containing two OPERABLE NSWS pumps has sufficient capacity to supply post loss of coolant accident (LOCA) loads on one unit and shutdown and cooldown loads on the other unit. Thus, the OPERABILITY of two NSWS loops assures that no single failure will keep the system from performing the required safety function. Additionally, one NSWS loop containing one OPERABLE NSWS pump has sufficient capacity to maintain one unit indefinitely in MODE 5 (commencing 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months following a trip from RTP) while supplying the post LOCA loads of the other unit. Thus, after a unit has been placed in MODE 5, only one NSWS pump and its associated emergency diesel generator are required to be OPERABLE on each loop, in order for the system to be capable of performing its required safety function, including single failure considerations.

Additional information about the design and operation of the NSWS, along with a list of the components served, is presented in the UFSAR, Section 9.2.1 (Ref. 1). The principal safety related function of the NSWS is the removal of decay heat from the reactor via the CCW System.

Catawba Units 1 and 2 B 3.7.8-1 Revision No.

NSWS B 3.7.8 BASES APPLICABLE The design basis of the NSWS is for one NSWS train, in conjunction SAFETY ANALYSES with the CCW System and a containment spray system, to remove core decay heat following a design basis LOCA as discussed in the UFSAR, Section 6.2 (Ref. 2). This prevents the containment sump fluid from increasing in temperature during the recirculation phase following a LOCA and provides for a gradual reduction in the temperature of this fluid as it is supplied to the Reactor Coolant System by the ECCS pumps. The NSWS is designed to perform its function with a single failure of any active component, assuming the loss of offsite power.

The NSWS, in conjunction with the CCW System, also cools the unit from residual heat removal (RHR), as discussed in the UFSAR, Section 5.4 (Ref. 3), from RHR entry conditions to MODE 5 during normal and post accident operations. The time required for this evolution is a function of the number of CCW and RHR System trains that are operating. Thirty six hours after a trip from RTP, one NSWS train is sufficient to remove decay heat during subsequent operations in MODES 5 and 6. This assumes a maximum NSWS temperature, a simultaneous design basis event on the other unit, and the loss of offsite power.

The NSWS satisfies Criterion 3 of 10 CFR 50.36 (Ref. 4).

LCO Two NSWS trains are required to be OPERABLE to provide the required redundancy to ensure that the system functions to remove post accident heat loads, assuming that the worst case single active failure occurs coincident with the loss of offsite power.

While the NSWS is operating in the normal dual supply and discharge header alignment, an NSWS train is considered OPERABLE during MODES 1, 2, 3, and 4 when:

a. 1. Both NSWS pumps on the NSWS loop are OPERABLE; or

2. One unit's NSWS pump is OPERABLE and one unit's flowpath to the non essential header, AFW pumps, and Containment Spray heat exchangers are isolated (or equivalent flow restrictions); and

b. The associated piping, valves, and instrumentation and controls required to perform the safety related function are OPERABLE.

Catawba Units 1 and 2 B 3.7.8-2 Revision No.

NSWS B 3.7.8 BASES LCO (continued)

The NSWS system is shared between the two units. The shared portions of the system must be OPERABLE for each unit when that unit is in the MODE of Applicability. Additionally, both normal and emergency power for shared components must also be OPERABLE. If a shared NSWS component becomes inoperable, or normal or emergency power to shared components becomes inoperable, then the Required Actions of this LCO must be entered independently for each unit that is in the MODE of applicability of the LCO, except as noted in a.2 above for operation in the normal dual supply header alignment. In this case, sufficient flow is available, however, this configuration results in inoperabilities within other required systems on one unit and the associated Required Actions must be entered. Use of a NSWS pump and associated diesel generator on a shutdown unit to support continued operation (> 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ) of a unit with an inoperable NSWS pump is prohibited. A shutdown unit supplying its associated emergency power source (1EMXG/2EMXH) cannot be credited for OPERABILITY of components supporting the operating unit.

APPLICABILITY In MODES 1, 2, 3, and 4, the NSWS is a normally operating system that is required to support the OPERABILITY of the equipment serviced by the NSWS and required to be OPERABLE in these MODES.

In MODES 5 and 6, the requirements of the NSWS are determined by the systems it supports.

ACTIONS A.1 Condition A is modified by a Note indicating that this Condition is not applicable while in Condition B, C, or D of this LCO unless entry is directed by Note 2 of Condition B, C, or D.

If one NSWS train is inoperable, action must be taken to restore OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . In this Condition, the remaining OPERABLE NSWS train is adequate to perform the heat removal function. However, the overall reliability is reduced because a single failure in the OPERABLE NSWS train could result in loss of NSWS function. Due to the shared nature of the NSWS, both units are required to enter a 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Action when a NSWS Train becomes inoperable on either unit. Required Action A.1 is modified by two Notes. The first Note indicates that the applicable Conditions and Required Actions of LCO 3.8.1, "AC SourcesOperating," should be entered if an inoperable Catawba Units 1 and 2 B 3.7.8-3 Revision No.

NSWS B 3.7.8 BASES ACTIONS (continued)

NSWS train results in an inoperable emergency diesel generator. The second Note indicates that the applicable Conditions and Required Actions of LCO 3.4.6, "RCS LoopsMODE 4," should be entered if an inoperable NSWS train results in an inoperable decay heat removal train (RHR). An example of when these Notes should be applied is with both units' loop 'A' NSWS pumps inoperable, both units' 'A' emergency diesel generators and both units' 'A' RHR systems should be declared inoperable and appropriate Actions entered. This is an exception to LCO 3.0.6 and ensures the proper actions are taken for these components. The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Completion Time is based on the redundant capabilities afforded by the OPERABLE train, and the low probability of a DBA occurring during this time period.

B.1 While the NSWS is operating in the single supply header alignment, one of the supply headers is removed from service in support of planned maintenance or modification activities associated with the supply header that is taken out of service. In this configuration, each NSWS train is considered OPERABLE with the required NSWS flow to safety related equipment being fed through the remaining OPERABLE NSWS supply header. While the NSWS is operating in the single supply header alignment, an NSWS train is considered OPERABLE during MODES 1, 2, 3, and 4 when:

a. The associated train related NSWS pumps are OPERABLE; and

b. The associated piping (except for the supply header that is taken out of service), valves, and instrumentation and controls required to perform the safety related function are OPERABLE.

If one NSWS supply header is inoperable due to the NSWS being aligned for single supply header operation, the NSWS supply header must be restored to OPERABLE status within 30 days. Dual supply header operation is the normal alignment of the NSWS. The Completion Time of 30 days is supported by probabilistic risk analysis. While in Condition B, the single supply header is adequate to perform the heat removal function for all required safety related equipment for both safety trains. Due to the shared nature of the NSWS, both units are required to enter this Condition when the NSWS is aligned for single supply header operation.

In order to prevent the potential for NSWS pump runout, the single NSWS pump flow balance alignment is prohibited while the NSWS is aligned for single supply header operation.

Catawba Units 1 and 2 B 3.7.8-4 Revision No.

NSWS B 3.7.8 BASES ACTIONS (continued)

Condition B is modified by three Notes. Note 1 states that entry into this Condition shall only be allowed for pre-planned activities as described in the Bases of this Specification. Condition B is only allowed to be entered in support of planned maintenance or modification activities associated with the supply header that is taken out of service. An example of a situation for which entry into this Condition is allowed is refurbishment or inspection of a supply header. Entry into this Condition is not allowed in response to unplanned events or for other events involving the NSWS.

Examples of situations for which entry into this Condition is prohibited are emergent repair of discovered piping leaks and other component failures.

For unplanned events or other events involving the NSWS, Condition A must be entered. Note 2 requires immediate entry into Condition A of this LCO if one or more NSWS components become inoperable while in this Condition and one NSWS train remains OPERABLE. With one remaining OPERABLE NSWS train, the NSWS can still perform its safety related function. However, with one inoperable NSWS train, the NSWS cannot be assured of performing its safety related function in the event of a single failure of another NSWS component. The most limiting single failure is the failure of an NSWS pit to automatically transfer from Lake Wylie to the SNSWP during a seismic event. While the loss of any NSWS component subject to the requirements of this LCO can result in the entry into Condition A, the most common example is the inoperability of an NSWS pump. This occurs during periodic testing of the emergency diesel generators. Inoperability of an emergency diesel generator renders its associated NSWS pump inoperable. Note 3 requires immediate entry into LCO 3.0.3 if one or more NSWS components become inoperable while in this Condition and no NSWS train remains OPERABLE. In this case, the NSWS cannot perform its safety related function.

C.1 While the NSWS is operating in the single Auxiliary Building discharge header alignment, one of the Unit 2 Auxiliary Building discharge headers is removed from service in support of planned maintenance or modification activities associated with the Auxiliary Building discharge header that is taken out of service. In this configuration, the corresponding (train related) Unit 1 NSWS train is inoperable and the required NSWS flow to safety related equipment is discharged through the remaining OPERABLE NSWS Auxiliary Building discharge header.

When in the single Auxiliary Building discharge header alignment with the NSWS Train A discharge header inoperable, the NSWS piping between Catawba Units 1 and 2 B 3.7.8-5 Revision No.

NSWS B 3.7.8 BASES ACTIONS (continued) valves 1RNP19 and 1RN63A is isolated. Likewise, when in the single Auxiliary Building discharge header alignment with the NSWS Train B discharge header inoperable, the NSWS piping between valves 1RNP20 and 1RN58B is isolated.

Operation of the NSWS in the single Auxiliary Building discharge header alignment while in either the single supply header alignment or the single Pond return header alignment at the same time is prohibited.

If one NSWS train is inoperable due to the NSWS being aligned for single Auxiliary Building discharge header operation, the NSWS train must be restored to OPERABLE status within 14 days. Dual Auxiliary Building discharge header operation is the normal alignment of the NSWS. The Completion Time of 14 days is supported by probabilistic risk analysis.

While in Condition C, the single Auxiliary Building discharge header is adequate to perform the heat removal function for all required safety related equipment for its respective safety train. Due to the design of the NSWS, only the operating unit is required to enter this Condition when the NSWS is aligned for single Auxiliary Building discharge header operation.

Pre-planned activities requiring entry into this Condition are only performed with Unit 2 in an outage (MODE 5, 6, or defueled).

Condition C is modified by three Notes. Note 1 states that entry into this Condition shall only be allowed for Unit 1 and for pre-planned activities as described in the Bases of this Specification. Condition C is only allowed to be entered in support of planned maintenance or modification activities associated with the Auxiliary Building discharge header that is taken out of service. An example of a situation for which entry into this Condition is allowed is refurbishment or inspection of an Auxiliary Building discharge header. Entry into this Condition is not allowed in response to unplanned events or for other events involving the NSWS. Examples of situations for which entry into this Condition is prohibited are emergent repair of discovered piping leaks and other component failures. For unplanned events or other events involving the NSWS, Condition A must be entered.

In addition, Note 1 states that entry into this Condition shall not be allowed while Unit 2 is in MODE 1, 2, 3, or 4. Entry into this Condition is only allowed while the LCO is not applicable to Unit 2. Note 2 requires immediate entry into Condition A of this LCO if one or more Unit 1 required NSWS components become inoperable while in this Condition and one NSWS train remains OPERABLE. With one remaining OPERABLE NSWS train, the NSWS can still perform its safety related function. However, with one inoperable NSWS train, the NSWS cannot be assured of performing its safety related function in the event of a ACTIONS (continued)

Catawba Units 1 and 2 B 3.7.8-6 Revision No.

NSWS B 3.7.8 BASES single failure of another NSWS component. While the loss of any NSWS component subject to the requirements of this LCO can result in the entry into Condition A, the most common example is the inoperability of an NSWS pump. This occurs during periodic testing of the emergency diesel generators. Inoperability of an emergency diesel generator renders its associated NSWS pump inoperable. Note 3 requires immediate entry into LCO 3.0.3 if one or more Unit 1 required NSWS components become inoperable while in this Condition and no NSWS train remains OPERABLE. In this case, the NSWS cannot perform its safety related function.

E.1 and E.2 If the NSWS train cannot be restored to OPERABLE status within the associated Completion Time, or if the NSWS single supply header, single INSERT 1 Aux Building discharge header, or single Pond return header cannot be restored to OPERABLE status within the associated Completion Time, the unit must be placed in a MODE in which the LCO does not apply. To achieve this status, the unit must be placed in at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.

SURVEILLANCE SR 3.7.8.1 REQUIREMENTS This SR is modified by a Note indicating that the isolation of the NSWS components or systems may render those components inoperable, but does not affect the OPERABILITY of the NSWS.

Verifying the correct alignment for manual, power operated, and automatic valves in the NSWS flow path provides assurance that the proper flow paths exist for NSWS operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since they are verified to be in the correct position prior to being locked, sealed, or secured. This SR does not require any testing or valve manipulation; rather, it involves verification that those valves capable of being mispositioned are in the correct position. This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves.

The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.

Catawba Units 1 and 2 B 3.7.8-7 Revision No.

NSWS B 3.7.8 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.7.8.2 This SR verifies proper automatic operation of the NSWS valves on an actual or simulated actuation signal. The signals that cause the actuation are from Safety Injection and Phase 'B' isolation. The NSWS is a normally operating system that cannot be fully actuated as part of normal testing. This Surveillance is not required for valves that are locked, sealed, or otherwise secured in the required position under administrative controls. The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.

This SR is modified by a Note that states that the SR is not required to be met for valves that are maintained in position to support NSWS single supply, single Auxiliary Building discharge header operation, or single Pond return header operation. When the NSWS is placed in this alignment, certain automatic valves in the system are maintained in position and will not automatically reposition in response to an actuation signal while the NSWS is in this alignment.

SR 3.7.8.3 This SR verifies proper automatic operation of the NSWS pumps on an actual or simulated actuation signal. The signals that cause the actuation are from Safety Injection and Loss of Offsite Power. The NSWS is a normally operating system that cannot be fully actuated as part of normal testing during normal operation. The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.

REFERENCES 1. UFSAR, Section 9.2.

2. UFSAR, Section 6.2.

3. UFSAR, Section 5.4.

4. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).

Catawba Units 1 and 2 B 3.7.8-8 Revision No.

INSERT 1 FOR TS 3.7.8 BASES D.1 While the NSWS is operating in the single Pond return header alignment, one of the shared discharge headers from the Aux Bldg to the SNSWP is removed from service in support of planned maintenance or modification activities associated with the Pond return header that is taken out of service. In this configuration, each NSWS train is considered OPERABLE with the required NSWS flow path from safety related equipment through the remaining OPERABLE NSWS Pond return header.

