ML033510109

Duke Energy Corporation - Response to NRC Generic Letter 2003-01, Control Room Habitability

ML033510109
Person / Time
Site:	Oconee, Mcguire, Catawba, McGuire
Issue date:	12/09/2003
From:	Mccollum W Duke Energy Corp
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
GL-03-001
Download: ML033510109 (35)

Text

• Duke Duke Energy Corporation 526 South Church Street Energy.

P.O. Box 1006 Charlotte, NC 28201-1006 December 9, 2003 U. S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 ATTENTION: Document Control Desk

SUBJECT:

Duke Energy Corporation Oconee Units 1, 2, & 3 Docket Nos. 50-269, 50-270, 50-287 Catawba Nuclear Station, Units 1 & 2 Docket Nos. 50413, 50414 Response to NRC Generic Letter 2003-01, Control Room Habitability On June 12, 2003, the NRC issued Generic Letter (GL) 2003-01 concerning control room habitability. The NRC issued the Generic Letter to alert addressees to findings at U. S. power reactor facilities suggesting that the control room licensing and design bases, and applicable regulatory requirements may not be met, and that Technical Specification surveillance requirements may not be adequate. Also, the Generic Letter emphasizes the importance of reliable, comprehensive surveillance testing to verify control room habitability.

The Generic Letter requests information to confirm that the facility's control room meets applicable control room habitability requirements, with emphasis to be placed on control room envelope inleakage testing and radiological analyses, hazardous chemical and smoke assessments, and Technical Specifications. The Generic Letter also requests information regarding compensatory measures in use and the design criteria to which the facility is licensed, if not licensed to the General Design Criteria.

Duke Energy Corporation's 180-day response to GL 2003-01 is provided in Attachments 1 and 2 for Oconee and Catawba Nuclear Stations, respectively. Duke previously submitted a 60-day response for McGuire Nuclear Station on August 7, 2003.

No commitments are contained in this letter. If you have questions or need additional information, please contact R. K. Nader at 704-382-0979.

Very truly yours, W. R. McCollum, Jr.

Senior Vice President, Nuclear Support Attachments

1;. S. Nuclear Regulatory Commission December 9, 2003 Page 2 I affirm that I, W. R. McCollum, am the person who subscribed my name to the foregoing, and that all the matters and facts herein are true and correct to the best of my knowledge.

Senior Vice President, Nucle

/RlI (cI,(Iet/ 1-osI Subscribed and sworn to me:

I Z/q/o 3 Date mA4t W L -

Notary Public My Commission Expires:

Adz Z2j 2&og Date SEAL N

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*7 MICHAEL T. CASH Notary Public Lincoln County, North Carolina Commission Expires January 22, 2008

IA. S. Nuclear Regulatory Commission December 9, 2003 Page 3 xc wlatt:

L. A. Reyes, Regional Administrator U. S. Nuclear Regulatory Commission, Region II Sam Nunn Atlanta Federal Center 23T85 61 Forsyth St., SW Atlanta, GA 30303 L. N. Olshan (Addressee only)

NRC Senior Project Manager (ONS)

U. S. Nuclear Regulatory Commission Mail Stop 0-8 H12 Washington, DC 20555-0001 S. E. Peters (Addressee only)

NRC Senior Project Manager (CNS)

U. S. Nuclear Regulatory Commission Mail Stop 0-8 H12 Washington, DC 20555-0001 M. E. Shannon, Senior Resident Inspector (ONS)

E. F. Guthrie, NRC Senior Resident Inspector (CNS)

U. S. Nuclear Regulatory Commission December 9, 2003 Page 4 bxc w/ att:

M. T. Cash (ECO50)

R. L. Gill (EC05P)

Lee A. Keller (CNO1RC)

K. E. Nicholson (CNOIRC)

A. P. Jackson (CNOIRC)

R. E. Banker (CN03SE)

J. F. Reed (CN03SE)

J. M. Ferguson - Date File (CNOISA)

L. E. Nicholson (ON03RC)

J. E. Smith (ON03RC)

R. V. Gambrell (ON03RC)

J. N. Robertson (ON03MC)

R. H. Burley (ON03MC)

H. T. Barrett (ON03MC)

S. P. Schultz (EC09A)

C. B. Taylor (EC09A)

M. V. Costello (EC09A)

MNS MasterFile MC-801.01 (MGOIDM)

CNS MasterFile CN-801.01 (CN04DM)

ONS MasterFile ON-801.01 (ON03DM)

ELL

U. S. Nuclear Regulatory Commission December 9, 2003 Page 5 bxc w/ attach: North Carolina Municipal Power Agency Number 1 1427 Meadowwood Boulevard P. 0. Box 29513 Raleigh, NC 27626 Saluda River Electric Cooperative, Inc.

P. 0. Box 929 Laurens, SC 29360 Piedmont Municipal Power Agency 121 Village Drive Greer, SC 29651 T. Richard Puryear North Carolina Electric Membership Corporation P. 0. Box 27306 Raleigh, NC 2761!

Oconee Nuclear Station's Response to Generic Letter 2003-01

OCONEE NUCLEAR STATION'S RESPONSE TO GL 2003-01 Per the above-referenced generic letter, licensees were requested to provide information on three issues relating to the importance of ensuring that the control room envelope is operated and maintained in accordance with the facility's design and licensing bases.

The exact requests and the specific Oconee responses follow:

1.

Provide confirmation that your facility's control room meets the applicable habitability regulatory requirements (e.g., GDC 1, 3, 4, 5, and 19) and that the CRHSs are designed, constructed, configured, operated, and maintained in accordance with the facility's design and licensing bases.

Oconee Nuclear Station Units 1, 2, and 3 were not designed to the General Design Criteria (GDC) currently outlined in 10CFR50 Appendix A. Howevcr, Oconee was designed based on consideration of the seventy General Design Criteria for Nuclear Power Plant Construction Permits proposed by the AEC in the Federal Register of July 11, 1967. Several of the AEC General Design Criteria are comparable to the GDCs outlined in 10CFR50 Appendix A. A detailed comparison between the Oconee level of commitment to the applicable AEC criteria and the IOCFR50 Appendix A criteria 1, 3, 4, 5, and 19 is given in the response to Question 3.

