Text

Proposed Technical Specification Amendment, Technical Specification 3.6.6, Containment Spray System; 3.7.5, Auxiliary Feedwater System
	ML081980198
Person / Time
Site:	Catawba
Issue date:	07/14/2008
From:	Morris J; Duke Energy Carolinas
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	Download: ML081980198 (60)
	v • d • e

Duke JAMES R. MORRIS, VICE PRESIDENT Energy Duke Energy Carolinas, LLC Carolinas Catawba Nuclear Station 4800 Concord Road! CN01 VP York, SC 29745 803-701-4251 803-701-3221 fax July 14, 2008 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20555

Subject:

Duke Energy Carolinas, LLC Catawba Nuclear Station, Unit I Docket Number 50-413 Proposed Technical Specification Amendment Technical Specification 3.6.6, Containment Spray System; 3.7.5, Auxiliary Feedwater System Pursuant to 10 CFR 50.4, 10 CFR 50.90, and 10 CFR 50.91(a)(5), the licensee for Catawba Nuclear Station proposes a one-time limited duration extension of the Technical Specifications (TS) 3.6.6, Containment Spray System (CSS); and TS 3.7.5, Auxiliary Feedwater (AFW) System for Unit I. These extensions are required to facilitate repair and replacement of the 1B NSWS pump and the activities associated with the repair.

On July 12, 2008 at approximately 1041 hours0.012 days 0.289 hours 0.00172 weeks 3.961005e-4 months the lB NSWS pump was declared inoperable and the 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months action statement of TS 3.7.8 was entered. On Sunday July 13, 2008, at approximately 1115 Operations realigned NSWS on Unit 1 utilizing operating procedure, OP/O/A/6400/006 C, Nuclear Service Water System, Enclosure 4.12B. This enclosure isolates NSWS flow to the 1B AFW pump, I B CSS heat exchanger and the Unit I nonessential NSWS header. This allowed the "B" NSWS train to be declared operable and both units exited TS 3.7.8.

This required Unit I to enter TS 3.6.6 Required Action A and TS 3.7.5 Required Action B, each with a 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Completion Time. These TS Required Actions will remain in effect until the repairs and restoration of the 1B NSWS pump are complete. At that time, NSWS will be realigned to these components and the applicable Required Actions exited.

Although efforts are underway to replace the lB NSWS pump will not be restored to operable status prior to expiration of the completion time. In order to avoid the shutdown of Catawba Unit 1, Duke proposes a one-time limited duration extension of the Technical Specification Required Action Completion Time associated with the Unit IB AFW pump and the IB CSS. The requested extension would allow continued operation of Unit 1 for an additional 144 hours0.00167 days 0.04 hours 2.380952e-4 weeks 5.4792e-5 months while repairs and related testing of the I B NSWS pump are completed.

www. duke-energy. comr

U.S. Nuclear Regulatory Commission July 14, 2008 Page 2 of 5 Both units are currently at 100% power. Completion Times of the applicable Required Actions expire on July 15, 2008 at 1041 hours0.012 days 0.289 hours 0.00172 weeks 3.961005e-4 months . An estimated repair time for the lB NSWS pump is nine (9) days and thus this will exceed the 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months allowed by the TS. Therefore, in order to avoid the shutdown of Catawba Unit 1, Duke requests approval of this license amendment on a one-time emergency basis by July 15, 2008 at 0800 hours0.00926 days 0.222 hours 0.00132 weeks 3.044e-4 months . provides a description of the proposed change and technical justification, an evaluation of significant hazards consideration pursuit to 10 CFR 50.92 (c) and an environmental assessment. provides the existing TS pages marked-up to show the proposed change. contains retyped (clean) TS pages. lists the regulatory commitments documented in this request. contains a Catawba PRA quality discussion.-

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

Implementation of this amendment request will not require changes to the Catawba Updated Final Safety Analysis Report (UFSAR).

Pursuant to 10 CFR 50.91, a copy of this proposed amendment is being sent to the appropriate State of South Carolina official.

Should you have any questions concerning this information, please call A. P. Jackson at (803) 701-3742.

Very truly yours, James R. Morris Attachments

U.S. Nuclear Regulatory Commission July 14, 2008 Page 3 of 5 James R. Morris affirms that he is the person who subscribed his name to the foregoing statement, and that all the matters and facts set forth herein are true and correct to the best of his knowledge.

James R. Morris, Site Vice President Subscribed and sworn to me:

QtAI V 4)ate 't My commission expires:

t-;ý11 20 Date SEAl

)

ON

U.S. Nuclear Regulatory Commission July 14, 2008 Page 4 of 5 xc (with attachments):

Luis Reyes, Region II Administrator U.S. Nuclear Regulatory Commission Sam Nunn Atlanta Federal Center, 23 T85 61 'Forsyth St., SW Atlanta, GA 30303-8931 J. F. Stang, Jr., Senior Project Manager (CNS & MNS)

U.. S. Nuclear Regulatory Commission 11555 Rockville Pike Mail Stop 8 G9A Rockville, MD 20852-2738 A. T. Sabisch Senior Resident Inspector U. S. Nuclear Regulatory Commission Catawba Nuclear Station S. E. Jenkins, Manager Division of Radioactive Waste Management Bureau of Land and Waste Management Department of Health and Environmental Control 2600 Bull Street Columbia, SC 29201

U.S. Nuclear Regulatory Commission July 14, 2008 Page 5 of 5 bxc (with attachments):

J. R. Morris (CN01VP)

J. W. Pitesa (CNO1SM)

T. M. Hamilton (CNSOISA)

R. D. Hart (CNO1RC)

R. L. Gill (EC05P)

A. P. Jackson (CNO1RC)

K. L. Ashe (MGO1RC)

NCMPA-1 NCEMC PMPA SREC Document Control File 801.01 RGC File ELL-EC050

ATTACHMENT 1 DESCRIPTION OF PROPOSED CHANGES, TECHNICAL JUSTIFICATION, SIGNIFICANT HAZARDS CONSIDERATION PURSUANT TO 10CFR50.92(C) AND ENVIRONMENTAL ASSESSMENT

1.

==

Description:==

Pursuant to 10 CFR 50.90 and 10 CFR 50.91 (a) (5), Duke Energy Carolinas, LLC (Duke), the licensee for Catawba Nuclear Station, proposes a one-time limited duration extension of the Technical Specification (TS) 3.7.5 Required Action B. 1 Completion Time associated with the lB Auxiliary Feedwater (AFW) pump, and TS 3.6.6 Required Action A.l Completion Time associated with the lB Containment Spray System (CSS).

The requested extension would allow continued operation of Unit 1 for an additional 144 hours0.00167 days 0.04 hours 2.380952e-4 weeks 5.4792e-5 months above the 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months action statement time while repairs and related testing of the lB nuclear service water system (NSWS) pump are completed.

The proposed amendment is being requested on an emergency basis pursuant to 10 CFR 50.91 (a) (5). On July 12, 2008 at approximately 1041 hours0.012 days 0.289 hours 0.00172 weeks 3.961005e-4 months the lB NSWS pump was declared inoperable and the 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months action statement of TS 3.7.8 was entered. On Sunday July 13, 2008, at approximately 1115 Operations realigned NSWS on Unit 1 utilizing operating procedure, OP/0/A/6400/006 C, Nuclear Service Water System,.12B. This enclosure isolates NSWS flow to the 1B AFW pump, lB CSS heat exchanger and the. Unit I nonessential NSWS header. This allowed the "B" NSWS train to be declared operable and both units exited TS 3.7.8. This required Unit 1 to enter TS 3.6.6 Required Action A and TS 3.7.5 Required Action B each with a 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Completion Time. These TS Required Actions will remain in effect until the repairs and restoration of the 1B NSWS pump are complete. At that time, NSWS will be realigned to these components and the applicable Required Actions exited.

Efforts are currently in progress to replace the lB NSWS pump; however, the repairs will not be completed prior to expiration of the current completion time at 1041 on July 15, 2008. Therefore, in order to avoid the shutdown of Catawba Unit 1, Duke requests approval of this license amendment application on a one-time emergency basis by July 15, 2008 at 0800.

2.

Proposed Change:

The proposed change would add two new License Conditions to Appendix B of the Catawba Nuclear Station Unit I Facility Operating License, License Number NPF-35.

The proposed License Conditions are as follows The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months allowed outage time of Technical Specification 3.7.5 Action "B" for the lB AFW pump which was entered at 1041 on July 12, 2008 may be extended by an additional 144 hours0.00167 days 0.04 hours 2.380952e-4 weeks 5.4792e-5 months . Upon completion of the repair and restoration of the 1B NSWS pump, this License Condition is no longer applicable and will expire at 1041 on July 21, 2008.

The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months allowed outage time of Technical Specification 3.6.6 Action "A" for the IB CSS which was entered at 1041 on July 12, 2008 may be extended by an additional 144 hours0.00167 days 0.04 hours 2.380952e-4 weeks 5.4792e-5 months . Upon completion of the repair and restoration of the lB NSWS pump, this License Condition is no longer applicable and will expire at 1041 on July 21, 2008.

Attachment I Page 1 of 20

3.

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 non-safety related components.

The NSWS consists of two independent loops (A and B) of essential equipment, each of which is shared between units. Each loopcontains two NSWS pumps, each of which is supplied from a separate emergency diesel generator. Each set of two pumps supplies two trains (IA and.2A, or lB and 2B) of essential equipment through common discharge piping. While the pumps are unit designated, i.e., IA, IB, 2A, 2B, albpumps 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 IA NSWS pump, powered by emergency diesel IA, will supply post accident cooling to NSWS trains IA and 2A.

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 flow path 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:

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.

The NSWS has another safety related function with regard to the AFW system. The condensate storage system supplies the AFW system suction source requirements during normal system operating modes; but, since the condensate storage system is not safety related its availability is not assured. The assured source 6f water supply to the AFW pumps is provided by the safety related portion of the Nuclear Service Water System.

Attachment I Page 2 of 20

Another safety related function of the NSWS is to supply cooling water to the CSS heat exchangers during the recirculation phase of a loss of coolant accident. In the recirculation mode of operation, containment spray pump suction is transferred from the refueling water storage tank (RWST) to the containment recirculation sump(s). When the containment spray system suction is from the containment recirculation sump, its associated heat exchanger receives NSWS flow for cooling.

On July 12, 2008 at approximately 0910 the control room operators started the lB NSWS pump and stopped the IA NSWS pump in support of a NSWS train B supply header flush. At approximately 1041 the control room operators received several NSWS alarms for B NSWS train header pressure and flow. In addition, the operators observed low discharge header pressure along with high NSWS flow for the 2B NSWS pump and low flow for the 1B NSWS pump. The operators entered their abnormal procedure for the NSWS and started the IA NSWS pump and stopped the lB NSWS pump. The lB NSWS pump was declared inoperable as of 1041 and both units entered TS 3.7.8 Action A with a 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months completion time.

On Sunday July 13, 2008, at approximately 1115 Operations realigned NSWS on Unit I utilizing operating procedure, OP/O/A/6400/006 C, Nuclear Service Water System,.12B. This enclosure isolates NSWS flow to the lB AFW pump, 1B CSS heat exchanger and the Unit I nonessential header. This allowed the "B"INSWS train to be declared operable and both units exited TS 3.7.8. This required Unit 1 to enter TS 3.6.6 Required Action A and TS 3.7.5 Required Action B each with a 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Completion Time.. These alignments allow the 2B NSWS pump to carry the loads for Unit I Train B except for the isolated sections discussed above. Since the required flows are not available for the lB AFW pump and the lB CSS heat exchangers, the start time of the LCO reverts back to the time the NSWS was taken out of service originally. These TS Required Actions will remain in effect until the repairs and restoration of the lB NSWS pump are complete. At that time, NSWS will be realigned to these components and the applicable Required Actions exited.