While the NSWS is operating in the single Pond return header alignment, an NSWS train is considered OPERABLE during MODES 1, 2, 3, and 4 when:

a. The associated train related NSWS pumps are OPERABLE; and

b. The associated piping (except for the Pond return header that is taken out of service), valves, and instrumentation and controls required to perform the safety related function are OPERABLE.

When in the single Pond return header alignment with the NSWS Train A Pond return header inoperable, the NSWS piping downstream of valves 1RN63A and 1(2)RN846A is isolated. Valve 1RNP20 is locked open, 1RN58B is open with power removed, and 1(2)RN848B are open with power removed to protect against closing that would isolate the discharge flow from both trains.

Similarly, when in the single Pond return header alignment with the NSWS Train B Pond return header inoperable, the NSWS piping downstream of valves 1RN58B and 1(2)RN849B is isolated. In this case valve 1RNP19 is locked open, 1RN63A is open with power removed, and 1(2)RN846A are open with power removed to protect against closing that would isolate the discharge flow from both trains.

When in the single Pond return header alignment the RN System is aligned to the SNSWP to preclude a single active failure that could result in the complete loss of one RN pit (two RN pumps). Aux Bldg discharge crossover piping valves 1RN53B and 1RN54A are open with power removed to allow both Aux Bldg trains to discharge through one header. Similarly, Unit 1 and Unit 2 D/G crossover valves 1(2)RNP08 and 1(2)RNP09 are locked open to allow both trains of D/Gs to discharge through one header one each unit. Finally, both Unit 1 and Unit 2 RN non-essential headers are isolated.

Operation of the NSWS in the single Pond return header alignment while in either the single supply header alignment or the single Auxiliary Building discharge header alignment at the same time is prohibited.

If one NSWS Pond return header is inoperable due to the NSWS being aligned for single Pond return header operation, the NSWS Pond return header must be restored to OPERABLE status within 30 days. The Completion Time of 30 days is supported by probabilistic risk analysis.

While in Condition D, the single Pond return header alignment is adequate to perform the heat removal function for all required safety related equipment for both safety trains of both units. Due to the shared nature of the NSWS, both units are required to enter this Condition when the NSWS is aligned for single Pond return header operation. In order to ensure adequate flow to essential components, the single NSWS pump flow balance alignment is prohibited while the NSWS is aligned for single Pond return header operation.

Condition D is modified by three Notes. Note 1 states that entry into this Condition shall only be allowed for pre-planned activities as described in the Bases of this Specification. Condition D is only allowed to be entered in support of planned maintenance or modification activities associated with the Pond return header that is taken out of service. An example of a situation for which entry into this Condition is allowed is refurbishment or inspection of a Pond return header. Entry into this Condition is not allowed in response to unplanned events or for other events involving the NSWS. Examples of situations for which entry into this Condition is prohibited are emergent repair of discovered piping leaks and other component failures. For unplanned events or other events involving the NSWS, Condition A must be entered. Note 2 requires immediate entry into Condition A of this LCO if one or more NSWS components become inoperable while in this Condition and one NSWS train remains OPERABLE. With one remaining OPERABLE NSWS train, the NSWS can still perform its safety related function. However, with one inoperable NSWS train, the NSWS cannot be assured of performing its safety related function in the event of a single failure of another NSWS component. While the loss of any NSWS component subject to the requirements of this LCO can result in the entry into Condition A, the most common example is the inoperability of an NSWS pump. This occurs during periodic testing of the emergency diesel generators.

Inoperability of an emergency diesel generator renders its associated NSWS pump inoperable. Note 3 requires immediate entry into LCO 3.0.3 if one or more NSWS components become inoperable while in this Condition and no NSWS train remains OPERABLE. In this case, the NSWS cannot perform its safety related function.

U.S. Nuclear Regulatory Commission CNS-17-014 Attachment 4 PRA Peer Review Findings and Resolutions

U.S. Nuclear Regulatory Commission PRA Peer Review Findings and Resolutions CNS-17-014 4.1 Internal Events, CDF There were 8 internal events PRA Findings which were considered to be open after the December 2015 Peer Review (Table 1 of Reference 3) and the 2017 independent F&O closure technical review. This included F&Os for which the Supporting Requirements (SRs) were met at Capability Category II, as well as SRs that were rated at Capability Category I. Reference 1 provides a resolution to each of these F&O Findings.

These are discussed along with their impact to the 30-day CT T.S. LAR in this Appendix.

F&Os Regarding PRA Supporting Requirements F&O ID: Associated SRs: Peer Review CC Assessment:

20-9 HR-I1, HR-D6 Met at CC I-III SR HR-I1 Capability Category I/II/III Requirements:

DOCUMENT the human reliability analysis in a manner that facilitates PRA applications, upgrades, and peer review.

F&O Issue and Proposed Resolution:

Table 2 of the HRA Notebook documents the mean and error factor for all the pre-initiator HFEs.

Several errors have been identified in this table that need to be corrected. For example the mean and error factor for 1(2)LOPER_KCNCPAF are listed as 8.00E-02 and 5, respectively. The correct mean and error factor are 8.00E-04 and 10, respectively. It was found that the correct values are used in the HRA calculator and in the PRA model files.

(This F&O originated from SR HR-I1)

Basis for Significance:

These are documentation errors that should be resolved.

Possible Resolution:

Review all of Table 2 in the HRA Notebook and ensure that all pre-initiator HFEs have the correct mean and error factor.

Disposition of the Peer Review Finding:

Resolution of F&O:

CNC-1535.00-00-0205 (Reference 4) contains updated tables reflecting the mean and error factors for the human failure events in the Catawba PRA model (see Table 7-1: Miscalibration HEP Summary and Table 7-4: Summary of Misalignment HEPs). As the correct mean and error factor values are now documented, this finding has been fully addressed and is now considered closed.

Evaluation of F&O impact on proposed application:

Updated mean and error factors for HRA events have been incorporated in the model used to support the Catawba NSWS 30-day CT T.S. LAR.

U.S. Nuclear Regulatory Commission PRA Peer Review Findings and Resolutions CNS-17-014 4.1 Internal Events, CDF F&Os Regarding PRA Supporting Requirements F&O ID: Associated SRs: Peer Review CC Assessment:

20-14 QU-E4, QU-F4 Met at CC I-III SR QU-E4 Capability Category I/II/III Requirements:

For each source of model uncertainty and related assumption identified in QU-E1 and QU-E2, respectively, IDENTIFY how the PRA model is affected (e.g., introduction of a new basic event, changes to basic event probabilities, change in success criterion, introduction of a new initiating event) [Note (1)].

F&O Issue and Proposed Resolution:

Modeling assumptions that are documented in the PRA notebooks should be accounted for in the uncertainty process. Assumptions made throughout the PRA should identify if they contribute to uncertainty and if they do the impact on the PRA model should be identified and the uncertainty characterized.

As part of the uncertainty characterization the level of significance on the baseline PRA should be assessed (e.g., important, not important, conservative, optimistic). This will facilitate future uncertainty analyses that are required for PRA applications.

This should be documented in accordance with QU-F4.

(This F&O originated from SR QU-E4).

Basis for Significance:

It is important that the uncertainty assessment has been performed that covers not just on the generic areas of uncertainty but also the plant-specific sources of modeling uncertainty.

Possible Resolution:

Review each of the technical elements (e.g., success criteria, accident sequences, system notebooks, HRA, etc.) to identify any plant-specific potential sources of uncertainty that are not already covered by the generic sources of uncertainty.

Disposition of the Peer Review Finding:

Resolution of F&O:

The specific assumptions from each PRA notebook have been compiled into generic categories for assessment based on EPRI report 1016737. These assumptions are documented in Table 16 of Section 6.3.4 of the quantification notebook (Reference 5), which closes this finding.

Evaluation of F&O impact on proposed application:

Assumptions impacting the Catawba NSWS T.S. CT LAR are documented in this calculation.

U.S. Nuclear Regulatory Commission PRA Peer Review Findings and Resolutions CNS-17-014 4.1 Internal Events, CDF F&Os Regarding PRA Supporting Requirements F&O ID: Associated SRs: Peer Review CC Assessment:

22-7 QU-C2 Met at CC I-III Capability Category I/II/III Requirements:

ASSESS the degree of dependency between the HFEs in the cutset or sequence in accordance with HR-D5 and HR-G7.

F&O Issue and Proposed Resolution:

Review of QU notebook (RSC 14-31/CNC-1535.00-00-0180) section 3.2 and recovery rule file (U1-CNSRecovery_final_CDF2a.txt) shows that dependencies between HFEs were captured using recovery rules. However, the recovery rules were not developed to capture the dependencies between HFEs correctly. An HEP combination with low probability was applied first in recovery rules rather than larger number of combination as below (U1-CNSRecovery_final_CDF2a.txt);

Line 3008

- Recovery** HFE_Combo_310 1.821e-005 1OPER_RCHTLPS 1OPER_FNBI 0OPER_RCDMWSO Line 5736

- Recovery** HFE_Combo_3804 3.323e-005 1OPER_KCMVAIS 1OPER-SITERM 1OPER_RCHTLPS 1OPER_FNBI 0OPER_RCDMWSO (This F&O originated from SR QU-C2).

Basis for Significance:

The impact of the CNS PRA quantification analysis is indeterminate because the second HEP combination which has higher joint probability was not able to be applied in the cutsets with command **MAX RECOVERIES** 1.

Possible Resolution:

The recovery rule files need to be revised to assess properly the degree of dependency between the HFEs in the cutset or sequence in conjunction with review of a joint human error probability of dependent HEP combination. Alternately, a basis needs to documented for acceptability of the existing recovery rules.

Disposition of the Peer Review Finding:

Resolution of F&O:

A new dependency analysis has been performed and the new recovery rules have been implemented in the model used for this application. The recovery approach is based on the current state of knowledge in using the EPRI HRA calculator application to define dependencies. The current HRA calculator generates, in some cases, higher probabilities for larger event combinations than possible. The ranking by smallest to largest probability avoids the use of illogical dependency results. This closes the finding.

Evaluation of F&O impact on proposed application:

The HRA dependencies representing the model were used to support the NSW 30-day CT.

U.S. Nuclear Regulatory Commission PRA Peer Review Findings and Resolutions CNS-17-014 4.1 Internal Events, CDF F&Os Regarding PRA Supporting Requirements F&O ID: Associated SRs: Peer Review CC Assessment:

23-8 SY-A19 Met at CC I-III Capability Category I/II/III Requirements:

In the systems model, INCLUDE out-of-service unavailability for components in the system model, unless screened, in a manner consistent with the actual practices and history of the plant for removing equipment from service.

(a) INCLUDE (1) unavailability caused by testing when a component or system train is reconfigured from its required accident mitigating position such that the component cannot function as required (2) maintenance events at the train level when procedures require isolating the entire train for maintenance (3) maintenance events at a sub-train level (i.e., between tagout boundaries, such as a functional equipment group) when directed by procedures (b) Examples of out-of-service unavailability to be modeled are as follows:

(1) train outages during a work window for preventive/corrective maintenance (2) a functional equipment group (FEG) removed from service for preventive/corrective maintenance (3) a relief valve taken out of service F&O Issue and Proposed Resolution:

The documentation of Maintenance events is inconsistent in the system notebooks. Some notebooks (e.g., Safety Injection, Chemical and Volume Control System) specify the basic event(s) used for unavailability due to maintenance or testing, whereas other notebooks (e.g., AFW, EDG) do not.

(This F&O originated from SR SY-A19)

Basis for Significance:

Without the basic events specified in the notebooks, it is difficult to ensure all of the events are properly modeled.

Possible Resolution:

Include the basic event designation(s) in the system notebook.

Disposition of the Peer Review Finding:

Resolution of F&O:

The maintenance events used in the model have been completely reviewed, and the system notebooks were updated to reflect the model of record and to be consistent with one another in listing the maintenance events. This closes the finding.

Evaluation of F&O impact on proposed application:

This is a documentation issue. During the development of the 30-day CT analysis, it was confirmed the PRA contains the test and maintenance basic events for the D/G and AFW systems (as well as the NSWS). This is a documentation issue which has no impact on this application.

U.S. Nuclear Regulatory Commission PRA Peer Review Findings and Resolutions CNS-17-014 4.1 Internal Events, CDF F&Os Regarding PRA Supporting Requirements F&O ID: Associated SRs: Peer Review CC Assessment:

24-1 AS-A2 Met at CC I-III Capability Category I/II/III Requirements:

For each modeled initiating event, IDENTIFY the key safety functions that are necessary to reach a safe, stable state and prevent core damage. [See Note (1).]

F&O Issue and Proposed Resolution:

According to Accident Sequence Analysis notebook Section 10.2, 100% AFW is defined as all three AFW pumps feeding to all SGs. However, gate #1B1 models this as 2 out of 4 SGs required. This is inconsistent with the documentation, and needs to be corrected.

Finally, it is not clear from Accident Sequence Analysis notebook Section 10.2 what are the Pressure Relief requirements for the 4 cases identified with success and failure of AFW and CRI.

(This F&O originated from SR AS-A2)

Basis for Significance:

This is a Finding F&O since it identifies a potential model error. In addition, it provides a recommendation to clarify ATWS documentation.

Possible Resolution:

Correct the inconsistency, and clarify the ATWS documentation.

Disposition of the Peer Review Finding:

Resolution of F&O:

A review of the ATWS event tree in Reference 9 and the CA system design indicated that a more fundamental change was needed to support the F&O resolution. As a result of these findings and to resolve the F&O the following changes are needed:

1. The generic WCAP assessment (Reference 20) assumes that the CA turbine driven pump (TDP) has a nominal flow rate that is twice that of a single CA motor driven pump (MDP). This is the basis for the 100% and 200% flow rates noted in the report. The CNS CA system has two MDPs and one TDP but the TDP flow rate is only slightly higher than a single MDP. As such, CNS can only provide 100% and 150% flow. In addition, the current CNS procedural guidance indicates to the operators that a minimum flow of 1,000 gpm is acceptable which is equivalent to combining any two of the three CA pumps. The combination of these factors eliminates the 200% flow cases provided in the generic WCAP and the CA flow assessment is limited to the so called 100% case.

2. Scenario IEKT related to turbine trip is missing from the CNS CAFTA model and needs to be added taking credit for closure of the MSIVs if the main turbine does not trip. The signal to close the MSIVs is generated automatically based on a mismatch of steam and feedwater flow rates.