The Control Room for Oconee Units I and 2 is shared for the operation of both units. A separate control room exists for the operation of Oconee Unit 3. The control room envelope includes the control room and other rooms that may require operator access during an emergency such as offices, computer rooms, the break area, and the toilet.

Supporting systems, structures, and components include the control room envelope structure, the doors, the Control Room Ventilation System (CRVS), and the Control Room Area Cooling System (CRACS). The CRVS provides filtered air to the control room and is used to pressurize the control room in the event of an accident. CRACS is used to cool the control room, the cable room, and the equipment room. The operation of these systems is controlled by Technical Specification (TS) 3.7.9, "Control Room Ventilation System (CRVS) Booster Fans" and TS 3.7.16, "Control Room Area Cooling Systems". Testing of the CRVS filters is controlled by TS 5.5.12, "Ventilation Filter Testing Program" and Selected Licensee Commitment 16.15, "Ventilation Filter Testing Program".

The performance of the CRVS and CRACS is monitored in accordance with Maintenance Rule requirements. System Health Reports are developed periodically and reviewed by senior station management.

As demonstrated with the responses to la, lb, and Ic that follow, the CRVS and CRACS systems are designed, constructed, configured, operated, and maintained in accordance with Oconee's design and licensing basis, with the exception of issues noted in the responses to Questions la and Ic below and in the response to Question 3.

Attachment I December 9, 2003 Page 2 of 17 Emphasis should be placed on confirming:

(a)

That the most limiting unfiltered inleakage into your CRE (and the filtered inleakage if applicable) is no more than the value assumed in your design basis radiological analyses for CRE habitability. Describe how and when you performed the analyses, tests, and measurements for this confirmation.

Tracer gas tests were performed at Oconee in 1998 and 2001 to confirm that the inleakage into the control room envelope is less than or equal to the value assumed in the plant radiological analysis. Tracer gas tests were performed in accordance with ASTM E741 by NCS Corporation. The written test report that was provided by NCS to Oconee detailing the methodology and results of the tracer gas test was submitted to the staff in May 2002 (Reference 2). Both Oconee control rooms were tested in three different modes of operation. The first test mode was normal operation with the booster fans off, the second test mode was the emergency mode with one booster fan operating and the third test mode was the emergency mode with both booster fans operating.

The results from the 1998 tracer gas test are as follows:

Control Room Ventilation Mode Inleakage U1IU2 NORMAL 1065 +/- 61 (ACFM)

Ul/U2 EMERGENCY - 1 FAN 80 +/- 55 (SCFM)*

UlIU2 EMERGENCY - 2 FAN 0-128 (SCFM)*

U3 NORMAL 534 +/- 30 (ACFM)

U3 EMERGENCY - 1 FAN 73 +/- 25 (SCFM)*

U3 EMERGENCY -2 FAN 0-236 (SCFM)*

• Referred to 70 Deg F and 14.7 psia A second test was performed in 2001 to verify that control room envelope integrity is being maintained. Between the 1998 test and the 2001 test duct sealing was performed and cable penetrations in the control room boundary were improved. The tracer gas injection location for the 2001 test was changed to promote better mixing, resulting in more accurate measurements with lower uncertainty.

Attachment I December 9, 2003 Page 3 of 17 The results from the 2001 tracer gas test are as follows:

Control Room Alignment Inleakage U1/U2 NORMAL 869+/-31 (ACFM)

U1IU2 EMERGENCY -1 FAN 0 +/-18 (SCFM)*

U1/U2 EMERGENCY - 2 FAN 0 +/-30 (SCFM)*

U3 NORMAL 467 +/- 16 (ACFM)

U3 EMERGENCY - 1 FAN 0 +-13 (SCFM)*

U3 EMERGENCY - 2 FAN 0 +-42 (SCFM)*

• Referred to 70 Deg F and 14.7 psia Oconee has a condition in its licensing basis radiological analysis that is currently operable but nonconforming. The operability evaluation is based on the 1998 test results that indicated the air inleakage is higher than that used in the licensing basis analysis. The control room air inleakage values used in the operability evaluation are conservative in that the values used are much greater than the most recent, 2001 test values.

In October 2001 Duke Energy Corporation submitted a license amendment request which includes full implementation of Alternative Source Terms for Oconee Nuclear Station.

This license amendment request includes an analysis of the radiological consequences of a design basis Loss of Coolant Accident (LOCA) and Fuel Handling Accident (FHA) postulated to occur at Oconee (Reference 1 through Reference 7).

The selection of bounding values for the control room unfiltered inleakage assumed in the analyses provides Duke with margin to accommodate changes in input assumptions that could be required to account for possible plant operational changes, such as increases in ECCS system leakage flow, imbalances in ventilation system flowrates, or reductions in filtration efficiencies. Analyses are performed to support a method designed to couple the evaluation of control room dose for control room unfiltered inleakage and ECCS leakage so that the control programs for the system performance can be designed using the margin tradeoff between these two parameters.

Duke has concluded that the appropriate input values for unfiltered inleakage as derived from the tracer gas test results should correspond to the nominal values determined from each of the testing programs. This conclusion is valid because the uncertainty values derived from the experimental results are within a reasonable range, as seen in the data set measurement results. Additionally, the measured nominal values for leakage during December 9, 2003 Page 4 of 17 operation of the control room booster fan pressurization system are very low and much less than 100 cfm.

Sensitivity studies have shown that the dose prediction is most sensitive to the post-booster fan value (after 30 minutes into the accident). Therefore, to accommodate operational flexibility for ECCS system leakage, a range of values for unfiltered inleakage in the post-booster fan configuration is used. Post-booster fan inleakage values ranging from 40 cfm to 90 cfm are evaluated. This range of values provides margin above the 2001 tracer gas test results of 0 +/- 18 cfm inleakage for the Units 1&2 control room. A bounding value of 1150 cfm is selected for the pre-booster fan flowrate and is applied for the first 30 minutes following the event.