Completion times for the applicable TS Required Actions for the I B AFW pump and lB CSS train expire at 1041 on July 15, 2008. Although efforts are underway to replace the lB NSWS pump, the pump will not be restored to operable status prior to expiration of the completion time.

In order to avoid the shutdown of Catawba Unit 1, Duke proposes a one-time limited duration extension of the Technical Specification Required Action Completion Time associatedwith the Unit LB AFW pump and the lB CSS. The requested extension would allowcontinued operation of Unit 1 for an additional 144 hours0.00167 days 0.04 hours 2.380952e-4 weeks 5.4792e-5 months while repairs and related testing of the IB NSWS pump are completed.

4.

Current Requirements:

TS 3.7.5, "Auxiliary Feedwater (AFW) System" contains LCO 3.7.5. This LCO governs the AFW system for Modes 1, 2, 3, and 4 when steam generator is relied upon for heat Attachment I Page 3 of 20

removal. LCO 3.7.5 requires three AFW trains to be operable. Condition B for this LCO states that with one AFW train inoperable, the inoperable train must be restored to operable status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . Condition C states that if the required action and associated completion time is not met, the unit must-be in Mode 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in Mode 4 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

TS 3.6.6, "Containment Spray System'" contains LCO 3.6.6. This LCO governs the CSS for Modes 1, 2, 3, and 4. LCO 3.6.6 requires two CSS trains to be operable. Condition A for this LCO states that with one CSS train inoperable, the inoperable train must be restored to operable status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . Condition B states that if the required action

and associated completion time is not met, the unit must be in 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 .

5.

Basis for Current Requirements:

LCO 3.7.5 Basis Discussion The AFW System mitigates the consequences of any event with loss of normal feedwater.

The design basis of the AFW System is to supply water to the steam generator to remove decay heat and other residual heat by delivering at least the minimum required flow rate to the steam generators at pressures corresponding to the lowest steam generator safety

,valve set pressure-plus 3%. In addition, the AFW System must supply enough makeup water to replace steam generator secondary inventory lost as the unit cools to MODE 4 conditions. Sufficient AFW flow must also be available to account for flow losses such as pump recirculation valve leakage and line breaks.

The limiting Design Basis Accidents (DBAs) and transients for the AFW System are as follows:

a. Feedwater Line Break (FWLB);

b. Loss of Main Feedwater (MFW)

In addition, the minimum, available AFW flow and system characteristics are considered in the analysis of a small break loss of coolant accident (LOCA) and events that could lead to steam generator tube, bundle uncovery for dose considerations.

The AFW System satisfies the requirements of Criterion 3 of 10 CFR 50.36.

LCO 3.6.6 Basis Discussion The limiting DBAs considered relative to containment OPERABILITY are the loss of coolant accident (LOCA) and the steam line break (SLB). The DBA LOCA and SLB are analyzed using computer codes designed to predictthe resultant containment pressure and temperature transients. No two DBAs are assumed to occur simultaneously or consecutively. The postulated DBAs are analyzed, in regard to'containment engineered Page 4 of 20

safety feature (ESF) systems, assuming the loss of one ESF bus, which is the worst case single active failure, resulting in one train of the Containment Spray System, the RHR System, and the Air Return System (ARS) being rendered inoperable.

The DBA analyses show that the maximum peak containment pressure results from the LOCA analysis, and is calculated to be less than the containment design pressure. The maximum peak containment atmosphere temperature results from the SLB analysis and was calculated to be within the containment environmental qualification temperature during the DBA SLB. The basis of the containment environmental qualification temperature is to ensure the OPERABILITY of safety related equipment inside containment.

The Containment Spray System satisfies Criterion 3 of 10 CFR 50.36.

6.

Reason for Requesting Emergency Amendment:

Regulation 10 CFR 50.91(a) (5) states that where the NRC finds that an emergency situation exists, in that failure to act in a timely way would result in derating or shutdown of a nuclear power plant, or in prevention or either resumption of operation or increase in power output up to the plant's licensed power level, it may issue a license amendment involving no'significant hazards consideration without prior notice and opportunity for a hearing or for public comment. The regulation also states that the NRC will decline to dispense with notice and comment on the no significant hazards if it determines that the licensee has abused the emergency provision by failing to make timely application for the amendment and thus itself creating the emergency. The regulation requires that a licensee requesting an emergency amendment explain why the emergency situation occurred and why the licensee could not avoid the situation. As explained below, an emergency amendment is needed to preclude a plant shutdown and cooldown, and Duke could not have reasonably avoided the situation or made timely application for an amendment.

7.

Reason the Emergency Situation Has Occurred:

On July 12, 2008 at approximately 0910 the control room operators started the IB NSWS pump and stopped the IA NSWS pump in support of NSWS train B supply header flushes. At approximately 1041 the control room operators received several NSWS alarms for B NSWS train header pressure and flow. The operators observed low discharge header pressure along with high NSWS flow for the 2B NSWS pump and low flow for the lB NSWS pump. The operators entered their abnormal procedure for the NSWS and started the IA NSWS pump and stopped the lB NSWS pump. The lB NSWS pump was declared inoperable as of 1041 and both units entered TS 3.7.8 Action A with a 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months completion time.

On Sunday July 13, 2008, at approximately 1115 Operations realigned NSWS on Unit I utilizing operating procedure, OP/0/A/6400/006 C, Nuclear Service Water System,.12B. This enclosure isolates NSWS flow to the lB AFW pump, lB CSS heat exchanger and the Unit 1 nonessential header. This allowed the "B" NSWS train to Attachment I Page 5 of 20

be declared operable and both units exited TS 3.7.8. This required Unit 1 to enter TS 3.6.6 Required Action A and TS 3.7.5 Required Action B each with a 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Completion Time. These TS Required Actions will remain in effect until the repairs and restoration of the I B NSWS pump are complete. At that time, NSWS will be realigned to these components and the applicable Required Actions exited.

Completion times for the applicable TS Required Actions for the lB AFW pump and lB CSS train expire at 1041 on July 15, 2008. Although efforts are underway to repair the lB NSWS pump and it will not be restored to operable status prior to expiration of the completion time.

In order to avoid the shutdown of Catawba Unit 1, Duke proposes a one-time limited duration extension of the Technical Specification Required Action Completion Time associated with the Unit lB AFW pump and the lB CSS. The requested extension would allow continued operation of Unit 1 for an additional 144 hours0.00167 days 0.04 hours 2.380952e-4 weeks 5.4792e-5 months while repairs and related testing of the lB NSWS pump are completed.

Both units are currently at 100% power. An estimated repair time for the lB NSWP is nine (9) days and thus this will exceed the 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months allowed by the TS. Therefore, in order to avoid the shutdown of Catawba Unit 1, Duke requests approval of this license amendment on a one-time emergency basis by July 15, 2008 at0800 hours.

8.

Reason the Situation Could Not Have Been Avoided:

Initial Incident Investigation At 0030, on July 13, 2008, a diver crew and maintenance pump team performed an inspection of the pump. The diver crew entered the suction pit and discovered several metallic pieces lying on the bottom of the pump house pit floor. The diver crew retrieved the pieces for further inspection. While on location at the entrance to the suction bell, the maintenance pump team proceeded to hand rotate the pump. The diver crew did not identify any movement in the first stage impeller while the pump crew successfully rotated the shaft at the pump and motor coupling. Therefore, it was evident the impeller assembly was no longer connected to the motor shaft.

The Nuclear Service Water (NSWS) Pump is a deep draft vertical pump. It is a 1000 HP, two stage Bingham-Willamette VTM 30 x 44C pump. It is assembled with a suction bell, two bowl assemblies, four columns, one discharge head and motor to make the complete vertical assembly approximately 65 feet tall. It consists of five shafts and correspondingly four couplings. Only the uppermost motor to pump head shaft is accessible without complete pump removal and disassembly. Therefore complete removal is necessary for further investigation and repair.

Attachment I Page 6 of 20

NSWS Pump Monitoring The Catawba NSWS pumps are high safety significance pumps and receive in depth monitoring, trending, and analysis.

1) Vibration: Data collected quarterly via IWP and Maintenance testing programs.

Amplitude, frequency, and time waveforms are reviewed in detail for any changes. Data is reviewed by Category III and IV certified vibration analysts.

The most recent vibration data collected on the lB NSWS pump was collected just prior to the recent IEOC 17 RFO. No abnormal data was evident.

2) Pump Pressure and Flow: Suction and discharge pressure as well as flow is monitored closely on a quarterly basis via required procedural TWP testing.

Suction and discharge pressure have remained within acceptable values.

3) Oil Analysis Data: Motor oil samples are collected on a quarterly basis. No significant changes in oil quality have been noted.

4) Preventive Maintenance: Pump assemblies are replaced on a periodic frequency based on vendor recommendations and industry experience for this type of pump.

The pump was last refurbished is 2003. Per the preventative maintenance program the next overall/rebuild should not be needed until 2015.

A comprehensive review of the previous 4 quarters of in-service test data for the lB NSWS Pump was reviewed to verify that no performance degradation had occurred prior to the failure of the pump on July 12, 2008. The pump flow rate and discharge pressure were well within the established acceptance criteria. The pump motor inboard and outboard vibration readings were well below the acceptance criteria. There were no negative trends noted on any of the measured parameters. A review of previous work history on the 1B NSWS Pump did not identify any work activity generated as a result of degrading pump conditions indicative of impending coupling failure. Therefore, the failure of the pump coupling was not predictable based on the quarterly test data. When completed, the results of the Root Cause Investigation will be incorporated into the NSWS pump monitoring program.

Additional Actions Based on the above discussion, Catawba has been actively monitoring pump data and this failure could not have been predicted. Neither a routine nor an exigent TS amendment request could have been processed within the 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months period. Therefore, an emergency TS amendment is required to preclude a shutdown.

Page 7 of 20

9.

Technical Evaluation:

Extent of Condition Discussion This is a proven pump design with an excellent operating history. NSWS Pumps are being changed out due to aging and have not had a history of failures or operational issues. This pump was completely refurbished in 2003. Each of the NSWS pumps have been refurbished twice during the operation of Catawba with only normal wear identified during the refurbishments/inspections. There have been no design changes to the pumps that would create a common mode failure. The A NSWS pump pit was verified free of foreign material in May 2008 during the unit I refueling outage. The B NSWS pump pit was inspected on July 12, 2008. The only foreign material identified in the B pit was associated with the failed coupling. Operating and maintenance practices have not changed. Therefore, it is concluded that this failure is not transportable to the other pumps.

The NSWS pump changeout schedule is shown below:

NSWS Pump Date of Last Date of Next Scheduled Changeout Changeout IA 2008 2020 lB 2003 Failed, Root Cause 2A 2004 2018 2B 1998 2012 Spare 1991 2009 (1B replacement)

Duke plans to changeout the lB pump at the next refueling outage, which is the lEOC18 outage in November 2009.