U.S. Nuclear Regulatory Commission PRA Peer Review Findings and Resolutions CNS-17-014 4.1 Internal Events, CDF

3. The PORV requirements must be related to the associated probability of a less negative UET leading to overpressure. This is also associated with items 1, 4, 5 and 6.

4. The current model assumes that for any scenario that the UET is 1.0 for the defined pressure scenarios. This is conservative and not reflective of the influence that CA, pressurizer PORVs and manual rod insertion have on RCS pressurization. This is discussed in the accident sequence notebook but conservatively neglected in the ATWS event tree.

5. The effect of manual rod insertion (MRI) is not included in the ATWS event tree but has essentially no impact on the overpressure assessment due to the lack of UET consideration.

The definition of ATWS is that 10 of 50 groups do not insert. That leaves up to 40 groups that can be inserted manually as long as the failure mode is not due to binding of the rods. The insertion of a group to 72 steps is sufficient to lower the pressure a measurable amount. The model should include MRI in combination with UET and pressurizer PORV capacity in assessing the potential for overpressure.

6. The overpressure event PA includes the pressurizer PORV model but the model is overly conservative since it requires operation of the level instrumentation which is not required. It also credits operator intervention which cannot occur in a timely manner and is not required. The PORV model utilizes the same number of PORVs regardless of the success of MRI or CA flow.

The conservatism introduced is small but should be refined due to the changes in CA flow options.

The inclusion of these changes to the ATWS model closes the finding.

Evaluation of F&O impact on proposed application:

The resulting changes were made to the model used to support the NSW 30-day CT LAR.

U.S. Nuclear Regulatory Commission PRA Peer Review Findings and Resolutions CNS-17-014 4.1 Internal Events, CDF F&Os Regarding PRA Supporting Requirements F&O ID: Associated SRs: Peer Review CC Assessment:

24-14 SY-B9 Met at CC I-III Capability Category I/II/III Requirements:

When modeling a system, INCLUDE appropriate interfaces with the support systems required for successful operation of the system for a required mission time (see also SY-A6).

Examples of support systems include (a) actuation logic (b) support systems required for control of components (c) component motive power (d) cooling of components (e) any other identified support function (e.g., heat tracing) necessary to meet the success criteria and associated systems F&O Issue and Proposed Resolution:

According to an assumption documented in the System notebooks, HVAC is not required for safety injection and component cooling water pumps since failure data are assumed to include failures due to high temperature caused by loss of HVAC. The peer review team is not convinced that is the case. In the absence of room heatup calculations, loss of HVAC should be modeled to fail PRA equipment in the area. Alternative, the screening can be justified by room heatup calculations, if available. Note that SR SY-B12 does not allow screening based on the availability of proceduralized recovery actions.

(This F&O originated from SR SY-B9).

Basis for Significance:

This is finding since it has the potential of resulting in a model change.

Possible Resolution:

Provide a documented basis for screening out HVAC systems that provide cooling to rooms with PRA components. The basis should include an engineering analysis (see SR SY-B6), simplified heatup calculation, or other references that support the modeling or screening of ventilation support systems.

Disposition of the Peer Review Finding:

Resolution of F&O:

The auxiliary building HVAC system (VA) is not required for design basis events. Since this is the only system providing cooling to identified loads and it is not required within in the conservative design basis, it is not required for the PRA model. This basis closes the finding.

Evaluation of F&O impact on proposed application:

Since the auxiliary building HVAC system (VA) is not required for design basis events, there is no impact on the NSW 30-day T.S. CT.

U.S. Nuclear Regulatory Commission PRA Peer Review Findings and Resolutions CNS-17-014 4.1 Internal Events, CDF F&Os Regarding PRA Supporting Requirements F&O ID: Associated SRs: Peer Review CC Assessment:

24-20 QU-F1 Met at CC I-III Capability Category I/II/III Requirements:

DOCUMENT the model quantification in a manner that facilitates PRA applications, upgrades, and peer review.

F&O Issue and Proposed Resolution:

The pie-charts provided in the Quantification notebook section 2.0

SUMMARY

AND CONCLUSIONS, provide CDF and LERF contributions due to initiating events. However, no description is provided on what the initiator basic events represent which makes the reviewer's job difficult to identify them. Also, listing SGTR and SLOCA initiators for each RCS loop adds no value.

Instead, it would be better to provide pie-charts on CDF contributions from hazard groups of event trees such as General Transients, SLOCA, MLOCA, etc.

(This F&O originated from SR QU-F1).

Basis for Significance:

The pie-charts that present CDF contributors need to be revised to facilitate the peer review.

Possible Resolution:

Revise the CDF and LERF contribution pie-charts to more clearly present the contributors. One suggestion is presenting the pie-charts in terms of the event tree contributions to CDF.

Disposition of the Peer Review Finding:

Resolution of F&O:

Information on results is updated based on other model changes. Figures include a legend to support interpretation of results. This closes the finding.

Evaluation of F&O impact on proposed application:

This finding is a documentation issue and has not impact on the risk information and insights associated with the NSWS 30-day CT T.S. LAR.

U.S. Nuclear Regulatory Commission PRA Peer Review Findings and Resolutions CNS-17-014 4.1 Internal Events, CDF F&Os Regarding PRA Supporting Requirements F&O ID: Associated SRs: Peer Review CC Assessment:

24-23 QU-F3 Met at CC II/III Capability Category II/III Requirements:

DOCUMENT the significant contributors (such as initiating events, accident sequences, basic events) to CDF in the PRA results summary. PROVIDE A DETAILED DESCRIPTION OF SIGNIFICANT ACCIDENT SEQUENCES OR FUNCTIONAL FAILURE GROUPS.

F&O Issue and Proposed Resolution:

Table 3 and 4 of the Quantification notebook provide accident sequence contributions. However, it appears that the ATWS contribution is provided for the whole ATWS event, and not on individual sequence level like the other event tree sequences. Since the ATWS CDF contribution appears to be significant, it is important to understand which sequence the most contribution comes from.

(This F&O originated from SR QU-F3).

Basis for Significance:

ATWS sequence contributions are not provided. Instead, overall ATWS contribution to CDF is provided.

Possible Resolution:

Provide in Tables 3 and 4 CDF contributions from ATWS sequences. If any of the sequences turn out to be significant, provide detailed description.

Disposition of the Peer Review Finding:

Resolution of F&O:

Revision of the ATWS event tree provides the inputs necessary to discuss the accident sequence level results that are provided in the quantification notebook (Reference 17). This closes the finding.

Evaluation of F&O impact on proposed application:

This finding is a documentation issue and has no impact on the CNS NSWS 30-day CT T.S. risk insights and information.

U.S. Nuclear Regulatory Commission PRA Peer Review Findings and Resolutions CNS-17-014 4.2 Internal Events, LERF The five (5) Supporting Requirements (SRs) that have been assessed as meeting only Capability Category (CC) I (Appendix D of Reference 3) are discussed, as follows.

F&Os Regarding PRA Supporting Requirements SR ID: Other Affected SRs: Peer Review CC Assessment:

LE-C4 (note, no F&O) Met CC I Capability Category II Requirements:

INCLUDE model logic necessary to provide a realistic estimation of the significant accident progression sequences resulting in a large early release. INCLUDE mitigating actions by operating staff, effect of fission product scrubbing on radionuclide release, and expected beneficial failures in significant accident progression sequences. PROVIDE technical justification (by plant-specific or applicable generic calculations demonstrating the feasibility of the actions, scrubbing mechanisms, or beneficial failures) supporting the inclusion of any of these features.

Issue and Proposed Resolution:

Catawba has a NUREG/CR-6595 LERF model. They updated the 6595 Containment Event Tree (CET) to reflect Catawba specifics. They have incorporated the top logic from the CET into their base model for ease of quantification and to directly capture dependencies from the internal initiator models. However, Catawba does not credit operator actions, release scrubbing or the effects of beneficial failures. The basis is the assumption that these elements do not contribute significantly to reducing the LERF so excluding them to simplify the model is slightly conservative. Catawba meets CC-I based on the use of the NUREG/CR-6595 model. The NRC has explicitly accepted the NUREG/CR-6595 approach as being sufficient for determination of LERF.

Possible Resolution:

Update LERF model to meet Capability Category II Requirements.

Disposition of the Peer Review Finding:

Resolution of F&O:

N/A. Catawba uses a LERF model based on the simplified LERF model in NUREG/CR-6595.

While a NUREG/CR-6595 model is classified as Capability Category I, historically the NRC has indicated that a NUREG/CR-6595 model is of sufficient capability to support risk-informed applications.

Evaluation of F&O impact on proposed application:

The existing LERF analysis is sufficient for the 30-day CT T.S. LAR. The LERF analysis uses the approach contained in NUREG/CR-6595. The use of NUREG/CR-6595 is conservative and overstates the change in LERF risk, thus CC-1 is acceptable for this application.

U.S. Nuclear Regulatory Commission PRA Peer Review Findings and Resolutions CNS-17-014 4.2 Internal Events, LERF F&Os Regarding PRA Supporting Requirements SR ID: Other Affected SRs: Peer Review CC Assessment:

LE-C9 (note, no F&O) LE-C11 Met CC I Capability Category II/III Requirements:

JUSTIFY any credit given for equipment survivability or human actions under adverse environments.

Issue and Proposed Resolution:

Catawba did not take credit for continued operation of equipment in adverse environments.

Catawba has a NUREG/CR-6595 LERF model. The NRC has explicitly accepted the NUREG/CR-6595 approach as being sufficient for determination of LERF.

Possible Resolution:

To move up from CC-I to CC-II/III, Catawba needs to provide documentation of their basis for not crediting continued operation of equipment during a severe accident or following containment failure.

Disposition of the Peer Review Finding:

Resolution of F&O:

N/A. Catawba uses a LERF model based on the simplified LERF model in NUREG/CR-6595.

While a NUREG/CR-6595 model is classified as Capability Category I, historically the NRC has indicated that a NUREG/CR-6595 model is of sufficient capability to support risk-informed applications.

Evaluation of F&O impact on proposed application:

The existing LERF analysis is sufficient for the 30-day CT T.S. LAR. The LERF analysis uses the approach contained in NUREG/CR-6595. The use of NUREG/CR-6595 is conservative and overstates the change in LERF risk, thus CC-1 is acceptable for this application.

U.S. Nuclear Regulatory Commission PRA Peer Review Findings and Resolutions CNS-17-014 4.2 Internal Events, LERF F&Os Regarding PRA Supporting Requirements SR ID: Other Affected SRs: Peer Review CC Assessment:

LE-C11 (note, no F&O) LE-C9 Met CC I Capability Category II Requirements:

JUSTIFY any credit given for equipment survivability or human actions that could be impacted by containment failure.

Issue and Proposed Resolution:

Catawba did not take credit for continued operation of equipment that could be impacted by containment failure. Catawba has a NUREG/CR-6595 LERF model. The NRC has explicitly accepted the NUREG/CR-6595 approach as being sufficient for determination of LERF.

Possible Resolution:

To move up from CC-I to CC-II/III, Catawba needs to provide documentation of their basis for not crediting continued operation of equipment during a severe accident or following containment failure.

Disposition of the Peer Review Finding:

Resolution of F&O:

N/A. Catawba uses a LERF model based on the simplified LERF model in NUREG/CR-6595.

While a NUREG/CR-6595 model is classified as Capability Category I, historically the NRC has indicated that a NUREG/CR-6595 model is of sufficient capability to support risk-informed applications.

Evaluation of F&O impact on proposed application:

The existing LERF analysis is sufficient for the 30-day CT T.S. LAR. The LERF analysis uses the approach contained in NUREG/CR-6595. The use of NUREG/CR-6595 is conservative and overstates the change in LERF risk, thus CC-1 is acceptable for this application.

U.S. Nuclear Regulatory Commission PRA Peer Review Findings and Resolutions CNS-17-014 4.2 Internal Events, LERF F&Os Regarding PRA Supporting Requirements F&O ID: Other Affected SRs: Peer Review CC Assessment:

LE-F1 (note, no F&O) LE-G3 Met CC I Capability Category II/III Requirements:

PERFORM a quantitative evaluation of the relative contribution to LERF from plant damage states and significant LERF contributors from Table 2-2.8-9.

Issue and Proposed Resolution:

Catawba identified the significant contributors to LERF in terms of the initiators and the contributions were quantified. They did not identify the significant contributors in terms of plant damage states or containment failure modes or any of the other contributors listed in Table 2-2.8-3 from RA-Sa-2009.

Possible Resolution:

To move from CC-I to CC-II/III, Catawba needs to evaluate the relative contributions to LERF by such things as plant damage states, accident progression sequences, phenomena, containment challenges, and containment failure modes.

Disposition of the Peer Review Finding:

Resolution of F&O:

The CNS Model Integration and Quantification Notebook (Reference 5) documents the contribution to LERF by initiator, as well as by the most significant LERF sequences. Since Catawba uses a LERF model based on the simplified LERF model in NUREG/CR-6595, and while a NUREG/CR-6595 model is classified as Capability Category I, historically the NRC has indicated that a NUREG/CR-6595 model is of sufficient capability to support risk-informed applications.

Evaluation of F&O impact on proposed application:

The LERF analysis documents the significant contributors to LERF, by initiating events. However, the LERF analysis does not evaluate (and document) the relative contributors to LERF by such things as plant damage states, accident progression sequences, phenomena, containment challenges, and containment failure modes, required to meet CC II/III. The added documentation of the relative contribution of accident progression sequences, phenomena, containment challenges and containment failures modes to LERF has no impact on the changes in LERF due to this application. Therefore, the existing LERF analysis is sufficient for this application.

U.S. Nuclear Regulatory Commission PRA Peer Review Findings and Resolutions CNS-17-014 4.2 Internal Events, LERF F&Os Regarding PRA Supporting Requirements F&O ID: Other Affected SRs: Peer Review CC Assessment:

LE-G3 (note, no F&O) LE-F1 Met CC I Capability Category II/III Requirements:

DOCUMENT the relative contribution of contributors (i.e., plant damage states, accident progression sequences, phenomena, containment challenges, containment failure modes) to LERF.

Issue and Proposed Resolution:

In CNC-1535.00-00-061, Catawba documents the significant contributors to LERF in terms of contribution by initiating events. However, they did not document the relative contribution of contributors such as plant damage states, accident progression sequences, phenomena, containment challenges and containment failure modes.

Possible Resolution:

To move from CC-I to CC-II/III, Catawba needs to evaluate the relative contributions to LERF by such things as plant damage states, accident progression sequences, phenomena, containment challenges, and containment failure modes.