Atmospheric dispersion factors for transport of radioactivity to the intakes of the Control Room Ventilation System (CRVS) have been calculated. All control room X/Q values for transport of radioactivity with dispersion from one release point to one CRVS outside air intake were calculated with the computer code ARCON96. The control room X/Q values used in the analysis of the design basis Fuel Handling Accident and Loss of Coolant Accident have been calculated in conformance to the regulatory positions of Regulatory Guide (RG) 1.194. Oconee Nuclear Station will have the dual intake configuration upon completion of plant modifications as described in References 1 and 2. As appropriate, composite control room X/Q values for transport of radioactivity with dispersion to both CRVS outside air intakes were calculated. The asymmetry in airflow in the two CRVS outside air intakes was set to a value of 55/45. This asymmetry will be verified following the modification to install the dual control room intake system.

Calculations of the design basis Fuel Handling Accidents were performed in conformance to the regulatory positions taken in RG 1.183. The associated license amendment requests proposed changes to the Technical Specifications for the Penetration Room Ventilation System (PRVS) and Spent Fuel Pool Ventilation System (SFPVS).

The Technical Specifications for the PRVS are deleted since no credit is taken for PRVS operation in any analyses. The SFPVS Technical Specification is changed to be applicable only during movement of recently irradiated fuel assemblies in the spent fuel pool (SFP). Control room dose calculations were performed for limiting fuel assemblies that are not recently irradiated and assume no credit for these ventilation systems. The limiting Total Equivalent Dose Equivalent (TEDE) to the control room operators for these accidents was calculated to be 3.4 rem following a transport cask drop in the Unit 1

& 2 SFP, using a conservative 60/40 control room intake flow imbalance. The control room TEDE for the limiting single assembly Fuel Handling Accident (postulated damage to one fuel assembly) was calculated to be 2.2 rem.

The analysis of radiological consequences of the design basis LOCA, and in particular the calculation of post-LOCA radiation doses to the control room operators was performed generally in conformance to the regulatory positions of RG 1.195. A detailed report of the analysis and comparison with RG 1.183 and Appendix A has been submitted to the NRC Staff (Reference 1 through Reference 7).

Attachment I December 9, 2003 Page 5 of 17 The LBLOCA dose analysis has been performed assuming a 55/45 dual control room intake flow imbalance, and using a range of potential ECCS leakage values, with corresponding assumed control room unfiltered inleakage values that demonstrate doses within regulatory limits for a range of parameter combinations. A control room dose in the range of 4.5 rem TEDE was chosen for a target value to provide margin to the regulatory limit of 5.0 rem TEDE. The sensitivity of the results to these key input parameters for four representative cases is as follows.

Case 1 Case 2 Case 3 Case 4 Base Control Room Unfiltered Inleakage 40 cfm 60 cfm 80 cfm 90 cfm (post-booster fan actuation) (cfm)

Total ECCS Leakage Allowable 25 gph 17.5 gph 12.5 gph 10 gph (gallons/hour)

• Control Room - Containment Model 1.3 1.6 1.9 2.1 Control Room - RBES Model 3.1 3.0 2.7 2.3 Total Control Room Dose (rem TEDE) 4.4 4.6 4.6 4.4

• The ECCS leakage used in the dose calculation is twice the allowable value.

For control room dose, a value of 40 cfm post-booster fan control room unfiltered inleakage will be used as the current licensing basis value based on the most recent tracer gas test results (performed in 2001). For the next tracer gas test planned in the Control Room Habitability Program, the Oconee program for monitoring and controlling total ECCS leakage will determine the current measured total ECCS leakage in the system.

This ECCS leakage value will be used to establish the post-booster fan actuation control room inleakage test criteria for the tracer gas test. For example, if total ECCS leakage is measured to be less than 10 gph, a tracer gas test criterion of 90 cfm for post-booster fan control room inleakage will be used. A value of 1150 cfm will remain the pre-booster fan actuation control room unfiltered inleakage licensing basis value. As long as the control room inleakage test satisfies these tracer gas test criteria, control room inleakage and ECCS leakage performance is satisfactory and no past or current operability evaluation or reportability is required.

If the test value for control room unfiltered inleakage has changed, a new value of total ECCS leakage will be determined from the approved curve of total ECCS leakage versus control room unfiltered inleakage. This value will apply until the subsequent control room test. The intention of both the Control Room Habitability and ECCS Leakage Control Programs is to maintain leakage rates at low levels.

(b)

That the most limiting unfiltered inleakage into your CRE is incorporated into your hazardous chemical assessment. This inleakage may differ from the value assumed in your design basis radiological analyses. Also confirm that

Attachment I December 9, 2003 Page 6 of 17 the reactor control capability is maintained from either the control room or the alternate shutdown panel in the event of smoke.

The following sections describe the hazardous chemical assessment and smoke evaluation for Oconee.

Hazardous Chemical Assessment An evaluation of onsite and offsite hazardous chemicals is documented in site controlled calculations that are periodically reviewed for changes that may require revision. The evaluation accounts for the most limiting unfiltered inleakage into the Oconee control room envelope, which occurs during normal operation with the booster fans OFF. This value is 1065 cfm for the U1/2 control room and 534 cfm for the U3 control room from the 1998 tracer gas test. The on-site hazardous chemical assessment is performed using the progressive screening steps and control room habitability evaluation guidance provided in RG 1.78 with the changes recommended in NUREG/CR-6624. These changes include the use of the computer code HABIT, Version 1.1 and its modules EXTRAN to estimate the atmospheric dilution from the release of a toxic constituent and CHEM to determine the respective concentration build up of the toxic constituent within the control room. The conclusion of the analysis is that on-site releases of hazardous chemicals do not result in concentrations within the control rooms that are unacceptable.

This on-site hazardous chemical assessment is presently being revised to evaluate the impact of the proposed control room dual air intake system. Preliminarily, results are that concentrations of hazardous chemicals within the control room are acceptable with the revised control room air intake locations and operation of the booster fans during a hazardous chemical release.

Oconee is located in a non-industrial area with no significant highway or river traffic.