Condition of IA NSWS Pump The JA NSWS pump was refurbished with a new rotating element during the Unit 1 refueling outage in May/June 2008. The pump was subsequently tested following replacement during this refueling outage and verified to meet its flow requirements for single pump and dual pump alignments. The performance parameters of the I A NSWS pump indicate the pump is in good running condition and considered reliable for many years of service. The next planned overhaul is the refueling outage in 2020. In addition to the overhaul completed during the refueling outage in May/June, 2008, the "A" NSWS pit was inspected for cleanliness of the suction intake of both the IA and 2A Nuclear Service Water Pumps. During this lB pump replacement, the IA NSWS pump and its support systems will be considered protected equipment. No scheduled maintenance will be performed on those systems.

Page 8 of 20.

Condition of 2A NSWS Pump The 2A NSWS pump was refurbished with a new rotating. element during the Unit 2 refueling outage in November 2004. The pump was subsequently tested following replacement during this refueling outage and verified to meet its flow requirements for single pump and dual pump alignments. Subsequently, since the November 2004 overhaul, the pump performance parameters are verified on a quarterly basis per the In-service Testing Requirements. The results indicate the 2A NSWS pump is in good running condition and considered reliable for many years of service. The next planned overhaul is the refueling outage in' 2018. During the refueling outage in May, 2008, the "A" NSWS pit was inspected for cleanliness of the suction intake of both the IA and 2A Nuclear Service Water Pumps. During this lB pump replacement, the IA NSWS pump and its support systems will be considered protected equipment. No scheduled maintenance will be performed on those systems.

Condition of the Spare Pump for lB The replacement pump for the failed lB Nuclear Service Water Pump is the IA Pump which was removed from service during the refueling outage (1EOC17) in May/June, 2008. The l A Pump was operating well and removed from service due to age as part of the station's pump overhaul/replacement plan for equipment reliability. The IA Pump performance prior to removal from service has been specifically reviewed to determine acceptable service for installation as the replacement I B pump. The In-service Test data from the previous four tests, specifically the flow rate, discharge pressure and pump vibration data demonstrate the pump will operate smoothly within the test acceptance criteria. Prior to installation of the spare pump, an inspection of the assembly will be performed to assess parts requiring refurbishment/replacement. This pump will be replaced in next refueling outage in November 2009 (lEOC18). While this pump is being replaced, the NSWS pit B will be drained and a cleanliness inspection will be performed. Once in service, the 1B NSWS pump will be operated to the extent practicable. This pump will be in the normal equipment rotation and will be operated as required to support routine train maintenance activities.

Condition of the 2B NSWS Pump The 2B NSWS pump was refurbished with a new rotating element during the Unit 2 refueling outage in September 1998. The pump was subsequently tested following replacement during this refueling outage and verified to meet its flow requirements for single pump and dual pump alignments. Subsequently, since the September 1998 overhaul, the pump performance parameters are verified on a quarterly basis per the In-service Testing Requirements. The results indicate the 2B NSWS pump is in good running condition and considered reliable for many years of service. The next planned overhaul is the refueling outage in 2012. While the lB pump is being replaced, the NSWS pit B will be drained and a cleanliness inspection will be performed for the suction intake of both the IB and 2B Nuclear Service Water Pumps.

Page 9 of 20

Testing Requirements:

Following the replacement of the IB NSWS pump, the pump will be tested for operational readiness in accordance to the 1998 ASME Code. The 1998 ASME Code for the In-service Test Requirements mandate the pump to be tested for performance flow and pressure parameters and axial and radial vibration. These parameters are tested quarterly per the program requirements. The results are evaluated against Acceptable, Alert or Unacceptable Limits. These tests include head curve verification, flow, pressure, vibration followed by train related flow balance which confirms operability.

Current Plant Status At the time of the incident NSWS pipe work on buried piping was in progress and various locations were uncovered for pipe inspections. Currently, this work has been put on hold and the buried piping has been covered per the requirements for tornado missile protection.

Additional Discussion:

The Containment Spray System consists of two separate trains of equal capacity, each capable of meeting the system design basis spray coverage. Each train includes a containment spray pump, one containment spray heat exchanger, spray headers, nozzles, valves, and piping. Each~train is powered from a separate Engineered Safety Feature (ESF) bus. The refueling water storage tank (RWST) supplies borated water to the Containment Spray System during the injection phase of operation. In the recirculation mode of operation, containment spray pump suction is transferred from the RWST to the containment recirculation sump(s).

The RWST is protected by a missile proof barrier wall which ensures a sufficient quantity of refueling water is retained in the tank to allow for an emergency cooldown in the event the tank is punctured by a missile. The RWST is a safety related seismic category I structure.

When the containment spray system suction is from the containment recirculation sump, its associated heat exchanger receives NSWS flow for cooling. During the extended time period this flow will not be available. However this does not affect the initial injection flow provided.

There are several sources of water available to the AFW pumps. The preferred sources are non-safety grade condensate quality, located in the Turbine and Service Buildings.

These are called the condensate storage system. The condensate storage system is formed from the Upper Surge Tanks (two 42,500 gallon tanks per unit) and the Condenser Hotwell (normal operating level of 170,000 gallons). The condensate storage system supplies the AFW requirements during normal system operating modes; but, since the condensate storage system is not safety related its availability is not assured. The Page 10 of 20

assured source of supply to the AFW pumps is provided by the safety related portion of the Nuclear Service Water System.

TS 3.7.6 requires the condensate storage system to be operable in modes 1, 2, 3 and mode 4 when steam generators are relied upon for heat removal. The condensate storage system contains sufficient cooling water to remove decay heat for 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months following a reactor trip from 100% Rated Thermal Power (RTP), and then to cool down the reactor coolant system (RCS) to RHR entry conditions, assuming a natural circulation cooldown.

In doing this, it retains sufficient water to ensure adequate net positive suction head for the AFW pumps during cooldown, as well as account for any-losses from the steam driven AFW pump turbine, or before isolating AFW to a broken line.

Another non-safety grade source of condensate water for the AFW pumps is the Auxiliary Feedwater Condensate Storage Tank (CACST). Each unit has a CACST that is maintained full by a recirculation flow of condensate from the condensate system and overflow to the condensate storage system. The CACST holds approximately 42,500 gallons of condensate grade water.

For emergency events, when none of the condensate grade sources are available, two redundant and separate trains of nuclear service water are available. The water supplied by the two nuclear service water sources is of lower quality; however, safety considerations override those of steam generator cleanliness. The NSWS assured source of water supply is configured into two trains. The turbine driven AFW pump receives NSWS from both trains of NSWS, therefore, the loss of one train of assured source renders only one AFW train inoperable. The remaining NSWS train provides an operable assured source to the other motor driven pump and the turbine driven pump. Therefore, during the extended time period, the lB AFW pump will be capable of starting and providing water to the steam generators from the non-safety sources and only its safety-related source from the NSWS will be affected.

Therefore, Catawba is requesting an extension of the Completion time to support repair of the 1B NSWS pump. The plant configuration during this time frame will still be able to support Chapter 15 accident analysis. The probabilistic risk assessment discussed below describes the effect of the extension of the Completion time.

Page 11 of 20

Risk Evaluation Duke has used a risk-informed approach to determine the risk significance of extending the current Technical Specification associated with the lB Nuclear Service Water System (NSWS) pump work. The Unit 1 extension is for an additional 144 hours0.00167 days 0.04 hours 2.380952e-4 weeks 5.4792e-5 months for a total time of 216 hours0.0025 days 0.06 hours 3.571429e-4 weeks 8.2188e-5 months .

The cumulative risk impact for this evolution is the sum of Part I and Part 2 risk numbers. This is shown below:

Catawba Unit 1 Unit 1 Part 1 (3 days)*

Part 2 (3 days)**

Total (Part I+Part (2 RN pumps)

(1 RN pump) 2)

6 days dCDF/yr 6.9E-07 4.5E-07 1.1E-06 dLERF/yr 2.7E-08 1.4E-08 4.1E-08 Part 1 (3 days)*

Part 2 (6 days)***

Total (Part 1+Part (2 RN pumps)

(I RN pump) 2)

9 days ICCDP 6.9E-07 8.9E-07 1.6E-06 ICLERP 2.7E-08 2.8E-08 5.5E-08

3 days beyond the original 72 hr TS CT

- 3 days of the 6 days extension period beyond the original 72 hr TS CT

- - 6 days representing the original 72 hr TS CT plus 3 days beyond the original 72 hr TS CT The delta CDF associated with the 6 day extension related to the NS and CA assured source is approximately 1.1E-06. The result is slightly above the RG-1.174 guidance of 1.OE-06. The ICCDP is estimated to be 1.6E-06. The result is above the RG-1.177 guidance (5E-07).for a permanent TS change.

The LERF results are less limiting than the CDF results.

Dominant Sequences The dominant sequences are reactor coolant pump seal LOCAs that occur when all RN is lost as an initiating event (dominated by some common cause failure of the available RN pumps); failure to restore cooling to the RCP seals (both SSF and YD) with failure to trip the RCPs prior to seal failure.

The dominant SSF failure is a failure to activate the SSF in time (human error).

The dominant YD failures are the human error of failure to activate or the inability to align YD because YD has been aligned to the other unit.

Attachment I Page 12 of 20

Impact of PRA Analysis on Fire and Flooding Events There were few fire initiated cut sets above the CDF and LERF truncation limits.

Additionally there were few flood cut sets. Fires and floods contributed negligibly to the CDF and LERF results.

RG 1.200 Assessment In accordance with the ASME standard [Reference 6] and RG 1.200 [Reference 5] Duke has made an assessment of all the ASME Supporting Requirements (SRs).

The Catawba PRA fully meets 224 of the 306 ASME PRA Standard Supporting Requirements (SRs), as modified by Reg. Guide 1.200. In addition, 24 of the SRs are not applicable to the Catawba PRA, either because the referenced techniques are not utilized in the PRA or because the SR is not required for Capability Category II.

Of the 58 open SRs, 14 are of a technical, nature. The remaining open SRs require enhanced documentation. However, none of the open items are expected to have a significant impact on the PRA results or insights, as discussed in Attachment 5 of this document.

PRA Model The Catawba PRA is a full scope PRA including both internal and external events. The model includes the necessary initiating events (e.g., LOCAs, transients) to evaluate the frequency of accidents. The previous reviews of the Catawba PRA, NRC and peer reviews, have not identified deficiencies related to the scope of initiating events considered.

The Catawba PRA includes models for those systems needed to estimate core damage frequency. These include all of the major support systems (e.g., ac power, service water, component cooling, and instrument air) as well as the mitigating systems (e.g., emergency core cooling). These systems are generally modeled down to the component level, pumps, valves, and heat exchangers. This level of detail is sufficient for this application.

Truncation Limit Truncation issues are not an issue with this risk calculation. The analysis for the current configuration was performed at the same truncation level as the base case (5.OE-10 for CDF and 5.OE-I1 for LERF). A review of the cut sets shows that loss of nuclear service water with a failure of drinking water backup cooling to the "A" charging pump with a corresponding failure to initiate the SSF are in most of the top cut sets. There is adequate representation of the expected failure in the results that drive the answer so that there was no need to solve to any lower truncation levels. The'issue identified in RG 1.177 (most of the failures appearing near the truncation cutoff) does not exist in this analysis. Additionally, an explicit Page 13 of 20

truncation level analysis was performed for Revision 3a of the PRA consistent with ASME standard and RG 1.200 requirements.