Disposition of the Peer Review Finding:

Resolution of F&O:

N/A. The peer review indicated that SRs LE-G3 and LE-F1 are met at Capability Category I. This category only requires that significant contributors to LERF are documented instead of the more thorough treatment in CC-II/III.

The Catawba LERF model is based on the simplified LERF model in NUREG/CR-6595. While a NUREG/CR-6595 model is classified as CC-I, it has been determined to be of sufficient capability to support risk-informed applications. Consistent with the definition in Section 1-2 of RA-Sa-2009, the following table provides the significant accident progression sequences for the base model used in the CNS LAR submittal:

% Cumulative %

Accident Progression Sequence Contribution Contribution to to LERF LERF Interfacing Systems LOCA 37% 37%

Low Pressure Early Containment Failure 30% 67%

Steam Generator Tube Rupture 26% 93%

ATWS-Related Failure 4% 97%

High Pressure Early Containment Failure 3% 100%

Evaluation of F&O impact on proposed application:

This finding is a documentation issue and will have no impact on the results or conclusions of the CNS NSWS 30-day CT LAR submittal.

U.S. Nuclear Regulatory Commission PRA Peer Review Findings and Resolutions CNS-17-014 4.3 Internal Flooding There are no open findings associated with the internal flooding PRA model.

4.4 High Winds Five (5) High Winds PRA Finding F&Os were generated during the peer review (Ref. 33). Each of the open finding-level F&Os are discussed, as follows:

F&Os Regarding PRA Supporting Requirements F&O ID: Other Affected SRs: Peer Review CC Assessment:

WPR-A1-02 WPR-A1 Met at CC I-III Capability Category II/III Requirements:

Ensure that wind-caused initiating events that give rise to significant accident sequences and/or significant accident progression sequences are included in the wind PRA system model using a systematic process.

F&O Issue and Proposed Resolution:

Assumption 5 in Section 6.1 of CNC-1535.00-00-0154 does not reflect the as built as operated plant and impacts the cutsets in the model.

Basis for Significance:

Assumption 5 in Section 6.1 of the CNC-1535.00-00-0154 states: A high wind initiating event is assumed to have the operators tripping the reactor if there is also high wind failure of a SSC modeled in the fault tree. This assumption is not correct based on the directions provided in procedure RP/0/A/5000/007. Refer to WPR-A1-01 for discussion on consideration of RP/0/A/5000/007 procedures. This does not reflect the as built as operated plant and impacts the cutsets in the model.

Possible Resolution:

Proposed solution: Delete this assumption and associated model logic.

Disposition of the Peer Review Finding:

Resolution of F&O:

CNC-1535.00-00-0154 (Reference 21) has been updated and the assumption has been deleted.

The model logic has been revised as discussed in the resolution to WPR A1-01 to induce a reactor trip for each applicable High Wind initiating event per the procedure.

Evaluation of F&O impact on proposed application:

The updated high winds model was used in the risk analysis to support the NSW 30-day CT LAR.

U.S. Nuclear Regulatory Commission PRA Peer Review Findings and Resolutions CNS-17-014 4.4 High Winds F&Os Regarding PRA Supporting Requirements F&O ID: Other Affected SRs: Peer Review CC Assessment:

WPR-A1-03 WPR-A1 Met at CC I-III Capability Category I/II/III Requirements:

Ensure that wind-caused initiating events that give rise to significant accident sequences and/or significant accident progression sequences are included in the wind PRA system model using a systematic process.

F&O Issue and Proposed Resolution:

No discussion was provided for screening out failure modes that could result in the loss of ultimate heat sink due to a high wind event other than those related to the class 1 structures housing the service water system.

Basis for Significance:

No discussion was provided for screening out failure modes that could result in the loss of ultimate heat sink due to a high wind event other than those related to the class 1 structures housing the service water system. This initiator is important since it can affect multiple units. For example, no discussion was provided to evaluate the consequences of wind borne debris being deposited in the lake supplying safety related cooling water and choking off the intake. The basis for significance is that a potential major initiator that can affect both units is not evaluated.

Possible Resolution:

Evaluate the potential for losing ultimate heat sink due to debris blocking the intake.

Disposition of the Peer Review Finding:

Resolution of F&O:

The potential for losing the ultimate heat sink due to debris blocking the intake has been evaluated and judged to be insignificant in the CNS HWPRA. The CNS units intake suction through piping is located near the bottom of the lake. It is considered unlikely that sufficient debris would be deposited in the lake and that the debris would sink to the low level intake and plug the intake.

Documentation of this judgment has been added to Section 7.3.2.16 of CNC-1535.00-00-0154, Revision 1 (Reference 21).

Evaluation of F&O impact on proposed application:

Due to the location of the intake suction at CNS, the potential for losing the ultimate heat sink is negligible. Therefore, this finding has no impact on the 30-day CT T.S. LAR.

U.S. Nuclear Regulatory Commission PRA Peer Review Findings and Resolutions CNS-17-014 4.4 High Winds F&Os Regarding PRA Supporting Requirements F&O ID: Other Affected SRs: Peer Review CC Assessment:

WPR-A4-01 WPR-A4 Met at CC I-III Capability Category I/II/III Requirements:

In each of the following aspects of the high wind PRA systems-analysis work, SATISFY the corresponding requirements in Part 2, except where they are not applicable or where this Part includes additional requirements. DEVELOP a defined basis to support the claimed non-applicability of any exceptions. The aspects governed by this requirement are:

(a) initiating-event analysis, (b) accident-sequence analysis, (c) success-criteria analysis, (d) systems analysis, (e) data analysis, (f) human-reliability analysis, (g) use of expert judgment.

When Part 2 requirements are used FOLLOW the capability category designations in Part 2, and for consistency use the same Capability Category in this analysis.

F&O Issue and Proposed Resolution:

There was no documented evidence in CNC-1535.00-00-0154, the CNS HWPRA report, to show that the high wind PRA systems-analysis work SATISFIES the corresponding requirements in Part 2 (of the PRA Standard). A defined basis to support non-applicability of any exceptions was not included. Peer Review Team did not have the time to perform a detailed review of the assessment of Part 2 i.e.,

P2A (calc DPC-1535.00-00-0013, PRA Quality Self-Assessment, received during the review). It is not within the scope that the Peer Review Team scope to perform the assessment of Part 2 SRs as part of this Peer Review. Without being provided with a compliance review of Part 2 SRs, the Peer Review also cannot judge that the technical elements as specified in the applicable Part 2 SRs are satisfied or not. So this F&O asks for more evidence and systematic assessment of the applicable SRs in Part 2 to meet this SR WPR-A4 and document it accordingly.

Basis for Significance:

No evidence of satisfying the requirements of Part 2, or basis for exceptions to the requirements, was provided or cross referenced in CNC-1535.00-00-0154, the CNS HWPRA report.

Possible Resolution:

Document that the requirements of Part 2 are satisfied. Whenever an exception is taken, the PRA team needs to be cognizant of the underlying rationale for the specific Part 2 requirement so as to ensure that this rationale is considered when the exception is taken.

U.S. Nuclear Regulatory Commission PRA Peer Review Findings and Resolutions CNS-17-014 4.4 High Winds Disposition of the Peer Review Finding:

Resolution of F&O:

Documentation of compliance to the part 2 SRs is provided in Appendix H of CNC-1535.00-00-0154 (Reference 21). The CNS Internal Events model has undergone an update to comply with R.G.

1.200, Rev. 2.

Evaluation of F&O impact on proposed application:

This finding is a documentation issue and resulted in no changes to the CNS High Winds PRA.

Therefore, this finding has no impact on the 30-day CT T.S. LAR.

U.S. Nuclear Regulatory Commission PRA Peer Review Findings and Resolutions CNS-17-014 4.4 High Winds F&Os Regarding PRA Supporting Requirements F&O ID: Other Affected SRs: Peer Review CC Assessment:

WPR-A4-02 WPR-A4 Met at CC I-III Capability Category I/II/III Requirements:

In each of the following aspects of the high wind PRA systems-analysis work, SATISFY the corresponding requirements in Part 2, except where they are not applicable or where this Part includes additional requirements. DEVELOP a defined basis to support the claimed non-applicability of any exceptions. The aspects governed by this requirement are:

(a) initiating-event analysis, (b) accident-sequence analysis, (c) success-criteria analysis, (d) systems analysis, (e) data analysis, (f) human-reliability analysis, (g) use of expert judgment.

When Part 2 requirements are used FOLLOW the capability category designations in Part 2, and for consistency use the same Capability Category in this analysis.

F&O Issue and Proposed Resolution:

Peer Review Team disagrees with some of the assessment results as stated in the P2A (DPC-1535.00-00-0013, PRA Quality Self-Assessment) report.

Basis for Significance:

For example, SY-B7 is a CCI because the base system modeling used conservative success criteria verses realistic as required to meet CCII - the peer review team would need to review the high wind analysis in detail to understand if the risk importance of low speed straight winds is justified. Also, given its significant importance the impact of the siding on the AC system should be documented in a system notebook to address compliance with SY-B9. We see no evidence that the AC notebook was modified or that a new notebook addressing structures was developed in compliance with SY-B9; e.g. siding integrity is essential for maintaining the integrity of the AC system during "low-speed" straight winds - these issues need to be documented in accordance with the SY requirements described in SY-C2.

Possible Resolution:

Review the P2A assessment in detail, correct any errors and enhance the documentation. If a materially important mistake is discovered, its impact shall be analyzed and appropriate action taken.

U.S. Nuclear Regulatory Commission PRA Peer Review Findings and Resolutions CNS-17-014 4.4 High Winds Disposition of the Peer Review Finding:

Resolution of F&O:

Documentation of compliance to the part 2 SRs is provided in Appendix H of CNC-1535.00-00-0154 (Reference 21). The CNS Internal Events model has undergone an update to comply with R.G.

1.200, Rev. 2.

Evaluation of F&O impact on proposed application:

This finding is a documentation issue and resulted in no changes to the CNS High Winds PRA.

Therefore, this finding has no impact on the 30-day CT T.S. LAR.

U.S. Nuclear Regulatory Commission PRA Peer Review Findings and Resolutions CNS-17-014 4.4 High Winds F&Os Regarding PRA Supporting Requirements F&O ID: Other Affected SRs: Peer Review CC Assessment:

WPR-C3-01 WPR-A10 Met at CC I-III Capability Category I/II/III Requirements:

DOCUMENT the sources of model uncertainty and related assumptions associated with the high wind plant response model development.

F&O Issue and Proposed Resolution:

Several assumptions in CNC-1535.00-00-0154, Section 6.1 need clarification:

Assumption 1 states that one tornado missile hit to a PRA SSC results in functional failure except for the main transformers in the Yard, which require two missile hits. No basis is provided for this assumption. Was fragility of the component considered or was any missile at any speed assumed to result in a functional failure? Why are two missiles needed for a functional failure of the main transformers and how were the two missiles modeled?

Assumption 4 states: F2 and greater peak gusty winds at CNS will automatically induce a LOOP.

Explain basis why an F2 or greater peak gust wind automatically induces LOOP verses >F1 or >F3, for example.

Assumption 5 states: A high wind initiating event is assumed to have the operators tripping the reactor if there is also high wind failure of a SSC modeled in the fault tree. Procedure RP/0/A/5000/007 requires operators take both units to hot shutdown for winds 73mph or higher -

without a concurrent high wind failure. As such this assumption does not reflect the way the operators will respond.

Assumption 6 states: Conservatism is introduced when initiating event %T3, LOOP, and the High Wind-Induced LOOP events are ORd under the same parent gate. Some High Wind- Induced LOOP events may be double counted due to inclusion in the %T3 model frequency. A characterization (e.g. sensitivity analysis) of the impact of this assumption on the model results is needed.

Assumption 7 states: Some components were modeled by high wind analysis. These components had no representation in the fault tree. Only one of the four MSSVs and one of the four MSIVs are modeled due to the Internal Event model assumption of a SGTR occurs on SG B. Thus high wind fragilities on the other three MSSV/MSIVs are not in the fault tree. The example did not clarify if this is a conservative assumption - please explain why this assumption is conservative and if a sensitivity analysis is needed.

Assumption 8 states: System YD is the Drinking Water System is assumed failed for all high wind events. A basis is needed for this assumption including the impact on model results.

Assumption 11 states: This analysis is for Unit 1 with shared Unit 2 SSCs. Applicability of Unit 2 with shared Unit 1 SSCs is assumed for this analysis. Explain why was Unit 1 selected and why a

U.S. Nuclear Regulatory Commission PRA Peer Review Findings and Resolutions CNS-17-014 4.4 High Winds Unit 2 model is not needed. Is the internal events CDF and LERF for Unit 1 significantly different from Unit 2?

Assumption 1 in CNC-1535.00-00-0154 Appendix A Section B.1. CREDITING RECOVERY OF SEAL INJECTION AFTER FIRST HOUR states: Straight Line or Tornado Wind conditions will not prevent access to the SSF after one hour has elapsed from the Wind-Induced LOOP Initiating Event. What Is the basis for this assumption? It is previously stated that reaching the SSF requires travelling 100 feet outside. Debris and structural integrity issues may preclude using this path, this would not only lead to path and door blockage but personnel safety. Access may require obtaining debris removal equipment and performing structural reviews which may not be possible within one hour. In addition, the Calculation Section Addressed reference for SR WPR-C3 in Table 4-1 should be Section 6.0, not Section 7.3.3.

Basis for Significance:

Basis for key assumptions must be clear to facilitate review, applications and future updates.

Possible Resolution:

Improve the quality of the assumptions in the High Wind PRA report.

Disposition of the Peer Review Finding:

Resolution of F&O:

CNC-1535.00-00-0154 (Reference 21) has been updated such that each of the assumptions has been reviewed and several have been clarified and/or enhanced. Assumptions 1, 5, 6, and 7 have been deleted as they were determined to be inapplicable. Assumptions 4, 8 and 11 have been revised and enhanced to provide a clearer basis for each. Assumption 1 has been moved to Appendix G and has also been revised and enhanced.

The only model change that resulted from the review of the aforementioned assumptions is the assumption that a failure of a PRA SSC is required to induce the wind initiating event in the model.

This assumption has been removed as a reactor trip was modeled in accordance with RP/0/A/5000/007.

Evaluation of F&O impact on proposed application:

The changes made in the analysis with respect to initiating events was explicitly encompassed in the analysis and thus no additional changes are required for this application.