The off-site hazardous chemical assessment concludes that there are no potential off-site fixed sources of toxic gas that adversely affect control room habitability.

Smoke Evaluation The Standby Shutdown Facility (SSF) provides a secondary alternate and independent means to achieve and maintain a hot shutdown condition for scenarios in which the Control Room is unavailable or equipment it controls is unavailable. The SSF was designed for safe shutdown during postulated fire, Turbine Building flooding, and physical security events.

The SSF is housed in a separate structure that is physically removed from the auxiliary building and control room. The SSF has its own ventilation system that is completely independent from any of the plant ventilation systems.

Attachment I December 9, 2003 Page 7 of 17 The following sections address smoke design, egress pathways, and smoke control to assure reactor control capability is maintained in the event of smoke.

Smoke Design The strategy used to assure the capability to safely shut the Oconee units down in the event of a fire that results in evacuation of the control room relies on the Standby Shutdown Facility (SSF). For the majority of fires that can potentially affect safe shutdown, the SSF is the credited safe shutdown train.

Significant physical separation exists both laterally and vertically between the Control Room Ventilation System outside air intakes and the SSF Control Room Ventilation System outside air intake. In addition, there is also significant separation from the point of discharge from the station exhaust points (Turbine Building Exhaust Fan Discharges and Station Ventilation Stack) to the SSF Ventilation System outside air intake.

In the unlikely event that a station fire was to generate significant smoke that entered the outside air intake of the Control Room, a contingency action would be to have the essential operators in the affected Control Room don Self-Contained Breathing Apparatus (SCBA). An operations management procedure requires that all fire brigade and control room personnel, including oversight personnel (e.g. STA, Plant SRO), maintain training and qualification for using a self-contained breathing apparatus (SCBA).

Egress Pathways Evaluation There are 3 primary fire-rated stair towers that are to be used for egress to the SSF with multiple alternate protected paths available should there be a need to bypass a fire or obstruction in any given area.

The egress path from the control rooms to the SSF is an operator training Job Performance Measure (JPM). Job Performance Measures verify, once per Operator Requalification Cycle, that all licensed operators can properly man the SSF. This JPM is a required part of both the initial and continuing requalification training.

Based on the numerous available alternate pathways the operators could use to get from either of the control rooms to the SSF, a fire in the Auxiliary Building is not expected to significantly challenge safe shutdown.

Smoke Control Oconee has developed detailed Fire Pre-Plans that include smoke control actions to be taken for each fire area/zone. Smoke control focuses on the use of portable blowers and flexible duct. In addition to the Fire Pre-Plans, smoke control activities are covered in the Fire Brigade training. The smoke control equipment is stored in the Fire Brigade lockers/equipment laydown area. The smoke control equipment is maintained in a ready state as part of the Fire Brigade equipment inventory control process.

Attachment I December 9,2003 Page 8 of 17 Overall Smoke Evaluation Conclusion Based on the fact that Oconee relies on the SSF as a secondary alternate and independent means to achieve and maintain a hot shutdown condition for scenarios in which the Control Room is unavailable or equipment it controls is unavailable and the physical layout of the site structures and the numerous egress pathways available for operators to reach the SSF, there is adequate assurance that safe shutdown will be achieved during a smoke event originating inside or outside the control room.

(c)

That your technical specifications verify the integrity of the CRE, and the assumed inleakage rates of potentially contaminated air. If you currently have a AP surveillance requirement to demonstrate CRE integrity, provide the basis for your conclusion that it remains adequate to demonstrate CRLE integrity in light of the ASTM E741 testing results. If you conclude that your AP surveillance requirement is no longer adequate, provide a schedule for: 1) revising the surveillance requirement in your technical specification to reference an acceptable surveillance methodology (e.g., ASTM E-741), and 2) making any necessary modifications to your CRE so that compliance with your new surveillance requirement can be demonstrated.

If your facility does not currently have a technical specification surveillance requirement for your CRE, explain how and on what frequency you confirm your CRE integrity.

Currently, Oconee Technical Specification Surveillance Requirement SR 3.7.9.3 periodically verifies the positive pressurization of the control room with respect to adjacent areas. Duke plans to implement TSTF-448, "Control Room Habitability," on a reasonable schedule once the TSTF is made available for use. The industry submitted TSTF-448, "Control Room Habitability," to the NRC for review on August 19, 2003.

This TSTF will follow the standard TSTF process and schedule.

No modifications to the control room envelope were required to perform the tracer gas tests. The Oconee response to Question la documents the results of tracer tests performed in 1998 and 2001. The results of these tests indicate that the work to seal ducts and electrical penetrations between 1998 and 2001 was very effective in reducing inleakage. In fact, the nominal value of the inleakage during the emergency mode was zero during the 2001 test. The results of the dose analysis are most sensitive to this mode.

Modifications to improve dose analysis results are described in the referenced submittals in response to Question 3.

2.

If you currently use compensatory measures to demonstrate CRE habitability, describe the compensatory measures at your facility and the corrective actions needed to retire these compensatory measures Oconee does not rely on compensatory measures to demonstrate control room envelope habitability.

Attachment I December 9,2003 Page 9 of 17

3.

If you believe that your facility is not required to meet either the GDC, the draft GDC, or the "Principle Design Criteria" regarding control room habitability, in addition to responding to items 1 and 2 above, provide the documentation (e.g., Preliminary Safety Analysis Report, Final Safety Analysis Report sections, or correspondence, etc.) of the basis for this conclusion and identify your actual requirements.

Oconee Units 1, 2, and 3 were designed in consideration of the seventy General Design Criteria for Nuclear Power Plant Construction Permits proposed by the AEC in a proposed rule-making published for 10 CFR Part 50 in the Federal Register of July 11, 1967. The AEC criteria and Oconee's level of commitment to the AEC criteria is described in UFSAR section 3.1.1. A detailed comparison between the applicable AEC criteria and the 10CFR50 Appendix A criteria 1, 3, 4, 5, and 19 and the current level of Oconee commitment to the AEC criteria is given in Table 1.