Uncertainty and Sensitivity Duke agrees with the RG 1.177 statement that risk analyses of CT extensions are relatively insensitive to uncertainties. The PRA did not credit equipment repair so there are no uncertainties to be evaluated for that issue. Important systems are required to remain in service during the CT so no issues with mean downtimes should exist. Thus uncertainty and sensitivity are not expected to alter the conclusions of the evaluation.

Results of Reviews with Respect to this LAR A review of the analyses (cut sets and pertinent accident sequences) was made for accuracy and completeness. Specifically, cut sets generated for the solutions were screened and invalid cut sets were removed and appropriate recovery events applied. This process is documented in Duke calculations. The review verified that the calculations adequately modeled the effects of the NSWS system unavailability.

Consistent with the work place procedures governing PRA analysis, this calculation has undergone independent checking by a qualified reviewer.

Additionally the Catawba Plant Operations Review Committee (PORC) and Duke Nuclear Safety Review Board (NSRB) reviewed and approved this amendment request package.

Tier 2 Assessment: Avoidance of Risk-significant Plant Equipment Outage Configurations Tier 2 provides reasonable assurance that risk-significant plant equipment outage configurations will not occur when specific plant equipment is out of service consistent with the proposed TS change. Specific components and trains have been identified that are not to be taken out of service during the period of the extended CT.

Duke has several Work Process Manual procedures and Nuclear System Directives that are in place at Catawba Nuclear Station to ensure that risk-significant plant configurations are avoided. The key documents are as follows:

" Nuclear System Directive 415, "Operational Risk Management (Modes 1-

3) per i0 CFR 50.65 (a.4)".

" Nuclear System Directive 403, "Shutdown Risk Management (Modes 4, 5, 6, and No-Mode) per 10 CFR 50.65 (a.4)".

" Work Process Manual, WPM-609, "Innage Risk Assessment Utilizing ORAM-SENTINEL".

o Work Pr'ocess Manual, WPM-608, "Outage Risk Assessment Utilizing ORAM-SENTINEL".

--Attachment 1 Page 14 of 20

The proposed changes are not expected to result in any significant changes to the current configuration risk management program. The existing program uses a blended approach of quantitative and qualitative evaluation of each configuration assessed. The Catawba on-line computerized risk tool, ORAM-Sentinel, considers both internal and external initiating events with the exception of seismic

events. Thus, the overall change in plant risk during maintenance activities is expected to be addressed adequately in accordance with RG 1.177 considering the.

proposed Technical Specifications.

Tier 3 Assessment: Maintenance Rule Configuration Control 10 CFR 50.65(a)(4), RG 1.182, and NUMARC 93-01 require that prior to performing maintenance activities, risk assessments shall be performed to assess and manage the increase in risk that may result from proposed maintenance activities. These requirements are applicable for all plant modes. NUMARC 91-06,requires utilities to assess and manage the risks that occur during the

performance of outages.

As stated above, Duke has approved procedures and directives in place at Catawba to ensure the requirements of the Maintenance Rule are implemented.

These documents are used to address the Maintenance Rule requirements, including the on-line (and off-line) Maintenance Policy requirement to control the safety impact of combinations of equipment removed from service.

More specifically, the Nuclear System Directives address the process; define the program, and state individual group responsibilities to ensure compliance with the Maintenance Rule. The Work Process Manual procedures provide a consistent process for utilizing the computerized software assessment tool, ORAM-.

SENTINEL, which manages the risk associated with equipment inoperability.

ORAM-SENTINEL is a Windows-based computer program designed by the Electric Power Research Institute as a tool for plant personnel to use to analyze and manage the risk associated with all risk significant work activities including assessment of combinations of equipment removed from service. It is independent of the requirements of Technical Specifications and Selected Licensee Commitments.

The ORAM-SENTINEL models for Catawba are based on a "blended" approach of probabilistic and traditional deterministic approaches. The results of the risk assessment include a prioritized listing of equipment to return to service, a prioritized listing of equipment to remain in service, and potential contingency considerations.

Additionally, prior to the release-of work for execution, Operations personnel must consider the effects of severe weather and grid instabilities on plant operations. This qualitative evaluation is inherent of the duties of the Work Control Center Senior Reactor Operator (SRO). Responses to actual plant risk Attachment I Page 15 of 20

due to severe weather or grid instabilities are programmatically incorporated into applicable plant emergency or response procedures.

Previous NRC RAIs on PRA Model Duke reviewed previous requests for additional information from a previous emergency TS submittal and provides the following responses:

Question 1:

The submittal identified administrative controls to assure plant changes are reflected in the PRA model, but has not stated whether there are outstanding plant changes not yet reflected in the model, and whether those would impact this analysis.

Response

All outstanding plant changes that are not included in the current base PRA model (Rev. 3a) were reviewed and evaluated for this application. Of this population, 6 plant changes were determined to require further evaluation. These are summarized below:

Issue Resolution Closure of two valves important to ISLOCA is important for LERF sequences.

ISLOCA sequences may not LERF is not the limiting metric.

occur.

Additionally ISLOCA sequences for LERF were not the dominant sequences.

Add alternate feedwater makeup This would be a risk reduction. No action line to each S/G.

was taken so the model is conservative.

The chemical and volume control The model was revised to include new failure system model does not capture all mechanism to reflect a 50-50 chance that an of the unavailability for drinking "A" charging pump will receive backup flow water backup cooling to the "A:

from drinking water.

charging pump.'

Increase the exposure time for 6 Exposure time was increased for the 6 basic basic events to reflect current events based on current testing schedule.

testing.

Replace recovery in model with Recovery was set equal to 1.0 in the model.

explicit logic.

This recovery event did not appear in dominant cut sets.

Modifications to the NSWS Expected CDF improvements. Current headers to add crossover between model is bounding.

EDGs Page 16 of 20

Question 2:

The submittal did not address truncation levels per RG 1.177 2.3.3.4.

Response

Truncation issues are not an issue with this risk calculation. The analysis for the current configuration was performed at the same truncation level as the base case (5.0E-10 for CDF and 5.0E-11 for LERF). A review of the cut sets shows that loss of nuclear service water with a failure of YD backup cooling to the "A" NV pump with a corresponding failure to initiate the SSF are in most of the top cut sets. There is adequate representation of the expected failure in the results that drive the answer so that there was no need to solve to any lower truncation levels.

The issue identified in RG 1.177 (most of the failures appearing near the truncation cutoff) does not exist in this analysis. Additionally, an explicit truncation level analysis was performed for Revision 3a of the PRA consistent with ASME standard and RG 1.200 requirements.

Question 3:

The submittal needs to identify if credit is taken for the SSF in the risk calculations, and should also address if equipment repair is credited.

Response

Credit is taken for the SSF. Catawba has taken action to ensure that the SSF will be available during the extended CT period. The Catawba PRA does not take credit for equipment repair.

Question 4:

The submittal did not address uncertainty or sensitivity issues per RG 1. 177 2.3.5.

Response

Duke agrees with the RG 1.177 statement that risk analyses of CT extensions are relatively insensitive to uncertainties. We did not credit equipment repair so there are no uncertainties to be evaluated for that issue. We required important systems to remain in service during the CT so no issues with mean downtimes should exist. Therefore, for the typical issues related to uncertainties, there should be no effect on our analysis.

Question 5:

Provide clarification that the seismic contribution is negligible compared to the non-seismic results.

Response

We have numerically reviewed the seismic impact for the nuclear service water system, including a loss of emergency diesel generator using the previous PRA model and determined that the seismic contribution is negligible compared to the Attachment I Page 17 of 20

non-seismic results. Based on the expected configuration during the time period of the CT extension, there is no reason to expect that that conclusion would change for the current model.

References

1. E-Mail: Tony Jackson to Randy Hart,

Subject:

Potential Emergency TS Change for lB RN Pump, July 13, 2008.

2. Phone Call: M.S. Kitlan, Jr, to Randy Hart, July 13, 2008.

3. U.S. Nuclear Regulatory Commission, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," Regulatory Guide 1.174, Revision 1, November 2002.

4. U.S. Nuclear Regulatory Commission, "An Approach for Plant-Specific, Risk-Informed Decisionmaking: Technical Specifications," Regulatory Guide 1.177, Revision 0, August 1998.

5.

U.S. Nuclear Regulatory Commission, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," Regulatory Guide 1.200, Revision 1, January 2007.

6. American Society of Mechanical Engineers, "Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications," ASME-RA-Sc-2007.

7.

SAAG 592, Rev.2: Determination of Risk Significance when Taking an RN Loop Out of Service Beyond Tech. Spec. Limits, (Catawba PRA Rev 2a) August 2000.

8. SAAG 725, Risk Significance Determination for RN Piping Replacement Project (CNCE 71.424) (Catawba PRA Rev 2b), August 2002.

9. SAAG 278, Catawba Rev. 3 Seismic PRA Analysis, November 2003.

10. CNC-1535.00-00-0025, Risk Significance Determination for Proposed Catawba RN Loop LCOs, Revision 3, April 2006.

11. Catawba PRA Revision 3a.

12. NUREG/CR-5497, Common Cause Failure Parameter Estimations, October 1998.

13. SAAG 710, Catawba RN Essential Header NOED, 2002.

14. SAAG 670, Catawba PRA Rev. 3 Common Cause Failure Analysis, Table 1, Revision 3, October 2005.

15. DPC-1535.00-00-0013, PRA Quality Self-Assessment, DRAFT.

16. TSAIL printout from site (see electronic documents).

Page 18 of 20

Operation and Maintenance Restrictions for the Duration of the Extension These items are listed in Attachment 4 to this document.

10.

Regulatory Safety Analysis:

10. 1 No Significant Hazards Consideration:

Duke has concluded that operation of the Catawba Nuclear Station Unit I in accordance with the proposed change to the Technical Specifications (TS) does not involve a significant hazards consideration. Duke's conclusion is based on its evaluation, in accordance with 10 CFR 50.91 (a) (1), of the three standards set forth in 10 CFR 50.92 (c).

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

Response: No.

The lB AFW pump and the IB CSS safety related functions are as accident mitigators and are not required unless an accident occurs. The 1B AFW pump and lB CSS do not affect any accident initiators or precursors. The proposed extension of the Required Action Completion Time does not affect the lB AFW pump's and lB CSS interaction with any system whose failure or malfunction could initiate an accident. Therefore the probability of an accident previously evaluated is not significantly increased.

The risk evaluation performed in support of this amendment request (Reference Section 9) demonstrates that the consequences of an accident are not significantly increased. As such, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

ii. Does the proposed amendment create the possibility of a new or different kind of accident from anypreviously evaluated?

Response: No.

This change does not create the possibility of a new or different kind of accident from any accident previously evaluated. No new accident causal mechanisms are created as a result of the NRC granting of this proposed change. No changes are being made to the plant which will introduce any new or different accident causal mechanisms.

Attachment I Page 19 of 20

i. Does the proposed amendment involve a significant reduction in the margin of safety?

Response: No.

Based on the availability of redundant systems, the restrictions on maintenance and operation of required systems, and the low probability of an accident, Catawba concludes that the reduction of availability of the lB AFW pump and the 1B CSS does not result in a significant reduction in the margin of safety.

The 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 significantly impacted by the proposed change. The risk implications of this request were evaluated and found to be acceptable.

10.2 Applicable Regulatory Requirements/Criteria:

The analysis presented in this LAR demonstrates that Catawba will remain in compliance with the applicable regulations and requirements. These are:

10 CFR 50.46 and 10 CFR 50, Appendix A, General Design Criterion (GDC) 44,45 and 46.

This LAR is being submitted in accordance with 10 CFR 50.90 and 50.91 (a) (5).