U.S. Nuclear Regulatory Commission PRA Peer Review Findings and Resolutions CNS-17-014 4.5 Fire Twenty (20) Fire PRA F&O Findings were generated during the FPRA Peer Review (Reference 31).

Additionally, 16 Supporting Requirements (SRs) have been assessed as meeting Capability Category (CC) I. Each of these finding-level F&Os, as well as the SRs assessed as meeting CC-I, are discussed, as follows.

F&Os Regarding PRA Supporting Requirements SR ID: Other Affected SRs: Peer Review CC Assessment:

PP-B3 (Note: not a finding) CC-I Capability Category II/III Requirements:

If spatial separation is credited as a partitioning feature, JUSTIFY the judgment that spatial separation is sufficient to substantially contain the damaging effects of any fire that might be postulated in each of the fire compartments created as a result of crediting this feature.

F&O Issue and Proposed Resolution:

Spatial separation not relied upon for compartment assignments. This SR, PPB3, is judged to be met at CC-I since no spatial separation is credited and CCII/III requires crediting of spatial separation as credited in the regulatory fire protection program.

Basis for Significance:

As performed, adequate compartmentalization is used for the Catawba Fire PRA.

Possible Resolution:

For CC-II/III, refinement of the compartmentalization to smaller compartment is required.

Disposition of the Peer Review Assessment:

Resolution of F&O:

No steps taken. CNS does not credit spatial separation as a partitioning feature.

Evaluation of F&O impact on proposed application:

CC-I is sufficient for this application as CC-II/III is not applicable to CNS. Therefore, this F&O has no impact on the NSWS 30-day CT T.S. LAR.

U.S. Nuclear Regulatory Commission PRA Peer Review Findings and Resolutions CNS-17-014 4.5 Fire F&Os Regarding PRA Supporting Requirements SR ID: Other Affected SRs: Peer Review CC Assessment:

PP-B5 (Note: not a finding) CC-I Capability Category II/III Requirements:

DEFINE and JUSTIFY the basis and criteria applied when active fire barrier elements (such as normally open fire doors, water curtains, and fire dampers) are credited in partitioning.

F&O Issue and Proposed Resolution:

No active fire barrier elements are credited for Catawba Fire PRA compartmentalization. The credited passive fire barriers correspond to barriers credited in the regulatory fire protection program.

Basis for Significance:

The Catawba Fire PRA credits only fire-rated passive barriers. In order to meet Capability Category II/III, crediting of active fire barrier elements in fire compartment boundaries, with appropriate justification, is necessary.

Possible Resolution:

If CC-II/III is desired, the PRA compartmentalization must be revised with credits for active fire barriers, with appropriate justification Disposition of the Peer Review Assessment:

Resolution of F&O:

No steps taken. CNS does not credit active fire barrier elements for partitioning.

Evaluation of F&O impact on proposed application:

CC-I is sufficient for this application as CC-II/III is not applicable to CNS. Therefore, this F&O has no impact on the NSWS 30-day CT T.S. LAR.

U.S. Nuclear Regulatory Commission PRA Peer Review Findings and Resolutions CNS-17-014 4.5 Fire F&Os Regarding PRA Supporting Requirements F&O ID: Other Affected SRs: Peer Review CC Assessment:

CS-A11-01 CS-C3 Not Met Capability Category I/II/III Requirements:

If assumed cable routing used in the Fire PRA, IDENTIFY the scope and extent, and PROVIDE a basis for the assumed cable routing.

F&O Issue and Proposed Resolution:

The "Y3" assessment in Appendix B of CNC-1535.00-00-0109 excludes cables for a small number of components that are not in the ARTRAK (e.g., 7 components for Air). While the routing of the cables from the electrical panel to the compressor may be sufficient to determine that power is available, the compressor itself has instrumentation and controls, that could cause spurious trips or spurious starts that do not appear to be included in the review of Y3 components and may not be limited to the routing areas in the assumed routing.

For instance, the compressor control cable will likely go to the control room for switches, alarms and other controls. Similar information would be needed for other systems credited in the Y3 list as well.

Basis for Significance:

Justify that the components in Appendix B are the only components needed to credit that system or train for "Y3". Also provide additional justification that the cable routing through the zones specified are the only routings needed to support the system Possible Resolution:

Provide any additional discussion and justification on the Y3 selection and routings that would be sufficient for an independent review to verify that the system can be credited in any area where it is excluded from the UNL.

Disposition of the Peer Review Finding:

Resolution of F&O:

An assumption was added to the FPRA Summary Report (Reference 22) to indicate Y3 components are based on assumed routing. The Y3 list of basic events was developed considering both power and control cables in which each Y3 component could be credited. Sensitivity analysis performed in the FPRA Summary Report show that the impact of the Y3 components on quantification is relatively minimal. Credit by exclusion was used as a reasonable alternative to cable routing of FPRA components of lesser importance.

Evaluation of F&O impact on proposed application:

Based on the aforementioned sensitivity, there is no impact to the NSWS 30-day CT T.S. LAR risk insights.

U.S. Nuclear Regulatory Commission PRA Peer Review Findings and Resolutions CNS-17-014 4.5 Fire F&Os Regarding PRA Supporting Requirements SR ID: Other Affected SRs: Peer Review CC Assessment:

CS-B1-01 CS-C4 CC-I Capability Category II/III Requirements:

Analyze all electrical distribution buses credited in the Fire PRA plant response model for proper overcurrent coordination and protection and IDENTIFY any additional circuits and cables whose failure could challenge power supply availability due to inadequate electrical overcurrent protective device coordination.

F&O Issue and Proposed Resolution:

CNS performed a review of their existing electrical over-current coordination and protection analysis. As a result of this review, CNS identified a number of deficiencies in terms of scope and level of detail. CNS is currently in the process of completely redoing their electrical over-current coordination and protection analysis. The new analysis will increase the level of detail and to increase the scope to include all Appendix R equipment, the PRA equipment and the NPO equipment. As part of this re-analysis, CNS is making plant modifications as needed. However, at this time, this analysis is not complete.

Basis for Significance:

A review of the existing electrical over-current coordination and protection analysis is required to meet the SR even at the CC-I level.

Possible Resolution:

To move from CC-I up to CC-II/III complete the on-going update of the electrical over-current coordination and protection analysis and formally issue the report.

Disposition of the Peer Review Assessment:

Resolution of F&O:

The update of the breaker coordination and protection analysis was completed subsequent to the peer review and has since been incorporated into the FPRA. Breaker coordination related interlocks from pseudo components modeled in DATATRAK that were tabulated in Section 6.0 of the Catawba Appendix R Coordination Study (Reference 23) have been included in the FPRA and is described in Section 5.0 of the CNS FPRA Cable Selection Report (Reference 24).

Evaluation of F&O impact on proposed application:

Breaker coordination issues are not a consideration for the SNSWP single header operation other than what has been included in the FPRA. Therefore, this F&O has no impact on the 30-day CT T.S. LAR risk insights.

U.S. Nuclear Regulatory Commission PRA Peer Review Findings and Resolutions CNS-17-014 4.5 Fire F&Os Regarding PRA Supporting Requirements F&O ID: Other Affected SRs: Peer Review CC Assessment:

ES-C1-01 ES-D1, HRA-A1, HRA-B4 Not Met Capability Category I/II/III Requirements:

IDENTIFY instrumentation that is relevant to the operator actions for which HFEs are defined or modified to account for the context of fire scenarios in the Fire PRA, per SRs HRA-B1 and HRA-B2.

F&O Issue and Proposed Resolution:

HRA events are reviewed for instrumentation in Attachment B of CNC-1535.00-00-0108, Rev. 0.

The documentation for HRA events that do not have instrumentation in the internal events model is not clear. Instrumentation is described in general terms without any information on the number of trains or the number of instruments available. There is not enough documentation to justify the diverse and redundant argument.

Basis for Significance:

Based on the available documentation, reviewers were unable to determine if the instrumentation supporting credited HRA events was diverse and redundant enough to credit associated events.

Possible Resolution:

Provided additional details on the number, type, and trains of instrumentation being credited.

Disposition of the Peer Review Finding:

Resolution of F&O:

Additional details were added to Appendix B of the Component Selection Calculation (Reference

25) to support the redundant (multiple trains) and diverse (multiple parameters such as level and pressure) argument.

Evaluation of F&O impact on proposed application:

This finding is a documentation issue and has no impact on the 30-day CT T.S. LAR risk insights.

U.S. Nuclear Regulatory Commission PRA Peer Review Findings and Resolutions CNS-17-014 4.5 Fire F&Os Regarding PRA Supporting Requirements F&O ID: Other Affected SRs: Peer Review CC Assessment:

ES-C2-01 ES-D1, HRA-A3 Not Met Capability Category I/II Requirements:

IDENTIFY instrumentation associated with each operator action to be addressed, based on the following: fire-induced failure of any single instrument whereby one of the modes of failure to be considered is spurious operation of the instrument. and fire-induced failure, including spurious indication, even if they are not relevant to the HFEs for which instrumentation is identified within the scope defined by ES-C1, if the failure could cause an undesired operator action related to that portion of the plant design credited in the analysis.

F&O Issue and Proposed Resolution:

The Equipment Selection Calculation CNC-1535.00-00-0108 revision 0, addresses spurious instrumentation under "Errors of Commission". This section states "No specific instruments were identified that would cause an undesired operator action without first taking one or more confirmatory actions". The results of the assessment are provided, but no details are provided on who performed the review, what method was used, and what procedures were reviewed.

Basis for Significance:

There is not sufficient documentation to determine the SR is met.

Possible Resolution:

Add documentation describing what procedures were reviewed, what method was applied during the review, and what the qualification of the individual performing the review was.

Disposition of the Peer Review Finding:

Resolution of F&O:

The Component Selection Calculation (Reference 25) has been updated to include additional details of the instrument review including the names of the reviewers. Using the guidance provided in Section 9.7 of the calculation and their firsthand knowledge of CNS, the reviewers evaluated the applicable EP(s), AOP(s) & AP(s) in order to determine the important parameters that would be relied on for successful execution of each modeled operator action.

Evaluation of F&O impact on proposed application:

The changes made to the modeled operator actions as a result of this finding were included in the analysis and thus no additional changes are required for this application.

U.S. Nuclear Regulatory Commission PRA Peer Review Findings and Resolutions CNS-17-014 4.5 Fire F&Os Regarding PRA Supporting Requirements F&O ID: Other Affected SRs: Peer Review CC Assessment:

PRM-B2-01 PRM-C1 Not Met Capability Category I/II/III Requirements:

VERIFY the peer review exceptions and deficiencies for the Internal Events PRA are dispositioned, and the disposition does not adversely affect the development of the Fire PRA plant response model.

F&O Issue and Proposed Resolution:

Section 4 of CNC-1535.00-00-0111 addresses PRA model quality for Fire PRA use. Two potentially significant items not addressed are the HRA pre-initiators (HR-A3 and HR-D6) and failure probability data (DA-B1) from DPC-1535.00-00-0013, Rev. 2. Section 4 of the FPRA Model Development should address these two items.

Basis for Significance:

The SR PRM-B2 was judged to be not met because of this deficiency.

Possible Resolution:

Address all internal events peer review findings.

Disposition of the Peer Review Finding:

Resolution of F&O:

The HEPs have been quantified using the mean values in the FPRA. There were no internal events peer review findings against HR-A3. No changes have been made to the FPRA. Compared to post-initiator HEPs and fire induced failures, latent human error probabilities, equipment random failure rates and maintenance unavailability, calibration HEPs and misalignment of multiple trains of equipment are not expected to contribute significantly to the overall equipment unavailability. Thus there is no material impact on the FPRA. The Internal Events Peer Review identified a finding against DA-B1 (F&O DA-01) which noted that the data development workplace procedure did not identify component boundaries. The finding went on to note that component boundaries are apparent from the data. The change to the workplace procedure does not impact the FPRA quantification and no examples where the data was found to be incorrect were identified. Modest changes to the random failure rates have little impact on the results as fire induced failures are far more significant in the FPRA results.

Evaluation of F&O impact on proposed application:

This finding has no impact on the FPRA. Therefore, there is no impact to the 30-day CT T.S. LAR risk insights.

U.S. Nuclear Regulatory Commission PRA Peer Review Findings and Resolutions CNS-17-014 4.5 Fire F&Os Regarding PRA Supporting Requirements F&O ID: Other Affected SRs: Peer Review CC Assessment:

PRM-B5-01 FQ-A2 Met Capability Category I/II/III Requirements:

For those fire-induced initiating events included in the Internal Events PRA model, REVIEW the corresponding accident sequence models and (a) IDENTIFY any existing accident sequences that will require modification based on unique aspects of the plant fire response procedures in accordance with HLR-AS-A and HLR-AS-B of Part 2 and their supporting requirements and (b) IDENTIFY any new accident sequences that might result from a fire event that were not included in the Internal Events PRA in accordance with HLR-AS-A and HLR-AS-B of Part 2 and their supporting requirements.

F&O Issue and Proposed Resolution:

Reactor trip was used for fire initiating events in the model, although feedwater is failed due to lack of routing information. The plant response model is not the same for the plant trip and loss of feedwater initiating events; for example the probability of lifting a PORV or SRV is 1E-2 for loss of feedwater and 1E-3 for plant trip.

Basis for Significance:

Plant response model does not represent bounding assumptions.

Possible Resolution:

Consider using a loss of feedwater initiator where appropriate.

Disposition of the Peer Review Finding:

Resolution of F&O:

The FPRA was modified to add gate IEFIRES which enables the fire initiating events to inherit the plant response logic for any transient event. The transient logic in the IEPRA and consequently the FPRA includes transfers to all of the necessary support systems logic. Updated Section 6.3 of the FPRA Model Development Report (Reference 26).

Evaluation of F&O impact on proposed application:

The results of this finding have been incorporated into the FPRA. Therefore, there is no further analysis needed.

U.S. Nuclear Regulatory Commission PRA Peer Review Findings and Resolutions CNS-17-014 4.5 Fire F&Os Regarding PRA Supporting Requirements F&O ID: Other Affected SRs: Peer Review CC Assessment:

PRM-B6-01 PRM-B11, PRM-C1, HRA-D1 Met Capability Category I/II/III Requirements:

MODEL accident sequences for any new initiating events identified per PRM-B3 and any accident sequences identified per PRM-B5 reflective of the possible plant responses to the fire-induced initiating events in accordance with HLR-AS-A and HLR-AS-B and their SRs in Part 2 with the following clarifications, and DEVELOP a defined basis to support the claim of non-applicability of any of the following requirements in Part 2:

(a) All the SRs under HLR-AS-A and HLR-AS-B in Part 2 are to be addressed in the context of fire scenarios including effects on equipment, associated cabling, operator actions, and accident progression and timing.