Prior to the issuance of the subject generic letter, Oconee identified areas where it may not currently conform to all aspects of the licensing basis outlined in Table 1. Oconee has several initiatives underway to return to a conforming status. A brief summary of each initiative and the related correspondence to the NRC follows:

Control Room Inleakage Results from the tracer gas tests indicate that the control room inleakage used in the dose analyses does not bound actual values. A license amendment request (Reference 1 through Reference 7) has been submitted for NRC review and approval that uses the Alternate Source Term methodology, bounding values for control room inleakage and ECCS post LOCA recirculation leakage, more restrictive Technical Specification values for containment leakage and CRVS iodine removal, and a modified plant configuration.

Tornado Protection In order to better understand the risks associated with tornadoes, a study was initiated in 2001 to evaluate tornadoes on a plant-wide basis. This study determined that a small portion of the Unit 3 control room north wall is not designed for tornado loads. Although analysis has shown that the wall is expected to survive direct wind loads up to 300 mph, it will not withstand the maximum negative pressure loads expected from a design basis tornado. The wall also does not have the ability to stop certain missiles. A Tornado Missile (TORMIS) evaluation has been completed and results have shown that the north wall area is a very low missile strike risk. Utilization of the TORMIS methodology and results of the recent evaluation are currently being incorporated into appropriate sections of the UFSAR. TORMIS is used to perform a Monte Carlo simulation that accounts for the geometry of important systems and structures and the locations of items near the plant that could become missiles in the event of a strike by a severe tomado. TORMIS, with some qualifications, has been found to be acceptable by the NRC and has been used by other utilities (Reference 8 through Reference 11) to demonstrate that exceptions to the December 9, 2003 Page I of 17 plant design basis are reasonable from a risk-informed perspective. The NRC recently performed a supplemental inspection related to tornado mitigation and closed the associated white finding (Reference 12).

High Energy Line Break (HELB)

Oconee has initiated a study to verify its compliance with various HELB licensing basis requirements. Although the control room envelope is protected from HELBs by the structure and HELB doors, the impact on supporting systems will be studied concurrent with the initiating event. The program used to conduct the study is outlined in Reference 13.

REFERENCES

1. W. R. McCollum (Duke Energy) to U.S. Nuclear Regulatory Commission, "License Amendment Request for Full-Scope Implementation of Alternate Source Term and Technical Specifications 3.3.5, Engineered Safeguards Protective System (ESPS)

Analog Instrumentation; 3.3.6, Engineered Safeguards Protective System (ESPS)

Manual Initiation; 3.3.7, Engineered Safeguards Protective System (ESPS) Digital Automatic Actuation Logic Channels; 3.7.10, Penetration Room Ventilation System; 3.7.17, Spent Fuel Pool Ventilation System; 3.9.3, Containment Penetrations; 5.5.2, Containment Leakage Rate Testing Program; and 5.5.12, Ventilation Filter Testing Program; Technical Specification Change Number 01-07," October 16, 2001.

2. W. R. McCollum to U.S. Nuclear Regulatory Commission, "Supplement to License Amendment Request for Full-Scope Implementati3n of the Alternate Source Term Technical Specification (TSC) Change Number 2001-07," May 20, 2002.

3. W. R. McCollum to U.S. Nuclear Regulatory Commission, "Supplement to License Amendment Request for Full-Scope Implementation of the Alternate Source Term Technical Specification (TSC) Change Number 2001-07," September 12, 2002.

4. R. A. Jones (Duke Energy) to U.S. Nuclear Regulatory Commission, "Supplement to License Amendment Request for Full-Scope Implementation of the Alternate Source Term Technical Specification (TSC) Change Number 2001-07," November 21, 2002.

5. Request for Additional Information from the U.S. Nuclear Regulatory Commission related to removing the Penetration Room Ventilation System and Spent Fuel Pool Ventilation System from Technical Specifications, January 27, 2003.

6. R. A. Jones to U.S. Nuclear Regulatory Commission, "Supplement to License Amendment Request for Full-Scope Implementation of the Alternate Source Term Technical Specification Change (TSC) Number 2001-007," September 22, 2003.

7. R. A. Jones to U. S. Nuclear Regulatory Commission, "Supplement to License Amendment Request for Full-Scope Alternate Source Term Technical Specification Change (TSC) Number 2001-007," November 20. 2003.

8. C. M. Dugger (Entergy Operations, Inc.) to U.S. Nuclear Regulatory Commission, "Request for Review and Approval of Design Basis Change Regarding Tornado Missiles", October 29, 1999. (Accession Number ML993080204)

9. R. P. Powers (Indiana Michigan Power Company) to U.S. Nuclear Regulatory Commission, "License Amendment Request to Permit Use of Probabilistic Risk

Attachment I December 9, 2003 Page 11 of 17 Assessment Techniques to Evaluate the Need for Tornado-Generated Missile Barriers", June 8, 2000. (Accession Number ML003723707)

10. U.S. Nuclear Regulatory Commission to L. W. Myers (Centerior Service Company),

"Amendment No. 90 to Facility Operating License No. NPF Perry Nuclear Power Plant, Unit 1 (TAC No. M99447)", SER to issue an amendment to change the design basis as described in the UFSAR by adding a description of the TORMIS methodology, November 4, 1997. (Accession Number ML021840234)

11. U.S. Nuclear Regulatory Commission to D. N. Morey (Southern Nuclear Operating Company), "Joseph M. Farley Nuclear Plant, Units 1 and 2 Re: Issuance of Amendments (TAC Nos. MA9495 and MA9496)", SER for the application of the TORMIS methodology for tornado missile risk analysis, September 26,2001.

(Accession Number ML012740299)

12. Letter EA-00-137 from the U.S. Nuclear Regulatory Commission's V.M. McCree to W. R. McCollum (Duke Energy), "ONSNRC Supplemental Inspection Report 50-269/02-07, 50-270/02-07, and 50-287/02-07," May 31, 2002.

13. R. A. Jones to U.S. Nuclear Regulatory Commission, "High Energy Line Break Outside Reactor Building," August 20, 2003.

December 9, 2003 Page 12 of 17 Table 1 Comparison bet een GDC, AEC Criteria and Oconee Licensing Basis Item 10CFR 50 Appendix A GDC Comparable AEC Criteria Current Oconee Licensing Basis No.