11.

Environmental Consideration:

The proposed change does not involve a significant hazards consideration, a significant change in the types of or significant increase in the amounts of effluents that may be released offsite, or a significant increase in individual or cumulative occupational radiation exposure. Therefore, the proposed changes meet the eligibility criteria for the categorical exclusion set forth in. 10 CFR 51.22 (c) (9).

Therefore, pursuant to 10 CFR 51.22 (b), an environmental assessment of the proposed change is not required.

12.

Precedent:

None Page 20 of 20

ATTACHMENT 2 MARKED-UP CATAWBA TECHNICAL SPECIFICATION

Amendment Implementation Number Additional Condition Date 173 The schedule for the performance of new and By January 31, 1999 revised surveillance requirements shall be as follows:

For surveillance requirements (SRs) that are new in Amendment No. 173 the first performance is due at the end of the first surveillance interval that begins at implementation of Amendment No. 173. For SRs that existing prior to Amendment No. 173, including SRs with modified acceptance criteria and SRs who intervals of performance are being extended, the firstperformance is due at the end of the first surveillance interval that begins on the date the surveillance was last performed prior to implementation of amendment No. 173. For SRs that existed prior to Amendment No. 173, whose intervals of performance are being reduced, the first reduced surveillance interval begins upon completion of the first surveillance performed after implementation of Amendment No. 173 180 The maximum rod average burnup for any rod Within 30 days of shall be limited to 60 GWd/mtU until the date of amendment.

completion of an NRC environmental assessment supporting an increased limit.

In association with the ECCS sump strainer.

Within 30 days of modification and Generic Safety Issue (GSI)-

date of amendment 191 requirements:

and no later than December 31, 2007

1.

Unit 1 shall enter Mode 5 for the outage to install the sump strainer modification no later than May 19, 2008 and

2.

The Unit 1 sump strainer modification shall be completed prior to entry into Mode 4 after May 19, 2008.

j Nsor A Amendment No. 237

'AA.

I#/'f-Insert A The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months allowed outage time of Technical Specification 3.7.5 Action "B" for the lB AFW pump which was entered at 1041 on July 12, 2008 may be, extended by an additional 144 hours0.00167 days 0.04 hours 2.380952e-4 weeks 5.4792e-5 months . Upon completion of the repair and restoration of the 1B NSWS pump, this License Condition is no longer applicable and will expire at 1041 on July 21,2008.

The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months allowed outage time of Technical Specification 3.6.6 Action "A" for the 1B CSS which was entered at 1041 on July 12, 2008 may be extended by an additional 144 hours0.00167 days 0.04 hours 2.380952e-4 weeks 5.4792e-5 months . Upon completion of the repair and restoration of the I B NSWS pump, this License Condition is no longer applicable and will expire at 1041 on July 21, 2008.

July 15, 2008 at 1041

ATTACHMENT 3 RE-TYPED CATAWBA TECHNICAL SPECIFICATION

Amendment Implementation Number Additional Condition Date The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months allowed outage time of July 15, 2008 at Technical Specification 3.7.5 Action "B" 1041 for the 1 B AFW pump which was entered at 1041 on July 12, 2008 may be extended by an additional 144 hours0.00167 days 0.04 hours 2.380952e-4 weeks 5.4792e-5 months .

Upon completion of the repair and restoration of the 1 B NSWS pump, this License Condition is no longer applicable and will expire at 1041 on July 21, 2008.

The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months allowed outage time of Technical Specification 3.6.6 Action "A" for the 1 B CSS which was entered at 1041 on July 12, 2008 may be extended by an additional 144 hours0.00167 days 0.04 hours 2.380952e-4 weeks 5.4792e-5 months . Upon completion of the repair and restoration of the 1 B NSWS pump, this License Condition is no longer applicable and will expire at 1041 on July 21,2008.

Renewed License No. NPF-35 Amendment No.

ATTACHMENT 4 REGULATORY COMMITMENTS

LIST OF REGULATORY COMMITMENTS The following list identifies those actions committed to by Catawba in this document, for the duration of the extension. Any other statements made in this licensing submittal are provide for informational purposes only and are not considered to be regulatory commitments. Please direct any questions you may have in this matter to A.P. Jackson at (803) 701-3742.

Regulatory Commitment Due Date The proposed changes to the Catawba TS will be July 15, 2008 implemented prior to the end of the original 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Completion Time (1041 on 7/15/08).

During the extended Completion Time period, no major July 15, 2008 maintenance or testing will be planned on the remaining operable NSWS "A" header. In addition, for Unit 1, during this period, no major maintenance or testing will be planned on the operable equipment that relies upon "A" Train NSWS as a support system.

During this extended Completion Time period, no major July 15, 2008 maintenance or testing will be planned on the Unit IA AFW pump and Unit I turbine driven AFW pump.

During this extended Completion Time period, no major July 15, 2008 maintenance or testing will be planned on the Unit I A train and B train CCW system.

During this extended CompletionTime period, no major July 15, 2008 maintenance or testing will be planned on the Unit I A train Chemical Volume and Control System.

During this extended Completion Time period, no major July 15, 2008 maintenance or testing will be planned on the SSF.

No major maintenance or testing will be planned on the July 15, 2008 portions of the drinking water system that are relied upon to provide backup cooling to the "A" charging pumps.

During the extended Completion Time period, the AFW July 15, 2008 system train I B motor driven will remain ayailable.

During this extended Completion Time period, no. major July 15, 2008 maintenance or testing will be planned on the 1 A and I B essential AC power switchgear including 4160 volt busses, load centers and motor control centers.

During this extended Completion Time period, no major July 15, 2008 maintenance or testing will be 'planned on switchyard components, the IA and 2A emergency diesel generators, and the transformers that feed the IA, I B, and 2A 4160 volt busses.

An action taken by Catawba to reduce the likelihood of an July 15, 2008 operator failing to get to the SSF and performing the required actions is to station an individual in the SSF continuously. This individual is trained on how to operate the SSF diesel generator and the standby makeup pump to establish an alternate method of reactor coolant pump seal injection. This will provide additional assurance that the SSF will be available during the extended completion time.

Prior to entering the extended Completion Time the July 15, 2008 operating crews will review the procedures regarding starting the SSF and establishing backup cooling to an "A" charging pump.

Catawba will perform a cleanliness inspection while the July 21, 2008 NSWS pit B is drained for pump change outl To mitigate the risk of a potential core damage event, an July 15, 2008 operator action has been identified. This involves dispatching operators to throttle key AFW valves to supply the flow to the steam generators prior to the depletion of the vital batteries, thereby preventing steam

generator overfill and thus protecting the steam supplies to the AFW turbine driven pump. Catawba will dedicate an operator on each shift with this responsibility.

Catawba has installed permanent flood protection barriers July 15, 2008 in the turbine building to mitigate turbine building flooding. In addition, to help reduce any potential flooding issues, no major maintenance or testing will be planned on the Condenser Circulating Water System.

Duke commits to a change out of the I B NSWS pump at Prior to the end of I EOC 18 Refueling outage the next scheduled refueling outage.

Catawba will perform a detailed cause evaluation of this November 30, 2008.

failure of the I B NSWS pump. This cause evaluation will be compared to the cause evaluation completed for the failure of the I B centrifugal charging pump earlier this year. This result of this comparison will be used to evaluate any potential enhancements to Catawba's pump preventive maintenance program.

NOTES:

The above commitments do not preclude the performance ofany TS required surveillances provided that the surveillances do not render equipment inoperable.

The above commitments do not preclude any work that becomes necessary due to emergent equipment issues. Any proposed work would be evaluated against this submittal and the appropriate risk mitigate measures would be put in place.

ATTACHMENT 5 CATAWBA PRA QUALITY DISCUSSION

ATTACHMENT 5 Table A-I Catawba PRA - Open ASME PRA Standard Supporting Requirements Cateory II

-equirv mn tRsol.tion Doc i, i ia,

o ti.ton

.tI cn SR

.C.,.cu n ato......

USE realistic, applicable (i.e., fiom similar plants)

Catawba No impact is expected thermal hydraulic analyses to determine the Thermal-Perform analyses with the since success criteria are accident progression parameters (e.g., timing Hydraulic consistent with peer temperature, pressure, steam) that could potentially Success MAAP.

plants per the PWROG affect the operability of the mitigating systems.

Criteria PSA Database.

(See SC-B4.)

calcs.

For each accident sequence, IDENTIFY the Cut set review during model phenomenological conditions created by the Catawba integration and when accident progression. Phenomenological impacts Rev 3a PRA supporting applications Many phenomenological include generation of harsh environments affecting Model should address this.

effects are already temperature, pressure, debris, water levels, Integration Suggest adding this considered in the model.

AS-B3 humidity, etc. that could impact the success of the Partial

Notebook, guidance to workplace Technical system or function under consideration [e.g., loss of CNC-procedure XSAA-103.

Cut set review considers pump net positive suction head (NPSH), clogging 1535.00 possible additional of flow paths]. INCLUDE the impact of the 0061, Rev.

phenomenological effects.

accident progression phenomena, either in the 2, July 2006 accident sequence models or in the system models.

ESTABLISH definitions of' SSC boundaries, failure Catawba modes, and success criteria consistent with Failure Rate corresponding basic event definitions in Systems

Database, DA-Analysis (SY-A5, SY-A7, SY-A8, SY-AIO through CNC-ReviseNo impact is expected for Ala SY-A13 and SY-B4) for failure rates and common No 1535.0000 discuss component Documentation documentation issues.

cause failure parameters, and ESTABLISH 0029, Rev.

boundaries definitions.

boundaries of unavailability events consistent with 2, January coriresponding definitions in Systems Analysis (SY-2006 A I8).

Page 1 of 25

ATTACHMENT 5 CS ateg-oryý I WR~cuiremen~ts Nlt i C-"!

ý Ref.,

b~~outo

~DoclimciltationA IIIedrpcto For parameter estimation, GROUP components Catawba Revise the data caic. to according to type (e.g.. motor-operated pump, air-Failure Rate segre san dThis is a refinemrent to the operated valve) and according to the characteristics

Database, rati equipment failure rates.

CNC opewever, compnee data.

of their usage to the extent supported by data: (a)

Partial CNCowoeenn sce groupnd DA-B I IPrilSegregate components by Technical Hoersnemt mission type (e.g., standby, operating) (b) service 1535.00 C. Zeacomponents are grouped I 1ý service condition to the condition (e.g., clean vs. untreated water, air) 0029, Rev.

extent supported by the appropriately, the overall 2, January data.

impact should be small.

2006 DO NOT INCLUDE outliers in the definition of a Catawba group (e.g., do not group valves that are never Failure Rate Revise the data calc. to tested and unlikely to be operated with those that

Database, include a specific discussion DA-.B2 are tested or otherwise manipulated frequently)

Partial CNC-of outlier treatment (i.e., do Documentation No impact is expected for 1535.00 any outliers exist? If so, documentation issues.

0029, Rev.

how are these events 2, January considered and grouped?)

2006 EXAMINE coincident unavailability due to Developing maintenance for redundant equipment (both PRA Data, intrasystem and intersystem) based on actual plant Workplace Put in place a mechanism experience. CALCULATE coincident maintenance Procedure for identifying and unavailabilities that reflect actual plant experience.