(b) When applying AS-A5 in Part 2 to Fire PRA, INCLUDE consideration of fire response procedures as well as emergency operating procedures and abnormal procedures.

F&O Issue and Proposed Resolution:

CNS added several new accident sequences to address some fire-specific issues that were not part of the base PRA. The model was reviewed and generally found to follow the process from the internal events PRA. One issue was identified: One of the new sequences included a new operator action, TSSPZRLRHE, but the documentation did not provide a basis for the assumed timing. In the HRA quantification section, CNS indicated that this was an ex-control room action with more an hour was available to perform the action.

However, CNS did not provide the basis for saying that more than an hour was available.

Basis for Significance:

This important information needs to be documented in relation to inclusion of a new operator action in the PRA.

Possible Resolution:

CNS needs to explicitly define the basis for stating that more than an hour is available to perform an ex-control room fire-specific action. Also, CNS should review all ex-control room actions to confirm that they have reasonable bases for the assumed time available.

Disposition of the Peer Review Finding:

Resolution of F&O:

Section 5.1 of the FPRA Model Development report has been updated to provide additional basis for the action and the assumed HEP value. Note that this HEP is not an NFPA 805 fire-specific recovery event.

Evaluation of F&O impact on proposed application:

This finding has no impact on the risk metrics used for the NSWS 30-day CT T.S. application.

U.S. Nuclear Regulatory Commission PRA Peer Review Findings and Resolutions CNS-17-014 4.5 Fire F&Os Regarding PRA Supporting Requirements F&O ID: Other Affected SRs: Peer Review CC Assessment:

PRM-B7-01 PRM-C1 Not Met Capability Category I/II/III Requirements:

IDENTIFY any cases where new or modified success criteria will be needed to support the Fire PRA consistently with the HLR-SC-A and HLR-SC-B of Part 2 and their supporting requirements.

F&O Issue and Proposed Resolution:

The self-assessment indicated that success criteria issues were considered in the Model Development Report. However, no evidence could be found that success criteria had been discussed in the Model Development report.

Basis for Significance:

No evidence could be found that success criteria had been discussed in the Model Development report. It is important to document the impact of fire on the PRA success criteria.

Possible Resolution:

Provide expanded documentation of fire-specific influences on success criteria.

Disposition of the Peer Review Finding:

Resolution of F&O:

A discussion addressing success criteria has been added to Section 3.5 of the FPRA Model Development Report (Reference 26).

Evaluation of F&O impact on proposed application:

This finding is a documentation issue and has no impact on the NSWS 30-day CT T.S. risk insights.

U.S. Nuclear Regulatory Commission PRA Peer Review Findings and Resolutions CNS-17-014 4.5 Fire F&Os Regarding PRA Supporting Requirements F&O ID: Other Affected SRs: Peer Review CC Assessment:

FSS-A1-01 Met Capability Category I/II/III Requirements:

IDENTIFY all risk-relevant ignition sources, both fixed and transient, in each unscreened physical analysis unit within the global analysis boundary.

F&O Issue and Proposed Resolution:

Documentation of the potential sources of fire in each compartment has not been completed.

Basis for Significance:

If all fire scenarios are not identified, the overall event frequency is potentially less than the value that should be used in the FPRA analysis. This has the potential to underestimate the CDF and LERF results of the FPRA Possible Resolution:

Utilize the NFPA 805 Ignition Source Walkdown report [CNC -1435.00-00-0048] to revise Appendix A of the CNS Fire Scenario Report [CNC-1535.00-00-0110] so that all fire scenarios are listed.

Disposition of the Peer Review Finding:

Resolution of F&O:

Previously, the CNS Fire Scenario Report, CNC-1535.00-00-0110 did not list ignition sources that were screened from quantification. Appendix A of the Fire Scenario Report (Reference 27) has been updated to list the ignition sources that were screened from quantification for each fire compartment.

Evaluation of F&O impact on proposed application:

This finding is a documentation issue and has no impact on the NSWS 30-day CT T.S. risk insights.

U.S. Nuclear Regulatory Commission PRA Peer Review Findings and Resolutions CNS-17-014 4.5 Fire F&Os Regarding PRA Supporting Requirements F&O ID: Other Affected SRs: Peer Review CC Assessment:

FSS-A2-01 Met Capability Category I/II/III Requirements:

GROUP all risk-relevant damage targets in each unscreened physical analysis unit within the global analysis boundary into one or more damage target sets and for each target set, SPECIFY the equipment and cable failures, including specification of the failure modes.

F&O Issue and Proposed Resolution:

Target Sets and related Failure Modes are not listed in a comprehensive and organized fashion, and then linked to ignition sources:

Example: For FA1, targets are: (1) trays in pump room, (2) NS Pump 1A motor, (3) NS Pump 1A itself, (4) NS Pump 1A motor junction box, etc. Tray identification may be needed in some fire areas. Then postulated ignition sources are linked to each target or group of targets (i.e., oil leak to pump and motor, etc.).

Basis for Significance:

Damage target sets may exist that are not evaluated.

Possible Resolution:

Provide additional analysis to identify and group the failure modes (i.e. damage targets) into a comprehensive table and indicate applicability of each grouping to each fire area.

Disposition of the Peer Review Finding:

Resolution of F&O:

No steps taken. However, Appendix A of the CNS Fire Scenario Development Report (Reference

27) lists the ignition source and targets within the zone of influence of the ignition source only. It was considered unpractical to group target sets and then locate ignition sources. However, a list of scenarios where a specific target was impacted (either a tray target or a specific component) can be derived using the FRANC database. Note that all of the targets in the NS and ND pump rooms in Fire Area 1 were assumed to be failed by the pump fire.

Evaluation of F&O impact on proposed application:

This finding is a documentation issue and has no impact on the NSWS 30-day CT T.S. risk insights.

U.S. Nuclear Regulatory Commission PRA Peer Review Findings and Resolutions CNS-17-014 4.5 Fire F&Os Regarding PRA Supporting Requirements F&O ID: Other Affected SRs: Peer Review CC Assessment:

FSS-B2 (Note: not an F&O) CC I Capability Category II Requirements:

SELECT one or more fire scenarios, either in the MCR or elsewhere, leading to MCR abandonment and/or a reliance on ex-control room operator actions including remote and/or alternate shutdown actions, consisting of a combination of an ignition source (or group of ignition sources) such that the selected scenarios provide reasonable assurance that the MCR abandonment fire risk contribution can be realistically characterized.

F&O Issue and Proposed Resolution:

Section 3.1.3 of CNC-1535.00-00-010 and Appendix E of that document identifying fire driven parameters necessitating abandonment discuss the conditions that are assumed for fire scenarios W1 and W2 addressed in the document. A bounding type analysis for the control room was performed. A realistic characterization is required to achieve Capability Category II. The scenario analyzed are bounding in nature but could be tweaked for more realism and analysis with additional detail in order to achieve a Capability Category II rating.

Basis for Significance:

Analysis presented satisfies Capability Category I requirements.

Possible Resolution:

If Capability Category II is desired, perform additional control room analysis with more realistically modeled scenarios, crediting panel design and other specific features of the Catawba control room design.

Disposition of the Peer Review Finding:

Resolution of F&O:

No additional scenarios created. Using a bounding analysis approach is sufficient.

Evaluation of F&O impact on proposed application:

The Main Control Room scenarios are sufficient for this application. No changes were made to the FPRA. Therefore, this finding has no impact on the NSWS 30-day CT T.S. risk insights.

U.S. Nuclear Regulatory Commission PRA Peer Review Findings and Resolutions CNS-17-014 4.5 Fire F&Os Regarding PRA Supporting Requirements F&O ID: Other Affected SRs: Peer Review CC Assessment:

FSS-C1 (Note: not an F&O) CC I Capability Category II Requirements:

For each selected fire scenario, ASSIGN characteristics to the ignition source using a two-point fire intensity model that encompass low likelihood, but potentially risk contributing, fire events in the context of both fire intensity and duration given the nature of the fire ignition sources present.

F&O Issue and Proposed Resolution:

A two-point treatment was used for isolated selected scenarios such as low energy panels but not for "each selected" scenario.

Basis for Significance:

Analysis performed addresses Capability Category I requirements and more but not to the extent to qualify for a Capability Category II rating.

Possible Resolution:

If Capability Category II rating is desired, a preponderance of evaluated scenarios should be evaluated using two-point methodology.

Disposition of the Peer Review Finding:

Resolution of F&O:

The Fire PRA analysis was updated to increase the number of scenario refinements using a 2-point treatment.

Evaluation of F&O impact on proposed application:

The resolution of this finding has been incorporated into the analysis. Therefore, no further analysis needs to be performed to determine the impact on the NSWS 30-day CT T.S. risk insights.

U.S. Nuclear Regulatory Commission PRA Peer Review Findings and Resolutions CNS-17-014 4.5 Fire F&Os Regarding PRA Supporting Requirements F&O ID: Other Affected SRs: Peer Review CC Assessment:

FSS-C2 (Note: not an F&O) CC I Capability Category II/III Requirements:

CHARACTERIZE ignition source intensity using a realistic time-dependent fire growth profile (i.e., a time-dependent heat release rate) for significant contributors as appropriate to the ignition source.

F&O Issue and Proposed Resolution:

Peak heat release rates reflected in NUREG 6850 were utilized. Time-dependent growth heat release rate curves were not discussed.

Basis for Significance:

Analysis performed meets industry practice.

Possible Resolution:

If Capability Category II/III is desired, additional analysis considering the impact of fuel exhaustion in each compartment is required.

Disposition of the Peer Review Finding:

Resolution of F&O:

Time dependent HRR profiles have since been incorporated into numerous high risk scenarios.

Analysis has since been updated.

Evaluation of F&O impact on proposed application:

The impact of this finding has been incorporated into the analysis. Therefore, no further action is needed.

U.S. Nuclear Regulatory Commission PRA Peer Review Findings and Resolutions CNS-17-014 4.5 Fire F&Os Regarding PRA Supporting Requirements F&O ID: Other Affected SRs: Peer Review CC Assessment:

FSS-C3 (Note: not an F&O) CC I Capability Category II Requirements:

JUSTIFY the heat release rate profile stages included in the analysis (i.e., fire growth, steady burning or decay stages).

F&O Issue and Proposed Resolution:

Burn out was considered in analysis for hot gas layer impact but did not seem to be based on fuel exhaustion but rather taking the room condition to total involvement. Additional discussion and detail addressing fuel exhaustion is required for improved rating.

Basis for Significance:

Analysis performed appears to satisfy requirement but does not address detail for higher than Capability Category I rating.

Possible Resolution:

If Capability Category II/III is desired, additional analysis considering the impact of fuel exhaustion in each compartment is required.

Disposition of the Peer Review Finding:

Resolution of F&O:

No further action required. CC-I for this SR is acceptable for this application.

Evaluation of F&O impact on proposed application:

No changes were made to the fire model due to this finding; therefore, it has no impact on the NSWS 30-day CT T.S. risk insights.

U.S. Nuclear Regulatory Commission PRA Peer Review Findings and Resolutions CNS-17-014 4.5 Fire F&Os Regarding PRA Supporting Requirements F&O ID: Other Affected SRs: Peer Review CC Assessment:

FSS-F2 (Note: not an F&O) CC I Capability Category II/III Requirements:

IF, per SR FSS-F1, one or more scenarios are selected, ESTABLISH and JUSTIFY criteria for structural collapse due to fire exposure.

F&O Issue and Proposed Resolution:

Structural collapse is not deemed likely or addressed further. This meets Capability Category I which does not have any requirements identified. The discussion of structural collapse is qualitative in nature which does not meet the requirements for Capability Category II/III structural collapse analyses.

Basis for Significance:

Capability Category I has no requirements identified, so that SR CC-I is met. Capability Category II/III required more in-depth scenario development, identifying the criteria for structural collapse.

Possible Resolution:

If Capability Category II//III is desired, then more detailed structural analysis is required to be incorporated into the model. However, this may not always be cost effective.

Disposition of the Peer Review Finding:

Resolution of F&O:

The FPRA locations were reviewed and determined to not meet the definition in FSS-F1. Therefore, this SR is N/A.

Evaluation of F&O impact on proposed application:

Capability Category I is sufficient for this application as there are no structural collapse scenarios determined to need a quantitative assessment. Therefore, this finding has no impact on the NSWS 30-day CT T.S. risk insights.

U.S. Nuclear Regulatory Commission PRA Peer Review Findings and Resolutions CNS-17-014 4.5 Fire F&Os Regarding PRA Supporting Requirements F&O ID: Other Affected SRs: Peer Review CC Assessment:

FSS-F3 (Note: not an F&O) CC I Capability Category II/III Requirements:

IF, per SR FSS-F1, one or more scenarios are selected, COMPLETE a quantitative assessment of the risk of the selected fire scenarios in a manner consistent with the FQ requirements, including collapse of the exposed structural steel.

F&O Issue and Proposed Resolution:

No quantitative discussion is provided. A qualitative discussion of structural collapse is provided in Section 3.2 of CNC-1535.00-00-011.

Basis for Significance:

Capability Category I has no requirements identified, so that SR CC-I is met. Capability Category II/III required more in-depth scenario development, identifying the criteria for structural collapse.

Possible Resolution:

If Capability Category II//III is desired, then more detailed structural analysis is required to be incorporated into the model. However, this may not always be cost effective.

Disposition of the Peer Review Finding:

Resolution of F&O:

The FPRA locations were reviewed and determined to not meet the definition in FSS-F1. Therefore, this SR is N/A.

Evaluation of F&O impact on proposed application:

Capability Category I is sufficient for this application as there are no structural collapse scenarios determined to need a quantitative assessment. Therefore, this finding has no impact on the NSWS 30-day CT T.S. risk insights.

U.S. Nuclear Regulatory Commission PRA Peer Review Findings and Resolutions CNS-17-014 4.5 Fire F&Os Regarding PRA Supporting Requirements F&O ID: Other Affected SRs: Peer Review CC Assessment:

FSS-G4-01 (Note: not an F&O) PP-B2 CC I Capability Category II Requirements:

If passive fire barriers with a fire-resistance rating are credited in the Fire PRA, (a) CONFIRM that the allowed credit is consistent with the fire-resistance rating as demonstrated by conformance to applicable test standards (b) ASSESS the effectiveness, reliability, and availability of any passive fire barrier feature credited, and (c) EVALUATE the potential for fire-induced or random failure of credited passive fire barrier features F&O Issue and Proposed Resolution:

Plans indicate that some three-hour boundaries are constructed with two-hour blocks with grout filled cells. No justification for this arrangement and its adequacy was provided. This is also a plant partitioning issue.