I _

Criterion I - Quality standards and records Criterion 1 - Quality Standards Those The QA-1 program implements the systems and components of reactor facilities requirements of IOCFR50 Appendix B.

Structures, systems, and components important to which are essential to the prevention of UFSAR Section 3.1.1 lists the "essential safety shall be designed, fabricated, erected, and accidents which could affect the public health systems and components" controlled under the tested to quality standards commensurate with the and safety or to mitigation of their QA-1 program. The control room structure is importance of the safety functions to be performed.

consequences shall be identified and then QA-1; however, CRVS and CRACs are not Where generally recognized codes and standards designed, fabricated, and erected to quality QA-1.

are used, they shall be identified and evaluated to standards that reflect the importance of the determine their applicability, adequacy, and safety function to be performed. Where Duke's quality assurance program conforms with sufficiency and shall be supplemented or modified generally recognized codes or standards on the requirements of 10CFR50, Appendix B, as necessary to assure a quality product in keeping design, materials, fabrication, and inspection Quality Assurance Criteria for Nuclear Power with the required safety function. A quality are used, they shall be identified. Where Plants. This Quality Assurance program is assurance program shall be established and adherence to such codes or standards does not described in Chapter 17 of the Oconee UFSAR.

implemented in order to provide adequate assurance suffice to assure a quality product in keeping Included in this quality assurance program is that these structures, systems, and components will with the safety function, they shall be specific direction for the maintenance of satisfactorily perform their safety functions.

supplemented or modified as necessary.

appropriate records.

Appropriate records of the design, fabrication, Quality assurance programs, test procedures, erection, and testing of structures, systems, and and inspection acceptance levels to be used Oconee has voluntarily adopted a QA-5 components important to safety shall be maintained shall be identified. A showing of sufficiency program to enhance the reliability of important by or under the control of the nuclear power unit and applicability of codes, standards, quality non-QA-1 equipment. This program is defined licensee throughout the life of the unit.

assurance programs, test procedures, and in the QA Topical Report. CRVS and CRACs inspection acceptance levels used is required.

are included in the QA-5 program.

Criterion 5-Records Requirements Records of the design, fabrication, and construction of UFSAR Section 3.1.1 states, "Duke Power essential components of the plant shall be Company will have under its control or will maintained by the reactor operator or under his have access to all records of major essential control throughout the life of the reactor.

components for the life of the plant.

2 Criterion 3 - Fire protection Criterion 3 - Fire Protection The reactor FSAR Section 9.5.1.6.2 identifies the Control facility shall be designed: 1) to minimize the Room as being isolated from other areas of the Structures, systems, and components important to probability of events such as fires and plant by 3 hr fire barriers, except for two 1+

safety shall be designed and located to minimize, explosions and, 2) to minimize the potential hour fire doors at stairways and a steel plate

Attachment I December 9, 2003 Page 13 of 17 Table 1 Comparison between GDC, AEC Criteria and Oconee Licensing Basis Item 10CFR 50 Appendix A GDC Comparable AEC Criteria Current Oconee Licensing Basis No.

consistent with other safety requirements, the effects of such events to safety.

adjacent to the lobby around the Control Room probability and effect of fires and explosions.

Noncombustible and fire-resistant materials entrance door. These have been reviewed and Noncombustible and heat resistant materials shall shall be used whenever practical throughout the found acceptable.

be used wherever practical throughout the unit, facility, particularly in areas containing critical particularly in locations such as the containment portions of the facility such as containment, UFSAR Section 3.1.1 states, "The reactor and control room. Fire detection and fighting control room, and components of engineered facility is designed to minimize the probability systems of appropriate capacity and capability shall safety features.

of fire and explosion. Noncombustibles and be provided and designed to minimize the adverse fire-resistant materials were used whenever

• ~

effects of fires on structures, systems, and practical throughout the facility. The control components important to safety. Firefighting rooms are constructed and furnished with non-systems shall be designed to assure that their flammable equipment. Adequate fire rupture or inadvertent operation does not extinguishers are supplied, and combustible significantly impair the safety capability of these materials, such as records, are kept to a structures, systems, and components.

minimum. The control rooms are equipped with emergency breathing apparatus to permit continuous occupancy in the unlikely event of a fire. Electrical distribution equipment will be physically located to reduce vulnerability of vital circuits to physical damage as a result of accidents."

3 Criterion 4 - Environmental and dynamic effects Criterion 40 - Missile Protection Protection Safe occupancy of the Control Room during design bases for engineered safety features shall be provided abnormal conditions is provided for in the against dynamic effects and missiles that might design of the Auxiliary Building. UFSAR Structures, systems, and components important to result from plant equipment failures.

Table 3-23 identifies the Control Rooms as safety shall be designed to accommodate the effects being designed for the maximum hypothetical of and to be compatible with the environmental Criterion 23 - Protection against AMl'ltiple earthquake and tornado loads (wind and conditions associated with normal operation, Disabilityfor Protection Systems The effects missiles) as well as turbine missiles. Adequate maintenance, testing, and postulated accidents, of adverse conditions to which redundant shielding shall be provided to protect occupants including loss-of-coolant accidents. These channels or Protection Systems might be of the Control Room for maximum structures, systems, and components shall be exposed in common, either under normal hypothetical accident conditions as required by appropriately protected against dynamic effects, conditions or those of an accident, shall not FSAR Section 7.7.5.

including the effects of missiles, pipe whipping, and result in a loss of the protection function.

Attachment I December 9, 2003 Page 14 of 17 Table 1 Comparison between GDC. AEC Criteria and Oconee Licensing Basis Item 10CFR 50 Appendix A GDC Comparable AEC Criteria Current Oconee Licensing Basis No. I I

1__

discharging fluids, that may result from equipment failures and from events and conditions outside the nuclear power unit. However, dynamic effects associated with postulated pipe ruptures in nuclear power units may be excluded from the design basis when analyses reviewed and approved by the Commission demonstrate that the probability of fluid system piping rupture is extremely low under conditions consistent with the design basis for the piping.