XSAA-110, quantifying coincident Such coincident maintenance unavailability can Rev. 4, July unavailabilities. Incorporate additional maintenance on arise, for example, for plant systems that have 2007; in the system models those CA3 "installed spares," i.e., plant systems that have more Partial Catawba maintenance events allowed Technical be restricted during the redundancy than is addressed-by tech specs. For Component by technical specifications amended completion example, the charging system in some plants has a Failure Rate where 2 or more time.

third train that may be ott of service for extended Denominato components have periods of time coincident with one of the other r Estimates, maintenance events that are trains and yet is in compliance with tech specs.

SAAG 492, correlated with each other.

December '

1997 Page.2 of 25

ATTACHMENT 5 DA-D4 When the Bayesian approach is used to derive a distribution and mean vfilue of a parameter, CHECK that the posterior distribution is reasonable given the relative weight of evidence provided by the prior and the plant-specific data. Examples of tests to ensure that the updating is accomplished correctly and that the generic parameter estimates are consistent with the plant-specific application include the following: (a) confirmation that the Bayesian updating does not produce a posterior distribution with a single bin histogram (b) examination of the cause of any unusual (e.g.,

multimodal) posterior distribution shapes (c) examination of inconsistencies between the prior distribution and the plant-specific evidence to.

confirm that they are appropriate (d) confirmation that the Bayesian updating algorithm provides meaningful results over the range of values being considered (e) confirmation of the reasonableness of the posterior distribution mean value Partial Catawba PRA Common Cause

Analysis, CNC-1535.00 0028, Rev.

0, December 2005 Enhance thedocumentation to include a discussion of the specific checks performed on the Bayesian-updated data, as required by this SR.

Documentation No impact is expected for documentation issues.

USE generic common cause failure probabilities, Catawba Providedocumentation in consistent with available plant experience.

PRA EVALUATE the common cause failure Common S comparison of the probabilities consistent with the component Cause compon ofnthe DA-D6 boundaries.

Analysis, component boundaries D

No impact is expected for DA-D6

~~~~Partial CCassumed for the generic Documentation douetinise.

.~~~CNC-.

CC siaest hs documentation i~ssues.

1535.0 CCF estimates to those 1535.00-00-*

S0028, Rev, assumed in the Catawba PRA to ensure that these 0, December boundaries are consistent.

2005 Page 3 of 25

ATTACHMENT 5 RI NA0fpliScationIlltoi

.SR [=...

,++

,*.*C tgoyH~ fu~emns* +,

V j

/et~f*

G SRe esl io

,+*ocumnenatinojiK<

,V........

I Ihipa, on IDENTIFY, through.a review of procedures and practices, those calibration activities that if performed incorrectly can have an adverse impact on the automatic initiation of standby safety equipment.

HR-A2 Partial Catawba Human Reliability

Analysis, CNC-1535.00 0030, Rev.

0, December 2005 Enhance the HRA to consider the potential for calibration errors.

Technical Based on preliminary evaluations using the EPRI HRA calculator, calibration errors that result in failure of a single channel are expected to fall in the low 10-3 range.

Calibration errors that result in failure of multiple channels are expected to fall in the low 10-5 range. Relative to post-initiator HEPs, equipment random failure rates and maintenance unavailability, calibration HEPs are not expected to contribute significantly to overall equipment unavailability.

Page 4 ot 25

ATTACHMENT 5

~CNS.Rf.c;

>-ic-Re0'rtidnýr.ýteDc htImpact oh~

SR Ca(_oý1 k~urnet Nlet~forC-4SZ j1>.! NS RO Reofiin 0

IDENTIFY which of those work practices identified above (HR-Al, HR-A2) involve a mechanism that simultaneously affects equipment This is the documentation HR-A3 in either different trains of a redundant system or No pt the issuendesrieNo impact is expected for diverse systems [e.g., use of common calibration No part of the issue described in Documentation documentation issues.

C, SR's HR-A I and HR-A2.

equipment by the same crew on the same shift, a maintenance or test activity that requires realignment of an entire system (e.g., SLCS)].

PROVIDE an assessment of the uncertainty in the Pre-initiator HEPs are HEPs. USE mean values when providing point generally set to relatively estimates of HEPs.

'high screening values.

HR-D6 No Develop mean values for Tehncl i

HR-D6 No peiiitrH s.Technical Thus the suggested data pre-initiator REPs.

refinement is not expected to have a significant impact on this application.

When estimating HEPs EVALUATE the impact of the following plant-specific and scenario-specific performance shaping factors: (a) quality [type (classroom or simulator) and frequency] of the operator training or experience (b) quality of the Catawba written procedures and administrative controls (c)

Human availability of instrumentation needed to take Reliability Document in more detail the corrective actions (d) degree of clahrity of the

Analysis, influence of performance No impact is expected for HR-G3 rnean'ing of the cues/indications (e) human-machine Partial CNC-D~ocumentation correctivepin actions (d) degeeofcluityofth documentation issues.

interface (f) time available and time required to 1535.00 shaping factors on executione complete the response (g) complexity of detection, 0030, Rev.

human error probabilities.

diagnosis and decision-making, and executing the 0, December required response (h) environment (e.g., lighting, 2005 heat, radiation) under which the operator is working (i) accessibility of the equipment requiring manipulation (j) necessity, adequacy, and availability of special tools, parts, clothing, etc.

Page 5 of 25

ATTACHMENT 5 SR CN RH R.~tgr ApRqulemetsation' 0

BASE the time available to complete actions on Catawba appropriate realistic generic thermal-hydraulic Human analyses, or simulation from similar plants (e.g.,

Reliability plant of similar design and operation) (See SC-

Analysis, Enhance HRA No impact is expected for HR-G4 B4.). SPECIFY the point in time at which operators Partial CNC-Documentation are expected to receive relevant indications.

1535.00 0030, Rev.

0, December 2005 CHECK the consistency of the post-initiator HEP Document a review of the quantifications. REVIEW the HFEs and their final HFEs and their final HEPs HEPs relative to each other to check their relative to each other to HR-G6 reasonableness given the scenario context, plant No confirm their reasonableness D

ta No impact is expected for history, procedures, operational practices, and given the scenario context, documentation issues.

experience, plant history, procedures, operational practices, and experience.

Characterize the uncertainty in the estimates of the Use of mean values for HEPs, and PROVIDE mean values for use in the HEPs is expected to result quantification of the PRA results.

in an increase in post-initiator HEP values, in the base case model as well as for applications.

HR-G9 No Develop mean values for Technical Implementing post-initiator HEPs.C compensatory actions for the important operator actions is expected to have an offsetting effect, thereby reducing the HEPs.

Page 6 of 25

ATTACHMENT 5 HR-H2

_I(L!I i operator recovery actions only It, on a plant-specific basis: (a) a procedure is available and operator training has included the action as part of crew's training, or justification for the omission for one or both is provided (b) "cues" (e.g., alarms) that alert the operator to the recovery action provided procedure, training, or skill of the craft exist (c) attention is given to the relevant performance shaping factors provided in HR-G3 (d) there is sufficient manpower to perform the action.

Partial Catawba Hurman Reliability

Analysis, CNC-1535.00 0030, Rev.

0, December 2005 Develop more detailed documentation of operator cues, relevant performance shaping factors, and availability of sufficient manpower to perform the action.

Documentation No impact is expected for documentation issues.

IDENTIFY those initiating events that challenge Catawba normal plant operation and that require successful Internal mitigation to prevent core damage using a Initiator structured, systematic process for identifying Event initiating events that accounts for plant-specific Frequency Enhance the IE features. For example, such a systematic approach Data, CNC-No impact is expected for IE-AI Partial documnentation (as was done Documentation may employ master logic diagrams, heat balance 1535.00 documentation issues.

fault trees, or failure modes and effects analysis 003 1, Rev.

(FMEA). Existing lists of known initiators are also 0, January commonly employed as a starting point.

2006; Systems Analysis Page 7 of 25

ATTACHMENT 5 Category If eiiiieit kc111(1111 SAefi~N?

CN Ref'pRsluii REVIEW the plant-specific initiating event experience of all initiators to ensure that the list of challenges accounts for plant experience. See also IE-A7 IE-A3 Partial Catawba Internal Initiator Event Frequency Data, CNC-1535.00 0031, Rev.

0, January 2006 Perform a review of the plant-specific initiating event experience of all initiators to ensure that the list of challenges accounts for plant experience.

Technical Initiating events (other than ATWS) result in a plant trip and the generation of an LER.

These events are reviewed as part of the initiating events analysis. Fire and flood events that don't result in a reactor trip could potentially impact the frequencies assigned to the fire and flood initiators. However, fire and flood sequences are not significant contributors to the delta CDF in the PRA analysis for the LAR. Thus this

-open SR does not have a significant impact.

REVIEW generic analyses of similar plants to Catawba assess whether the list of challenges included in the Internal Ensure the list of challenges model accounts for industry experience.

Initiator included in the Catawba Event PRA accounts for industry IE-A3a Partial Frequency experience using a more Documentation No impact is expected for Data, CNC-recent reference, such as the documentation issues.

1535.00 WOG PSA Model and 0031, Rev.

Results Comparison 0, January Database - Revision 4.

2006 Page 8 of 25

ATTACHMENT 5 PERFORM a systematic evaluation of each system Catawba where necessary (e.g., down to the subsystem or Internal Provide documentation of a train level), including support systems, to assess the Initiator systematic evaluation of all possibility of an initiating event occurring due to a Eventplant systems, including failure of the system. USE a structured approach Frequency support systems (including No impact is expected for IE-A4

[such as a system-by-system review of initiating Partial those not explicitly modeled Documentation Data, CNC-documentation issues.

event potential, or an FMEA (failure modes and 1535.00 in the PRA), to assess the effects analysis), or other systematic process] to 00 00 possibility of an initiating assess and document the possibility of an initiating 0031, Rev.

event occurring due to a event resulting from individual systems or train 0failure of the system.

failures.

2006 When performing the systematic evaluation Catawba required in IE-A4, INCLUDE initiating events Internal resulting frlom multiple failures, if the equipment Initiator failures result from a common cause, and from Event Enhance the JE IE-A4a system alignments resulting from preventive and Pi Frequency documentation (as was done Documentation docimpact is expected for corrective maintenance.

Data, CNC-docuientation issues.

1535.0 in OSC-9068).

0031, Rev.

0, January 2006 In the identification of the initiating events, Catawba INCORPORATE (a) events that have occurred at Internal conditions other than at-power operation (i.e.,

Initiator during low-power or shutdown conditions), and for Event Enhance the IE No impact is.expected for IE-A5 which it is determined that the event could also Partial FreqDuency docuentation (as was done Documentation occur during at-power operation. (b) events Data, CNC-documentation issues.

resulting in a controlled shutdown that includes a 1535.00 scram prior to reaching low-power conditions, 0031, Rev.

unless it is determined that an event is not 0, January applicable to at-power operation.

2006 INTERVIEW plantpersonnel (e.g., operations, IE-A6 maintenance, engineering, safety analysis) to No Obtain plani personnel input Documentation No impact is expected for determine if potential initiating events have been (as was done in OSC-9068).

documentation issues.

overlooked.

Page 9 of 25

ATTACHMENT 5

~~~

~~

~

chnmcal or~~

xetdIiatn REVIEW plant-specific operating experience for Catawba initiating event precursors, for the purpose of Internal identifying additional initiating events. For Initiator example, plant specific experience with intake Event Include review of precursor E-A7 structure clogging might indicate that loss of intake Partial Frequency No impact is expected for JE-7 ICC._.ari.

events for their potential to Documentation structures should be identified as a potential Data, CNC-documentation issues.

initiating event.