Basis for Significance:

Used three-hour fire-rated fire area boundaries and allowed for barrier failure in screening analysis, of the Fire Scenario Report [CNC-1535.00-00-0110].

Possible Resolution:

To achieve Capability Category II, provide original plant construction documents and/or industry test information and building code acceptance information to justify the validity of two-hour block with grout filled cells being equivalent to a three-hour barrier.

Disposition of the Peer Review Finding:

Resolution of F&O:

No further action required; CC-I for this SR is acceptable for the application.

Evaluation of F&O impact on proposed application:

The resolution of this finding has been incorporated into the analysis. Therefore, no further analysis needs to be performed to determine the impact on the NSWS 30-day CT T.S. risk insights.

U.S. Nuclear Regulatory Commission PRA Peer Review Findings and Resolutions CNS-17-014 4.5 Fire F&Os Regarding PRA Supporting Requirements F&O ID: Other Affected SRs: Peer Review CC Assessment:

FSS-H2-01 (Note: not an F&O) CC I Capability Category II/III Requirements:

DOCUMENT a basis for target damage mechanisms and thresholds used in the analysis, including references for any plant-specific performance criteria applied in the analysis.

F&O Issue and Proposed Resolution:

Duke testing was not used. Hughes report was the default report for damage mechanisms resulting in zone of influence damage criteria.

Basis for Significance:

Used zone of influence scoping and documented in Generic Fire Modeling Treatments Report for project 1SPH.02902.030 and CNC-1535.00-00-0110. Thresholds for target damage were based on industry criteria for damage with zone of influence assessment for Catawba. Catawba specific damage criteria were not used.

Possible Resolution:

In order to meet Capability Category II/III classification, the use of Catawba plant-specific damage criteria is required. Determination of plant-specific damage criteria is required with a well document technical basis. Revise and update the Fire PRA as noted above.

Disposition of the Peer Review Finding:

Resolution of F&O:

No further action required; CC-I for this SR is acceptable for the application.

Evaluation of F&O impact on proposed application:

This finding is a documentation issue and has no impact on the NSWS 30-day CT T.S. risk insights.

U.S. Nuclear Regulatory Commission PRA Peer Review Findings and Resolutions CNS-17-014 4.5 Fire F&Os Regarding PRA Supporting Requirements F&O ID: Other Affected SRs: Peer Review CC Assessment:

FSS-H10-01 Met Capability Category I/II/III Requirements:

DOCUMENT the walkdown process and results.

F&O Issue and Proposed Resolution:

Fire Area walkdown notes were input to a computer database, but no output has been created for documentation purposes. In addition, plant drawings identifying the fire areas and the ratings of boundaries to these fire areas have not been found.

Basis for Significance:

For long term coordination and monitoring of the plant configuration relative to FSS, an appendix to report [CNC-1535.00-00-0110] should be created.

Possible Resolution:

Create report from the fire scenario database. Locate fire area configuration drawings or create drawings from existing plant information that can be used to support the FPRA report and to evaluate changes in the future.

Disposition of the Peer Review Finding:

Resolution of F&O:

The fire scenario walkdown information is included in Appendix A of the CNS Fire Scenario Development Report (Reference 27). Additionally, Section 9.1 of the CNS FPRA Plant Partitioning

& Frequency Calculation (Reference 29) discusses the review of drawings to confirm fire area boundaries.

Evaluation of F&O impact on proposed application:

This finding is a documentation issue, which has been addressed. Therefore, there is no impact on the NSWS 30-day CT T.S. risk insights.

U.S. Nuclear Regulatory Commission PRA Peer Review Findings and Resolutions CNS-17-014 4.5 Fire F&Os Regarding PRA Supporting Requirements F&O ID: Other Affected SRs: Peer Review CC Assessment:

HRA-A2-01 Not Met Capability Category I/II/III Requirements:

For each fire scenario, IDENTIFY any new fire-specific safe shutdown actions called out in the plant fire response procedures (e.g., de-energizing equipment per a fire procedure for a specific fire location) in a manner consistent with the scope of selected equipment from the ES and PRM elements of this Standard, and in accordance with HLR-HR-E and its SRs in Part 2 with the following clarifications:

(a) where SR HR-E1 discusses procedures, this is to be extended to procedures for responding to fires (b) where SR HR-E1 mentions in the context of the accident scenarios, specific attention is to be given to the fact that these are fire scenarios (c) another source for SR HR-E1 is likely to be the current Fire Safe Shutdown/Appendix R analysis and DEVELOP a defined basis to support the claim of nonapplicability of any of the requirements under HLR-HR-E in Part 2.

F&O Issue and Proposed Resolution:

This requirement states that HRAs are identified in a manner similar to HLRHR-E from part 2 of the Standard with emphasis on fire scenarios. SR HR-E1 discusses a systematic review of the applicable procedures for operator actions of interest. However, the Fire Modeling documentation does not discuss the review of Plant Fire procedure or other applicable procedures to identify fire specific actions. If this review was performed, then some evidence of the actions considered should be provided.

Basis for Significance:

Most operator actions should already be addressed in the internal events PRA. However, the Plant Fire procedure might indicate new actions or impacts on existing actions that should be addressed.

Possible Resolution:

Perform (or provide more evidence of performance) a review of Plant Fire procedures for impacts to the FPRA.

Disposition of the Peer Review Finding:

Resolution of F&O:

No fire specific fire actions have been added to the FPRA model. No fire operator actions have been identified at this time; the requirement is N/A at this time.

Evaluation of F&O impact on proposed application:

No fire specific operator actions have been developed. Therefore, there is no impact on the NSWS 30-day CT T.S. risk insights.

U.S. Nuclear Regulatory Commission PRA Peer Review Findings and Resolutions CNS-17-014 4.5 Fire F&Os Regarding PRA Supporting Requirements F&O ID: Other Affected SRs: Peer Review CC Assessment:

HRA-A4-01 PRM-B11 CC-I Capability Category I/II/III Requirements:

TALK THROUGH (i.e., review in detail) with plant operations and training personnel the procedures and sequence of events to confirm that interpretation of the procedures relevant to actions identified in SRs HRA-A1, HRA-A2, and HRA-A3 is consistent with plant operational and training practices.

F&O Issue and Proposed Resolution:

Information on operator walk-throughs or talk-throughs for the impact of fires on the operator actions is not presented in CNC-1535.00-00-0111. There is information in the HRA Calculator sheets for the new operator actions developed, but it has no information concerning when these actions were discussed or with whom. This information should be maintained as backup information or included in the applicable document. Also, if the talk-throughs have not been updated since the IPE or IPEEE days, the procedural changes may require updating for the FPRA.

Basis for Significance:

A review of procedural impacts for the fire is required to determine correct impacts on the HEPs due to events such as fire. Talk-throughs will also help verify that any additional actions are not missed.

Possible Resolution:

If talk-throughs were performed for this FPRA, the information should be maintained as backup information or included in the applicable document. If the talk-throughs have not been performed or adequately documented since the IPEEE, then the talk-throughs should be performed and documented in a manner that will help future updates.

Disposition of the Peer Review Finding:

Resolution of F&O:

The FPRA uses a set of multipliers as described in the model development report to account for fire impacts on human reliability. This process is intended to implicitly account for (in a conservative manner) factors influencing operator performance such as fire effects on instrumentation, operator stress, and possible impact on timing. This conservative approach is judged to be consistent with a CC-I approach as indicated in SR HRA-Cl of the standard. With the HRA at CC-I, the FPRA results possess a conservative bias from this aspect of the analysis. With overall risk metric results of the FPRA acceptable, the conservatism does not impede the use of the FPRA for t This came from the F&O response for ESPS. I put the original wording back in.he transition to NFPA 805. No actions have been taken to bring this HRA element to CC-II.

Evaluation of F&O impact on proposed application:

The resolution of this finding has been incorporated into the analysis. Therefore, no further analysis needs to be performed to determine the impact on the NSWS 30-day CT T.S. risk insights.

U.S. Nuclear Regulatory Commission PRA Peer Review Findings and Resolutions CNS-17-014 4.5 Fire F&Os Regarding PRA Supporting Requirements F&O ID: Other Affected SRs: Peer Review CC Assessment:

HRA-B3-01 Not Met Capability Category I/II/III Requirements:

COMPLETE the definition of the HFEs identified in SRs HRA-B1 and HRA-B2 by specifying the following, taking into account the context presented by the fire scenarios in the Fire PRA:

(a) Accident sequence specific timing of cues, and time widow for success completion (b) Accident sequence specific procedural guidance (e.g., AOPs, EOPs)

(c) The availability of cues or other indications for detection and evaluation errors (d) The specific high-level tasks (e.g., train-level required to achieve the goal of the response).

F&O Issue and Proposed Resolution:

The methodology for HRA adjustments does not explicitly address instrumentation, timing and procedural impacts other than simple vs. complex actions, which per HRA-B1-01 were noted as not defined in the documentation.

Basis for Significance:

Addressing the instrumentation, timing and procedural impacts are required to meet any Capability Category for HRA-B3.

Possible Resolution:

Express in the documentation how the instrumentation, timing and procedural impacts are either addressed or bounded by the HRA adjustments provided.

Disposition of the Peer Review Finding:

Resolution of F&O:

In response to CNS NFPA 805 LAR RAI PRA-01b (i), risk significant operator actions credited in the FPRA were re-analyzed using the HRA Calculator. As discussed in Section 4.0 of the FPRA Detailed HRA Report (Reference 30), a Catawba operator was interviewed to verify procedure usage, timing and cues used in the analysis. The results of detailed HRA analysis were included in the response to CNS NFPA 805 LAR RAI PRA-03.

Evaluation of F&O impact on proposed application:

The resolution of this finding has been incorporated into the analysis. Therefore, no further analysis needs to be performed to determine the impact on the NSWS 30-day CT T.S. risk insights.

U.S. Nuclear Regulatory Commission PRA Peer Review Findings and Resolutions CNS-17-014 4.5 Fire F&Os Regarding PRA Supporting Requirements F&O ID: Other Affected SRs: Peer Review CC Assessment:

HRA-C1-02 CC-I Capability Category II Requirements:

For each selected fire scenario, QUANTIFY the HEPs for all HFEs and ACCOUNT FOR relevant fire-related effects using detailed analyses for significant HFEs and conservative estimates (e.g.,

screening values) for non-significant HFEs, in accordance with the SRs for HLR-HR-G in Part 2 set forth under at least Capability Category II, with the following clarification:

(a) Attention is to be given to how the fire situation alters any previous assessments in nonfire analyses as to the influencing factors and the timing considerations covered in SRs HR-G3, HR-G4, and HR-G5 in Part 2 and (b) DEVELOP a defined basis to support the claim of nonapplicability of any of the requirements under HLR-HR-G in Part 2.

F&O Issue and Proposed Resolution:

A finding from the FPIE evaluation stated that HEPs are not converted from medians to means.

This issue was said to be addressed with a sensitivity case. However, this issue should be addressed in the Fire PRA.

Basis for Significance:

This finding will have a minor impact on post-accident HEP, but will cause a 2-3 times increase in pre-accident HEPs.

Possible Resolution:

Ensure that the HEPs are completely based on means rather than medians.

Disposition of the Peer Review Finding:

Resolution of F&O:

The HEP values have been converted from median to mean in the FPRA model.

Evaluation of F&O impact on proposed application:

The results of this finding have been incorporated into the analysis. Therefore, there is no further analysis that needs to be performed to support the NSWS 30-day CT T.S. LAR.

U.S. Nuclear Regulatory Commission PRA Peer Review Findings and Resolutions CNS-17-014 4.5 Fire F&Os Regarding PRA Supporting Requirements F&O ID: Other Affected SRs: Peer Review CC Assessment:

HRA-D2-01 Not Met Capability Category I/II/III Requirements:

For any operator recovery actions identified in HRA-D1:

(a) ACCOUNT FOR relevant fire-related effects, including any effects that may preclude a recovery action or alter the manner in which it is accomplished, in accordance with HR-H2 and HR-H3 in Part 2; and (b) DEVELOP a defined basis to support the claim of nonapplicability of any of the requirements under HR-H2 and HR-H3 in Part 2.

F&O Issue and Proposed Resolution:

The one recovery action developed for the Fire PRA (TSSPZRLRHE) is not proceduralized nor is it trained on. There is no discussion of why this action can be credited, which is contrary to the requirements of HR-H2 so this SR is Not Met.

Basis for Significance:

Per HR-H2, not able to credit an operator action that is not proceduralized nor trained on.

Possible Resolution:

Provide basis for crediting this action despite lack of procedures or training.

Disposition of the Peer Review Finding:

Resolution of F&O:

The operator action referenced in the finding is not a "fire recovery" in the context of NFPA 805.

This is an action added to the FPRA model to address a specific accident sequence (not fire specific) that was not yet included in the internal events model. Section 5.1 of the Model Development Report has been updated to better describe the basis for crediting this operator action.

Evaluation of F&O impact on proposed application:

This finding has no impact on the risk metrics used for the NSWS 30-day CT T.S. LAR.

U.S. Nuclear Regulatory Commission PRA Peer Review Findings and Resolutions CNS-17-014 4.5 Fire F&Os Regarding PRA Supporting Requirements F&O ID: Other Affected SRs: Peer Review CC Assessment:

SF-A3-01 (Note: not a finding) Not Met Capability Category I/II/III Requirements:

ASSESS the potential for common-cause failure of multiple fire suppression systems due to the seismically induced failure of supporting systems such as fire pumps, fire water storage tanks, yard mains, gaseous suppression storage tanks, or building standpipes.

F&O Issue and Proposed Resolution:

There is no indication in the documentation that both the seismic-induced fire as well as seismic-induced failure of fire mitigation systems has been considered as required by the SR.

Basis for Significance:

Even though the SR is not met, this does not affect the Fire PRA results.

Possible Resolution:

Provide the necessary documentation.

Disposition of the Peer Review Finding:

Resolution of F&O:

Section 3.13 of the FPRA Summary Report (Reference 22) has been updated to include a discussion of the impact of a seismic and seismically-induced fire events. No impact on quantification.

Evaluation of F&O impact on proposed application:

This finding is a documentation issue, which has been resolved. Therefore, there is no impact on the NSWS 30-day CT T.S. risk insights.