Turbine Missile characteristics are given in FSAR Table 3-10. Areas of the Auxiliary Building subject to review for turbine missile impact are identified in FSAR Table 3-23.

Reinforced concrete walls 3-6" thick are provided on the east side of the Control Rooms to resist these missile impacts. Other areas designated for turbine missile protection are located below the Turbine Building Operating Floor and are not subject to missile impacts.

By AEC Letter dated 12/15n2 from A.

Giambusso to Austin C. Thies, Oconee was required to provide assurance that the control room will be habitable and its equipment functional after a steam line or feedwater line break er that the capability for shutdown and cooldown of the units(s) will be available in another habitable area.

As stated in UFSAR Section 3.6.1, The analysis of effects resulting from postulated piping breaks outside containment is contained in Duke Power MDS Report No. OS-73.2, dated July 16, 1973 including supplement 2, dated March 12, 1974.

UFSAR Section 7.1.2.1 states, "Protective system equipment located in the Control Room...is designed for a mild environment, not LOCA conditions."

December 9, 2003 Page 15 of 17 Table 1 Comparison between GDC, AEC Criteria and Oconee Licensing Basis Item 10CFR 50 Appendix A GDC Comparable AEC Criteria Current Oconec Licensing Basis No.

4 Criterion 5 - Sharing of structures, systems, and Criterion 4 - Sharing of Systems Reactor UFSAR Section 3.1.1 states, "Portions of the components facilities shall not share systems or components following systems are shared as indicated.

unless it is shown safety is not impaired by the Where sharing between Oconee 1 and 2 is Structures, systems, and components important to sharing.

indicated, a separate system is provided for safety shall not be shared among nuclear power Oconee 3. Safety is not impaired by the units unless it can be shown that such sharing will sharing."

not significantly impair their ability to perform their safety functions, including, in the event of an The Control Room Ventilation System is given accident in one unit, an orderly shutdown and as a system that is shared between Units I and cooldown of the remaining units.

2 because the Unit I and 2 control rooms are in the same room or envelope.

5 Crterion 19 - Control room Criterion 11 - Control Room The facility shall UFSAR Section 3.1.1 states that, "The reactors be provided with a control room from which and associated equipment are controlled from A control room shall be provided from which actions to maintain safe operational status of panels located in the control rooms. The actions can be taken to operate the nuclear power the plant can be controlled. Adequate radiation control rooms are designed to permit unit safely under normal conditions and to maintain protection shall be provided to permit access, continuous occupancy following a maximum it in a safe condition under accident conditions, even under accident conditions, to equipment in hypothetical accident (MHA) (Section 7.7.5).

including loss-of-coolant accidents. Adequate the control room or other areas as necessary to All controls and instrumentation required to radiation protection shall be provided to permit shut down and maintain safe control of the monitor and operate the reactors and electric access and occupancy of the control room under facility without radiation exposures of power generating equipment are located within accident conditions without personnel receiving personnel in excess of 10CFR20 limits. It shall the control rooms. This includes indication of, radiation exposures in excess of 5 rem whole body, be possible to shut the reactor down and power level; process variables such as or its equivalent to any part of the body, for the maintain it in a safe condition if access to the temperatures, pressures, and flows; valve duration of the accident. Equipment at appropriate control room is lost due to fire or other cause.

positions; and control rod positions. All locations outside the control room shall be provided Engineered Safety Systems equipment are (1) with a design capability for prompt hot controlled and monitored from the control shutdown of the reactor, including necessary rooms. The status of all dynamic equipment instrumentation and controls to maintain the unit in (pumps, valves, etc.)--as well as pertinent a safe condition during hot shutdown, and (2) with a pressures, temperatures, and flows--is potential capability for subsequent cold shutdown displayed. The Radiation Monitoring System of the reactor through the use of suitable has provisions for alarms and for display of procedures.

instrumentation readouts in the control room.

II

Attachment I December 9, 2003 Page 16of 17 Table 1 Comparison between GDC, AEC Criteria and Oconee Licensing Basis Item 10CFR 50 Appendix A GDC Comparable AEC Criteria l

Current Oconec Licensing Basis No.1 I1 Applicants for and holders of construction permits and operating licenses under this part who apply on or after January 10, 1997, applicants for design certifications under part 52 of this chapter who apply on or after January 10, 1997, applicants for and holders of combined licenses under part 52 of this chapter who do not reference a standard design certification, or holders of operating licenses using an alternative source term under §50.67, shall meet the requirements of this criterion, except that with regard to control room access and occupancy, adequate radiation protection shall be provided to ensure that radiation exposures shall not exceed 0.05 Sv (5 rem) total effective dose equivalent (TEDE) as defined in §50.2 for the duration of the accident.

The concrete Reactor Buildings and control room walls and roofs are designed to provide adequate protection against direct radiation to control room personnel at all times."

UFSAR Section 9.6.1 states, "The Standby Shutdown Facility (SSF) is designed as a standby system for use under extreme -c emergency conditions. The system provides additional 'defense in-depth' protection for the health and safety of the public by serving as a backup to existing safety systems. The SSF is provided as an alternate means to achieve and maintain mode 3...following postulated fire, sabotage, or flooding events, and is designed in accordance with criteria associated with these events. Loss of all other station power does not impact the SSFs capability to mitigate each event. The SSF is also credited as the alternate AC (AAC) power source and the source of decay heat removal required to demonstrate safe shutdown during the required station blackout coping duration."

The use of the SSF for achieving and maintaining mode 3 permits the capability of making all necessary repairs to achieve cold shutdown within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after a fire accident.

Repairs, including replacement of power cabling, pump motors, valve controls, and switchgear associated with LPI, HPI, or LPSW may be required for cold shutdown. All material and components necessary to

Attachment I December 9, 2003 Page 17 of 17 Table 1 Comparison between GDC, AEC Criteria and Oconee Licensing Basis Item 10CFR 50 Appendix A GDC Comparable AEC Criteria Current Oconee Licensing Basis No.

accomplish required repairs are available on site. Procedures are available and training has been provided to implement the required repairs and replacements.