1535.00 be initiating events.

0031, Rev.

0, January 2006 COMBINE initiating events into groups to facilitate definitioni of accident sequences in the Enhance the IE No impact is expected for IE-B 1 Accident Sequence Analysis element (para. 4.5.2)

No documentation (as was done Documentation documentation issues.

and to facilitate quantification in the Quantification in OSC-9068).

element (para. 4.5.8).

USE a structured, systematic process for grouping Catawba initiating events. For example, such a systematic Internal approach may employ master logic diagrams, heat Initiator balance fault trees, or failure modes and effects Event Document a structured, E-B2 analysis (FMEA).

Partial Frequency systematic grouping of No impact is expected for Data, CNC-initiating events (as was documentation issues.

1535.00 done in OSC-9068).

0031, Rev.

0, January 2006 Page 10 of 25

ATTACHMENT 5 11 Requireioi

,TcuicI.<<>u GROUP initiating events only when the following can be assured: (a) events can be considered similar Catawba in terms of plant response, success criteria, timing, Internal and the-effect on the operability and performance Initiator of operators and relevant mitigating systems; or (b)

Event Enhance documentation of events can be subsumed into a group and bounded Partial Frequency the grouping process (as Documentation oimpact is expected for by the worst case impacts within the "new" group.

Data, CNC-documentation issues.

DO NOT SUBSUME events into a group unless:

1535.00 (I) the impacts are comparable to or less than those 003 1, Rev.

of the remaining events in that group, AND (2) it is 0, January demonstrated that such grouping does not impact 2006 significant accident sequences.

DOCUMENT the assumptions and sources Enhance the IE No impact is expected for IE-D3 uncertainty with the initiating event analysis.

No documentation (as was done Documentation in OSC-9068).

documentation issues.

For each source and its identified failure Catawba Enhance the Internal Flood mechanism, IDENTIFY the characteristic of Flood analysis to address the release and the capacity of the source. INCLUDE:

Analysis, potential for spray, jet IF-B3 (a) a characterization of the breach, including type Partial CNC-impingement, and pipe whip Documentation No impact is expected for (e.g., leak, rupture, spray) (b) range of flow rates 1535.00 failures. Additionally, documentation issues.

(c) capacity of source (e.g., gallons of water) (d) 0058, Rev.

document how these failures the pressure and temperature of the source 0, December are included in the 1

2005 quaintification.

Page 11 of 25

ATTACHMENT 5 C

I I I:

I IV te h i t" -It~

l 4:XI I 1 For each flood area not screened out using the For those flood areas requirements under IF-BIb, IDENTIFY the SSCs addressed in the current located in each defined flood area and IF-A2) along flooding analysis, flood propagation paths that are modeled in the equipment important to internal events PRA model as being required to accident mitigation and the respond to an initiating event or whose failure associated critical flood would challenge normal plant operation, and are heights are identified.

susceptible to flood. For each identified SSC, Catawba However, given the IDENTIFY, for the purpose of determining its Flood expected increase in number Internal flood sequences susceptibility per IF-C3, its spatial location in the

Analysis, of flood areas needed -to are not significant IF-C2c area and any flooding mitigative features (e.g.,

Partial CNC-satisfy requirement IF-AI, Technical contributors in the present shielding, flood or spray capability ratings).

1535.00 additional equipment will analysis. No significant 0058, Rev.

need to be identified and impact associated with 0, December discussed in order to meet this open SR.

2005 the requirements of the ASME Standard. The current flooding analysis does not discuss flood mitigative features and this will have to be corrected to satisfy the requirements of the ASME Standard. "

For the SSCs identified in IF-C2c, IDENTIFY the The current flooding Internal flood sequences susceptibility of each SSC in a flood area to flood-Ctb analysis identifies the are not significant induced failure mechanisms. INCLUDE failure by aawo submergence failure height contributors in the present submergence and spray in the identification Alyss of the equipment important analysis. No significant process. ASSESS qualitatively the impact of flood-CNC-to accident mitigation, but impact associated with IF-C3 induced mechanisms that are not formally Partial 1535.00 never addresses the impact Technical this open SR.

addressed (e.g., using the mechanisms listed under 0058, Rev, of spray. Spray as a failure Capability Category III of this requirement), by December mechanism needs to be using conservative assumptions.

0, addressed in the analysis or 2005 a note made explaining why it was omitted.

Page 12 of 25

ATTACHMENT 5 IDENTIFY inter-area propagation through the Internal flood sequences normal flow path frlom one area to another via drain Catawba Provide more analysis of are not significant lines; and areas connected via back flow through Flood flood propagation contributors in the present drain lines involving failed check valves, pipe and

Analysis, flowpaths. Address analysis. No significant IF-C3b cable penetrations (including cable trays), doors, Partial CNC-potential structural failure of Technical impact 5.ssociated with stairwells, hatchways, and HVAC ducts.

1535.00 doors or walls due to this open SR.

INCLUDE potential for structural failure (e.g., of 0058, Rev.

flooding loads and the doors or walls) due to flooding loads and the 0, December potential for barrier potential for barrier unavailability, including 2005 unavailability.

maintenance activities.

INCLUDE, in the quantification, both the direct Catawba Internal flood sequences effects of.the flood (e.g., loss of cooling from a Flood are not significant service water train due to an associated pipe

Analysis, contributors in the present rupture) and indirect effects such as submergence, CNC-Address potential indirect analysis. No significant IF-E6b CPartial 13.00-efcsTechnical imatasoite ih jet irmpingemnent, and pipe whip, as applicable.

1535.00 effects.

impact associated with.

.0058, Rev.

this open SR.

0, December 2005 Page 13 of 25

ATTACHMENT 5 RIi I"for I N cec tedmpacton 1 1.

H D5,eslt

.:*ktocumentaXn*.,

pplication DOCUMENT the process used to identify flood sources, flood areas, flood pathways, flood scenarios, and their screening, and internal flood model development and quantification. For example, this documentation typically includes (a) flood sources identified in the analysis, rules used to screen out these sources, and the resulting list of sources to be further examined (b) flood areas used in the analysis and the reason for eliminating areas from further analysis (c) propagation pathways between flood areas and assumptions, calculations, or other bases for eliminating or justifying Catawba propagation pathways (d) accident mitigating Flood features and barriers'credited in the analysis, the Flood extent to which they were credited, and associated

Analysis, Need to document how the exett hc he

eeceie, n

soitdCNC-analysis addressed all of the No impact is expected for IF-F2 justification (e) assumptions or calculations used in Partial 550-items idenied in thi Documentation documeat isses.

th eemnto fteipcso umrec,1535.00 items identified in this documnentation issues.

the determination of the impacts of submergence,requirement.

spray, temperature, or other flood-induced effects 0, December on equipment operability (f) screening criteria used 0,

e in the analysis (g) flooding scenarios considered, screened, and retained (h) description of how the internal event analysis models were modified to model these remaining internal flooding scenarios (i) flood frequencies, component unreliabilities/unavailabilities, and HEPs used in the analysis (i.e., the data values unique to the flooding analysis) (j) calculations or other analyses used to support or refine the flooding evaluation (k) results of the internal flooding analysis, consistent with the quintification requirements provided in HLR QU-D Page 14 of 25

ATTACHMENT 5 PERFORM a containment bypass analysis in a The conservative realistic manner. JUSTIFY any credit taken for Perform plant-specific T/H treatment will not mask scrubbing (i.e., provide an engineering basis for the calculations for SGTR.

the contribution of non-decontamination factor used).

Catawba Consider some credit for bypass events, because Simplified ISLOCA scrubbing; if no even if some credit were LERF credit can be given, then this given to scrubbing, the LE-Partial Methodolog should be documented. It is Technical unscrubbed bypasses CIO y, SAAG not known whether or not would still dominate 817, Rev. 1, the additional analysis will LERF over the non-October alter the LERF, but because bypass events. In 2004 these items dominate LERF, addition, the limiting risk a more realistic analysis metric in the present should be considered.

analysis is CDF, not LERF.

In crediting HFEs that support the accident The only operator action progression analysis, USE the applicable Catawba expected to be important is requirements of para. 4.5.5, as appropriate for the Simplified RCS depressurization for level of detail of the analysis.

LERF small LOCAs. However, ofe thereanalysis.

laLERF Methodolog the current analysis lacks a No impact is expected for LE-C6 Partial y, SAAG formal dependency analysis Documentation documentation issues.

817, Rev. 1, for this action.. The result is October expected to be insensitive to 2004 this impact given that the SGTR so dominates the result.

PERFORM a realistic interfacing system failure probability analysis for the significant accident Catawba C

For MNS/CNS, the ND heat progression sequences resulting in a large early ISLOCA ISLOCA sequences are release. USE a conservative or a combination of Analysis exchanger is assumed tonot sinificant provide the largest break conservative and realistic evaluation of interfacing CNC-C contributors in the present LE-D3 system failure probability for non-significant No 1535.00 flow area. The ISLOCA is Technical significant a dominant contributor and aayi.N infcn accident progression sequences resulting in a large 0053, Rev.

impact associated with a,

Zý the evaluation is relativelythsoeSR early release. INCLUDE behavior of piping relief 0, January conservativethis open SR.

valves, pump seals, and heat exchangers at 2006 applicable temperature and pressure conditions.

Page 15 of 25

ATTACHMENT 5

, j*CteorfrCN9;* iC*R

-Expected-Irnpact on>?

PROVIDE uncertainty analysis that identifies the Catawba sources of uncertainty and includes sensitivity Simplified studies for the significant contributors to LERF.

LERF Methodolog y, SAAG 8 17, Rev. l, 8ctobev. l Perform and document 2004; sensitivity studies to LE-F2 Partial 2004; determine the impact of the No impact is expected for Rev 3a PRA assumptions and sources of documentation issues.

Rev 3aP model uncertainty on the Model LERF results.

Integration

Notebook, CNC-1535.00 0061, Rev.

2, July 2006 IDENTIFY contributors to LERF and characterize LERF uncertainties consistent with the applicable requirements of Tables 4.5.8-2(d) and 4.5.8-2(e).

NOTE: The supporting requirements in these tables Catawba are written in CDF language. Under this Simplified requirement, the applicable requirements of Table LERF Compare LERF results and LE-F3 4.5.8 should be interpreted based on LERF, Partial Methodolog uncertainties to similar Documentation No impact is expected for including characterizing key modeling uncertainties y, SAAG plants and include in the documentation issues.

associated with the applicable contributors from 817, Rev. 1, LERF documentation.

Table 4.5.9-3. For example, supporting requirement October QU-D5 addresses the significant contributors to 2004 CDF. Under this requirement, the contributors would be identified based on their contribution to LERF.

Page 16 of 25

ATTACHMENT 5 1,c m

a.or.,

SCategory H 'Requirehienit,,

NlMetN*

o*i

..' Y N R: f.eu i*D6cum1entation 4A-+.*:%+,';

-Re*<,+soliit i*

)*.;

,,,ppllcation DOCUMENT the relative contribution of Catawba contributors (i.e., plant damage states, accident Simplified progression sequences, phenomena, containment LERF Evaluate the relative challenges, containment failure modes) to LERF.

Methodolog contribution of the various Documentation No impact is expected for LE-G3 Partial y, SAAG contributors to the total documentation issues.