U.S. Nuclear Regulatory Commission PRA Peer Review Findings and Resolutions CNS-17-014 4.5 Fire F&Os Regarding PRA Supporting Requirements F&O ID: Other Affected SRs: Peer Review CC Assessment:

SF-A5-01 Not Met Capability Category I/II/III Requirements:

REVIEW (a) plant fire brigade training procedures and ASSESS the extent to which training has prepared firefighting personnel to respond to potential fire alarms and fires in the wake of an earthquake and (b) the storage and placement of firefighting support equipment and fire brigade access routes, and (c) ASSESS the potential that an earthquake might compromise one or more of these features F&O Issue and Proposed Resolution:

This SR basically requires that CNS qualitatively assess their existing fire brigade training procedures to determine if the training has prepared the brigade to respond to fire alarms after an earthquake, to review their staging of fire mitigation equipment and to assess whether or not the occurrence of a seismically-induced fire and any associated damage might compromise either of these elements. The CNS seismic/fire interaction evaluation is discussed in Section 3.13 of CNC-1535.00-00-0112. In general, CNS relies upon the assessments performed for the IPEEE analyses, in particular, the walkdowns. The IPEEE walkdown is documented in CNC-1435.00-00-0007 and the overall IPEEE is documented in the IPEEE Submittal Report. A review of these documents does not show any evaluation of a seismically-induced fire and the potential impacts on brigade response and equipment staging.

Basis for Significance:

Direct violation of the SR requirements.

Possible Resolution:

Take steps to conform to the SR.

Disposition of the Peer Review Finding:

Resolution of F&O:

Section 3.13 of the FPRA Summary Report (Reference 22) has been updated to include an evaluation of seismically induced fire and the potential impacts on fire brigade response and equipment staging. No impact on quantification.

Evaluation of F&O impact on proposed application:

The documentation has been updated to address this finding. Quantification results were not impacted. Therefore, this finding has no impact on the NSWS 30-day CT T.S. risk insights.

U.S. Nuclear Regulatory Commission PRA Peer Review Findings and Resolutions CNS-17-014 4.5 Fire F&Os Regarding PRA Supporting Requirements F&O ID: Other Affected SRs: Peer Review CC Assessment:

FQ-A2-01 PRM-B5 Met Capability Category I/II/III Requirements:

For each fire scenario selected per the FSS requirements that will be quantified as a contributor to fire-induced plant CDF and/or LERF, IDENTIFY the specific initiating event or events (e.g., general transient, LOOP) that will be used to quantify CDF and LERF.

F&O Issue and Proposed Resolution:

Loss of Offsite Power (LOOP) events are not adequately represented in the Fire PRA model.

Scenarios resulting in a LOOP are modeled by setting %T1 to TRUE along with the basic events for 6900V Switchgear 1TA/1TD and transformers 1ATC/1ATD. However, this does not satisfy all the LOOP logic, such as the PORV and SRV response following a LOOP, impact on Instrument Air and the ability to recover Main Feedwater.

Basis for Significance:

Does not capture all the effects of a LOOP and may underestimate the CDF/LERF for fire scenarios involving LOOP.

Possible Resolution:

Make change to FRANC database to set additional basic event(s) (e.g., PACBOFTDEX) to TRUE.

Disposition of the Peer Review Finding:

Resolution of F&O:

The FPRA model has been updated to collect offsite power cables under 1/2SYS-OSP which have been linked to basic event PACBOFTDEX under new gate TQ76A to address LOOP logic. This assures that the LOOP affects are reflected in the PORV, IA, MFW and SRV logic structures.

Evaluation of F&O impact on proposed application:

The analysis has been updated to address this finding and is included in the quantification results for the NSWS 30-day CT T.S. LAR.

U.S. Nuclear Regulatory Commission PRA Peer Review Findings and Resolutions CNS-17-014 4.5 Fire F&Os Regarding PRA Supporting Requirements F&O ID: Other Affected SRs: Peer Review CC Assessment:

FQ-F1-01 FQ-E1 Not Met Capability Category I/II/III Requirements:

DOCUMENT the CDF and LERF analyses in accordance with HLR-QU-F and HLR-LE-G and their SRs in Part 2 with the following clarifications:

(a) SRs QU-F2 and QU-F3 of Part 2 are to be met including identification of which fire scenarios and which physical analysis units (consistent with the level of resolution of the Fire PRA such as fire area or fire compartment) are significant contributors (b) SR QU-F4 of Part 2 is to be met consistently with 4-2.13 (c) SRs LE-G2 (uncertainty discussion) and LE-G4 of Part 2 are to be met consistently with 4-2.13 and DEVELOP a defined basis to support the claim of nonapplicability of any of the requirements under these sections in Part 2.

F&O Issue and Proposed Resolution:

There are asymmetries in the model results between trains A and B; this is due to the assumption that A train components are normally running and B train components are in standby (and thus all maintenance is assigned to that train). This results in asymmetrical results and is not discussed in the document.

Basis for Significance:

This is a significant shortcoming/impact on the results that needs to be documented.

Possible Resolution:

At a minimum, add a discussion to the document regarding the asymmetry. It is recommended that the model be modified to remove asymmetries Disposition of the Peer Review Finding:

Resolution of F&O:

A discussion of the model asymmetries and the potential impact on the FPRA results has been added to the FPRA Model Development Report (Reference 26). Additionally, a comparison of risk results and importance measures for A-train versus B-train fire areas and selected equipment demonstrated the impact from model asymmetry to be insignificant with respect to FRE conclusions based on RG 1.174 acceptance thresholds.

Evaluation of F&O impact on proposed application:

This finding is a documentation issue and has been resolved. Therefore, there is no impact on the NSWS 30-day CT T.S. risk insights.

U.S. Nuclear Regulatory Commission PRA Peer Review Findings and Resolutions CNS-17-014 4.5 Fire F&Os Regarding PRA Supporting Requirements F&O ID: Other Affected SRs: Peer Review CC Assessment:

FQ-F1-02 Not Met Capability Category I/II/III Requirements:

DOCUMENT the CDF and LERF analyses in accordance with HLR-QU-F and HLR-LE-G and their SRs in Part 2 with the following clarifications:

(a) SRs QU-F2 and QU-F3 of Part 2 are to be met including identification of which fire scenarios and which physical analysis units (consistent with the level of resolution of the Fire PRA such as fire area or fire compartment) are significant contributors (b) SR QU-F4 of Part 2 is to be met consistently with 4-2.13 (c) SRs LE-G2 (uncertainty discussion) and LE-G4 of Part 2 are to be met consistently with 4-2.13 and DEVELOP a defined basis to support the claim of nonapplicability of any of the requirements under these sections in Part 2.

F&O Issue and Proposed Resolution:

Many specific details from HLR-QU-F and HLR-LE-G are not documented. Specifically:

a) QU-F2: Review process, identification of key equipment and operator actions, bases for mutually exclusive events, and the process used to illustrate computer code correctness.

b) QU-F5 and LE-G5: Limitations in the quantification process that would impact applications.

c) QU-F6 and LE-G6: Quantitative definition for significant.

d) LE-G2: Containment failure analysis and failure probability estimate for containment implosion due to spurious NS or VX activation.

Basis for Significance:

A substantial number of document discrepancies are noted.

Possible Resolution:

Incorporate additional document items into the Summary Report.

Disposition of the Peer Review Finding:

Resolution of F&O:

Appendix E has been added to the CNS Fire PRA Summary Report (Reference 22) to include an importance measure report from the integrated cutset results to address QU-F2 (the key equipment/actions). Section 3.1 and 3.2 of the CNS FPRA Model Development Report (Reference

26) includes a discussion to address LE-G2 (Spurious NS, VX). Section 4.0 of the Model Development Report addresses the quantitative definition of significant, QU-F6 and LE-G6. Section 2.2 of the Model Development Report have been updated to address QU-F5 and LE-G5 (computer code correctness and limitations). Section 5.2 of the Model Development Report was updated to provide the basis for mutually exclusive event recovery rules.

Evaluation of F&O impact on proposed application:

This finding is a documentation issue, which has been resolved. Therefore, there is no impact on the 30-day CT T.S. risk insights.

U.S. Nuclear Regulatory Commission Attachment 4 PRA Peer Review Findings and Resolutions CNS-17-014 References

1. CNC-1535.00-00-0220, Catawba Nuclear Station Probabilistic Risk Assessment Resolution of Peer Review Facts and Observation, Revision 1

2. CNC-1535.00-00-0217, Catawba Nuclear Station Post-Initiator Human Reliability Analysis, Rev. 1

3. CNC-1535.00-00-0200, Catawba Nuclear Station PRA Peer Review F&O Resolutions, Rev. 2

4. CNC-1535.00-00-0205, Catawba Nuclear Station Pre-Initiator Human Reliability Analysis, Revision 0

5. CNC -1535.00-00-0180, Catawba Nuclear Station Probabilistic Risk Assessment Section 8.0:

Model Integration and Quantification, Revision 1

6. CNC-1535.00-00-0173, Catawba Nuclear Station Probabilistic Risk Assessment Initiating Events Development Notebook, Revision 1

7. Summitt, R., Catawba Nuclear Station Plant Systems Engineering and Plant Operations Interviews, Rev. 0, RSC Engineers, RSC 16-06/CNC-1535.00-00-0176.26, November 2016

8. CNC-1535.00-00-0174, Catawba Nuclear Station Success Criteria Notebook, Revision 1

9. CNC-1535.00-00-0175, CNS Accident Sequence Development Notebook, Revision 1

10. CNC -1535.00-00-0182, Catawba Nuclear Station Probabilistic Risk Assessment Section 1.0:

Component Data Development, Revision 1

11. Reagan, A., Catawba Nuclear Station Probabilistic Risk Assessment Section 5.0, Appendix 10:

Containment Spray System Notebook (NS), Rev. 1, RSC Engineers, Inc., RSC 14-42/CNC-1535.00-00-176.10, November 2016

12. Dudley, D, et al., Catawba Nuclear Station Probabilistic Risk Assessment System Walkdown Documentation, Rev. 0, RSC Engineers, Inc., RSC 14-26/CNC-1535.00-00-0181, December 2014

13. Summitt, R., and D. Croft, Catawba Nuclear Station Probabilistic Risk Assessment Section 5.0, Appendix 0: Modeling Groundrules and Assumptions, Rev. 1, RSC Engineers, Inc., RSC 14-19/CNC-1535.00-00-176, December 2016

14. CNC-1535.00-00-0201, CNS Loss of Offsite Power (LOOP) Restoration Assessment, Revision 3.

15. CNC-1535.00-00-0176.04, Catawba Nuclear Station Probabilistic Risk Assessment, Appendix 4:

CNS Auxiliary Feedwater System Notebook (CA), Revision 1

16. Morris, S., and J. Jansen Vehec, Catawba Nuclear Station Probabilistic Risk Assessment Section 5.0, Appendix 11: AC Power System Notebook (ACP), Rev. 1, RSC Engineers, Inc., RSC 14-43/CNC-1535.00-00-176.11, October 2016

17. Reagan, A., et al., Catawba Nuclear Station Probabilistic Risk Assessment Section 8.0: Model Integration and Quantification, Rev. 1, RSC Engineers, Inc., RSC 14-31/CNC-1535.00-00-180, December 2016

18. Duke Calc. CNC-1223.24-00-0072; RN Single Pond Return Header Design Basis; Rev. 1

U.S. Nuclear Regulatory Commission Attachment 4 PRA Peer Review Findings and Resolutions CNS-17-014 References

19. CNC-1535.00-00-0176.01, Error! Reference source not found.CNS Residual Heat Removal System Notebook (ND), Revision 1

20. WCAP-16304-P, Strategy for Identifying and Treating Modeling Uncertainties in PRA Models:

Issues Concerning LOCA and LOOP, Revision 0

21. CNC-1535.00-00-0154, Catawba Nuclear Station Units 1 and 2 High Wind/Missile PRA Analysis, Revision 2

22. CNC-1535.00-00-0112, CNS Fire PRA (FPRA) Summary Report, Revision 2

23. CNC-1381.05-00-0251, Units 1 and 2 NFPA 805 Circuit Breaker and Fuse Coordination Study, Revision 16

24. CNC-1535.00-00-0109, CNS Fire PRA Cable Selection Calculation, Revision 1

25. CNC-1535.00-00-0108, CNS Fire PRA Component Selection, Revision 1

26. CNC-1535.00-00-0111, CNS Fire PRA (FPRA) Model Development Report, Revision 2

27. CNC-1535.00-00-0110, CNS Fire PRA (FPRA) Scenario Development Report, Revision 2

28. DPC-1535.00-00-0024, Generic Fire Modeling Treatments, Revision 0

29. CNC-1535.00-00-0107, CNS Fire PRA Ignition Source Frequency Calculation, Revision 2

30. CNC -1535.00-00-0189, Catawba Fire PRA: HRA Detailed Analysis, Revision 0

31. CNC-1535.00-00-0113, CNC FPRA (Fire PRA) Application Calculation, Revision 2

32. PWROG-16032-P, Peer Review of the Catawba Nuclear Station Internal Events Probabilistic Risk Assessment, Revision 0, April 2016 (Proprietary to Duke Energy Corporation)

33. Catawba Nuclear Plant RG 1.200 High Wind PRA Peer Review Report, Attachment to LTR-RAM-II-13-077, Westinghouse Electric Company, January 2014 (Proprietary to Duke Energy Corporation)

34. ASME/ANS RA-Sb-2013, Addenda to ASME/ANS RA-S-2008 Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, American Society of Mechanical Engineers, New York, NY, September 2013.

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

36. Process for Performing Internal Events PRA Peer Reviews Using the ASME/ANS PRA Standard, NEI 05-04, Revision 2, Nuclear Energy Institute, November 2008. [ML083430462]

v t e Letter Sequence Request
CAC:MG0245, Control Room Habitability (Approved, Closed)
Initiation Request Acceptance Supplement, Supplement, Supplement Administration Questions 9 March 2018 Answers 8 May 2018 17 August 2018 Results Approval Other: ML18117A187
MONTHYEAR2018-11-28 [Table View]

CNS-17-014, License Amendment Request to Revise Technical Specification Section 3.7.8, Nuclear Service Water System, to Add a New Condition to Allow Single Pond Return Header Operation of the Nsws with a 30-Day Completion Time

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CNS-17-014, License Amendment Request to Revise Technical Specification Section 3.7.8, Nuclear Service Water System, to Add a New Condition to Allow Single Pond Return Header Operation of the Nsws with a 30-Day Completion Time

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