817, Rev. 1, LERF.

October 2004 DOCUMENT assumptions and sources of Catawba uncertainty associated with the LERF analysis, Simplified Perform and document including results and important insights from LERF sensitivity studies to LE-G4 sensitivity studies.Partial Methodolog determine the impact of the D

tti No impact is expected for y, SAAG assumptions and sources of documentation issues.

817, Rev. 1, model uncertainty on the October LERF results.

2004 IDENTIFY limitations in the LERF analysis that Include in the LERF would impact applications, documentation an LE-G5

'No assessment that identifies Documentation No impact is expected for the limitations in the LERF documentation issues.

analysis that could impact applications.

DOCUMENT the quantitative definition used for Catawba significant accident progression seqtience. If other Simplified than the definition used in Section 2, JUSTIFY the LERF Provide a discussion of the LE-G6 alternative.

Partial Methodolog significant cut sets and Documentation No impact is expected for y, SAAG C,

documentation issues.

817, Rev. 1, sequences.

October 2004 COMPARE results to those from similar plants and Perform and document a UD3comparison of resul~ts Dcmnain No impact is expected for QU-D3 IDENTIFY causes for significant differences. For No en t e

C P

Documentation example: Why is LOCA a large contributor for one between the CNS PRA and.

documentation issues.

plant and not another?

other similar plants.

Page 17 of 25

ATTACHMENT 5 II ;J4.

~t~fr;NS?

e..

~~~eouo Bocuentti~:

Expected 1pac"on S Rcategory I 'qurnti I(tf~ N?

C cS~ kO:-

Rc"idutio Ij I. c

~

t t

dT'leII EVALUATE the sensitivity of the results to model Perform and document a set uncertainties and assumptions using sensitivity of sensitivity cases to QU-E4 No determine the impact of the Documentation No impact is expected for assumptions and sources of documentation issues.

model uncertainty on the results.

Page 18 of 25

ATTACHMENT 5 DOCUMENT the model integration process, including any recovery analysis, and the results of the quantification including uncertainty and sensitivity analyses. For example, documentation typically includes (a) records of the process/results when adding nonrecovery terms as part of the final quantification (b) records of the cutset review process (c) a general description of the quantification process including accounting for systems successes, the truncation values used, how recovery and post-initiator HFEs are applied (d) the process and results for establishing the truncation Catawba screening values for final quantification Rev 3a PRA demonstrating that convergence towards a stable Model result was achieved (e) the total plant CDF and Integration of CNS delurestsN ped Cof CNS PRA model results No impact is expected -for QU-F2 contributions from the different initiating events Partial

Notebook, Documentation Ito address all required documnentation issues.

and accident classes (fl the accident sequences and CNC-items their contributing cutsets (g) equipment or human 1535.00 actions that are the key factors in causing the 0061, Rev.

accident sequences to be nonsignificant (h) the 2, July 2006 results of all sensitivity studies (i) the uncertainty distribution for the total CDF (j) importance measure results (k) a list of mutually exclusive events eliminated from the resulting cutsets and their bases for Elimination (1) asymmetries in quantitative modeling to provide application users the necessary understanding regarding why such asymmetries are present in the model (m) the process used to illustrate the computer code(s) used to perform the quantification will yield correct results process.

Page 19 of 25

ATTACHMENT 5 UULUVItLiN i me quanttative aeinmitlion usea ior significant basic event, significant cutset, significant accident sequence. If other than the definition used in Section 2, JUSTIFY the alternative.

QU-F6 Partial k.aaw Da Rev 3a PRA Model Integration

Notebook, CNC-1535.00 0061, Rev.

2, July 2006 Document the required definitions.

D a No impact is expected for DocumIentation documentation issues.

SPECIFY success criteria for each of the key safety Catawba functions identified per SR AS-A2 for each Thermal-Improve the documentation moCele iniit

[Note (2)].

ParHydraulic on the TH bases for all Documentation No impact is expected for SC-A4 m

n Partial Success safety function success doCumentation issues.

Criteria criteria for all initiators.

calcs.

CHECK the reasonableness and acceptability of the results of the thermal/hydraulic, structural, or other supporting engineering bases used to support the Catawba success criteria. Examples of methods to achieve Thermal-Provide evidence that an No impact is expected for this include: (a) comparison with results of the Hydraulic acceptability review of the Documentation SC-B5 same analyses performed for similar plants, Partial Success documentation issues.

accounting for differences in unique plant features Criteria T/H analyses is performed.

(b) comparison with results' of similar analyses calcs.

performed with other plant-specific codes (c) check by other means appropriate to the particular analysis DOCUMENT the success criteria in a manner that Catawba facilitates PRA applications, upgrades, and peer Thermal-Improve the documentation SC-Cl review.

Hydraulic on the TH bases for all Doetatio No impact is expected for Partial Success safety function success documentation issues.

Criteria criteria for all initiators.

calcs.

Page 20 of 25

ATTACHMENT 5 SC-C2 IutUIVIIN 1 me processes useo to leveiop overall PRA success criteria and the supporting engineering bases, including the inputs, methods, and results. For example, this documentation

.typically includes: (a) the definition.of core damage used in the PRA including the bases for any selected parameter value used in the definition (e.g., peak cladding temperature or reactor vessel level) (b) calculations (generic and plant-specific) or other references used to establish success criteria, and identification of cases for which they are used (c) identification of computer codes or other methods used to establish plant-specific success criteria (d) a description of the limitations (e.g., potential conservatisms or limitations that could challenge the applicability of computer models in certain cases) of the calculations or codes (e) the uses of expert judgment within the PRA, and rationale for such uses (t) a summary of success criteria for the available mitigating systems and. human actions for each accident initiating group modeled in the PRA (g) the basis for establishing the time available for human actions (h) descriptions of processes used to define success criteria for grouped initiating events or accident sequences Partial Catawba Thermal-

.Hydraulic Success Criteria calcs.

Improve the documentation on the TH bases for all safety function success criteria for all initiators.

Documentation No impact is expected for documentation issues.

Page 21 of 25

ATTACHMENT 5 In meeting SY-A12 and SY-A 13, contributors to system unavailability and unreliability (i.e.,

components and specific failure modes) may be excluded from the model if one of the following screening criteria is met: (a) A component may be excluded friom the system model if the total failure probability of the component failure.modes resulting in the same effect on system operation is SY-at least two orders of magnitude lower than the Partial System Provide quantitative Documentation No impact is expected for A 14 highest failure probability of the other components analyses evaluations for screening.

documentation issues.

in the same system train that results in the same effect on system operation. (b) One or more failure modes for a component may be excluded from the systems model if the contribution of them to the total failure rate or probability is less than I % of the total failure rate or probability for that component, when their effects on system operation are the same.

COLLECT pertinent information to ensure that the systemns analysis appropriately reflects the as-built and as-operated systems. Examples of such information include system P&IDs, one-line diagrams, instrumentation and control drawings, spatial layout drawings, systemn operating spatial laotsytmoprtn Systemn Need to update references Dt No impact is expected for SY-A2 procedures, abnormal operating procedures, Partial Documentation ise abngnporm operi ceres analyses per XSAA-115.

documentation issues.

emergency procedures, success criteria calculations, the final or updated SAR, Technical Specifications, training information, system descriptions and related design documents, actual system operating experience, and interviews with system engineers and operators.

Page 22 of 25

ATTACHMENT 5

~SR 1X

.Category H

1 Require'i n ent I c

)rC:

CNS Ref.

l- 'Reoluition cun

'ExectdinnaW:n PERFORM plant walkdowns and interviews with Enhance the system system engineers and plant operators to confirm documentation to include an that the systems analysis correctly reflects the as-up-to-date system built, as-operated plant.

walkdown checklist and system engineer review for SY-A4 Partial System each system. Consider Documentation No impact is expected for analyses revising workplace documentation issues.

procedure XSAA-106 to require that such documentation be revisited with each major PRA revision.

ESTABLISH the boundaries of the components required-for system operation. MATCH the definitions used to establish the component failure data. For example, a control circuit for a pump does not need to be included as a separate basic event (or events) in the system model if the pump failure data Enhance systems analysis No is for SYS used in quantifying the systemn model include No impact isexpectedfo SY-A8 usdC uniyn hesse oe nld No documentation to discuss Documentation control circuit failures. MODEL as separate basic component boundaries documentation issues.

events of the model, those subcomponents (e.g., a valve limit switch that is associated with a permissive signal for another component) that are shared by another component or affect another component, in order to account for the dependent failure mechanism.

Page 23 of 25

ATTACHMENT 5 SY-B15 IDENTIFY SSCs that may be required to operate in conditions beyond their environmental qualifications. INCLUDE dependent failures of multiple SSCs that result from operation in these adverse conditions. Examples of degraded environments include: (a) LOCA inside containment with failure of containment heat removal (b) safety relief valve operability (small LOCA, drywell spray, severe accident) (for BWRs)

(c) steam line breaks outside containment (d) debris that could plug screens/filters (both internal and external to the plant) (e) heating of the water supply.

(e.g., BWR suppression pool, PWR containment sump) that could affect pump operability (f) loss of NPSH for pumps (g) steam binding of pumps (h) harsh environments induced by containment venting or failure that may occur prior to the onset of core damage Partial System analyses Cut set review during applications should address this. Suggest adding this guidance to workplace procedure XSAA-103.

Documentation No impact is expected for documentation issues.

+

I I

IDENTIFY spatial and environmental hazards that may impact multiple systems or redundant components in the same system, and ACCOUNT for them in the system fault tree or the accident sequence evaluation. Example: Use results of plant walkdowns as a source of information regarding spatial/environmental hazards, for resolution of spatial/environmental issues, or evaluation of the impacts of such hazards.

SY-B8 Partial System analyses Per Duke's PRA modeling guidelines, ensure that a walkdown/system engineer interview checklist is included in each system notebook. Based on the results of the system walkdown, summarize in the system write-up any possible spatial dependencies or environmental hazards that may impact system operation.

Documentation No impact is expected for documentation issues.

DOCUMENT the system functions and boundary, System Enhance system model No impact is expected for SY-C2 the associated success criteria, the modeled Partial menhane te mo lDocumentation documentationtocompyiDocumenlanalyses documentation issues.

components and failure modes including human with all ASME PRA Page 24 of 25

ATTACHMENT 5 11 Reqirmcl, leif r C'NS' C Sef4

?

actions, and a description of modeled dependencies including support system and common cause failures, including the inputs, methods, and results.

For example, this documentation typically includes:

(a) system function and operation under normal and emergency operations (b) system model boundary (c) system schematic illustrating all equipment and components necessary for system operation (d) information and calculations to support equipment operability considerations and assumptions (e) actual operational history indicating any past problems in the system operation (f) system success criteria and relationship to accident sequence models (g) human actions necessary for operation of system (h) reference to system-related test and maintenance procedures (i) system dependencies and shared component interface (j) component

'spatial information (k) assumptions or simplifications made in development of the system models (1) the components and failure modes included in the model and justification for any exclusion of components and failure modes (m) a description of the modularization process (if used)

(n) records of resolution of logic loops developed during fault tree linking (if used) (o) results of the system model evaluations (p) results of sensitivity studies (if used) (q) the sources of the above information (e.g., completed checklist from walkdowns, notes from discussions with plant personnel) (r) basic events in the system fault trees so that they are traceable to modules and to cutsets.

(s) the nomenclature used in the system models.

Standard requirements.

Page 25 of 25

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