Catawba Nuclear Station's Response to Generic Letter 2003-01

CATAWBA NUCLEAR STATION'S RESPONSE TO GL 2003-01 Per the above-referenced generic letter, licensees were requested to provide information on three issues relating to the importance of ensuring that the control room envelope is operated and maintained in accordance with the facility's design and licensing bases.

The exact requests and the specific Catawba responses follow:

1.

Provide confirmation that your facility's control room meets the applicable habitability regulatory requirements (e.g., GDC 1, 3,4,5, and 19) and that the CRHSs are designed, constructed, configured, operated, and maintained in accordance with the facility's design and licensing bases.

Catawba Nuclear Station Units 1 and 2 were designed to the General Design Criteria (GDC) outlined in 10CFR50, Appendix A. A detailed comparison of the GDC 1, 3,4,5, and 19 versus the facility's current design and licensing bases is shown in Table 1.

The control room for Catawba Units I and 2 is shared for the operation of both units. The control complex envelope includes the control room proper, with the pressure, boundary consisting of the control room walls, floor, roof and doors and any penetrations of those in addition to the ductwork, dampers, filter units, and fan housings of the control room ventilation sub-system. The control room habitability system is comprised of the equipment, components, and the building enclosure that are provided to ensure a suitable environment is maintained for personnel and equipment in the control room for safe, long-term occupancy during both normal and emergency operation of the plant. The control complex is served by the Control Room Area Ventilation System (CRAVS).

Operation of the ventilation system is the same for all plant operational modes. The pressurization system utilizes filters for filtration of pressurizing air during all modes of plant operation. Technical Specification (TS) 3.7.10, "Control Room Area Ventilation System (CRAVS)," governs the operation of the CRAVS system. Selected Licensee Commitments (SLCs) 16.64, "Chlorine Detectors and Associated Circuitry," and 16.9-22, "Control Room Area (CRAVS) - Intake Alarms," govern the instrumentation at the CRAVS intakes. The Control Room Area Chilled Water System (CRACWS) provides temperature control for the control room and the control room area. The CRACWS consists of two independent and redundant trains that provide cooling to the control room and control room area. The CRACWS provides both normal and emergency cooling to the control room and control room area. Technical Specification TS 3.7.11, "Control Room Area Chilled Water System (CRACWS)," governs the operation of the CRACWS system.

The performance of the CRAVS and CRAWCS systems are monitored in accordance with Maintenance Rule requirements. System Health Reports are developed periodically and reviewed by senior station management.

As demonstrated with the responses to la, lb, and Ic that follow, the CRAVS and CRACWS systems are designed, constructed, configured, operated, and maintained in DATE Page 2 of 11 accordance with Catawba's design and licensing basis, with the exception of a CRACWS issue noted in the response to Question la below.

Emphasis should be placed on confirming:

(a)

That the most limiting unfiltered inleakage into your CRE (and the filtered inleakage if applicable) is no more than the value assumed in your design basis radiological analyses for CRE habitability. Describe how and when you performed the analyses, tests, and measurements for this confirmation.

Catawba Nuclear Station contracted NUCON International, Inc. to perform a tracer gas leak test and flow measurement for the Catawba control room. The testing was performed and successfully completed during November 12 - 15, 2002. NUCON provided Catawba with a written test report detailing the methodology and results of the tracer gas test. The Control Room Area Ventilation System and Control Room Area Chilled Water System provide the normal and emergency ventilation requirements to the control room. The control room is maintained at a positive pressure, > 1/8 inch water gauge with respect to all adjacent areas. The control room is maintained habitable for personnel and equipment by filtering outside air for pressurization and filtering a portion of the return air from the control room to clean-up the control room environment. The following is a description of the test method and results.

Outside air is drawn through each of two outside air intakes; therefore, sulfur hexafluoride (SF6) was used as a tracer gas to measure the outside airflow from one intake while a pitot tube was used for the other unit during the tests. The total air leakage rate from the control room was measured using the tracer gas technique. Testing was based on the constant injection test method described in ASTM E741 and ASTM E2099.

The outside airflows sample points for tracer gas testing were validated with pitot tube traverses for each line up tested. The line ups tested were A-train operating and B-train off in normal mode, B-train operating and A-train off in normal mode, and A-train and B-train operating at the same time. During all of the tests, the flow path to the control room Area was blanked off to allow for the inleakage to be simply estimated by taking the difference between the total control room leak rate and the sum of the outside airflows.

The uncertainty values were calculated at the 95% confidence level and include random and systematic uncertainties.

The results from the 2002 tracer gas test are as follows:

Alignment Envelope Leakage Outside Airflow Inleakage A-Train 1758 +/- 93 scfm 1729 +/- 128 scfm 29 +/- 158 scfm B-Train 1500 +/- 88 scfm 1497 +/-47 scfm 3 +/- 100 scfm A&B 3375 +/- 153 scfm 3355 +/- 72 scfm 20 +/- 169 scfm DATE Page 3 of 11 The analysis of radiological consequences of the design basis LOCA in the current licensing basis of Catawba was performed in conformance to Regulatory Guide 1.4. In that analysis the rate of unfiltered inleakage to the control room was set to a total of 30 cfm, including 10 cfm unfiltered inleakage attributable to control room access and egress.

The integral tracer gas test measurements account for leakage from pneumatic controls in the control room. A Catawba evaluation of instrument air leakage into the control room documented in an approved site calculation has estimated the total rate of pneumatic inleakage at 15 cfm. Given the configuration of the instrument air lines leading to the control room (long thin tubes), it is expected that most of the fission products that might be entrained in instrument air flow will plate onto the lining of the tubes. This instrument air inleakage has been judged to be filtered rather than unfiltered inleakage and the test results therefore confirm the value assumed in the design basis radiological analysis.

The Staff previously approved a license amendment request which included partial implementation of Alternative Source Ter. based on analyses of radiological consequences of a Fuel Handling Accident and Weir Gate Drop in April 2002 (Reference 1 through Reference 4). Duke Energy Corporation has submitted a license amendment request which includes full implementation of Alternative Source Terms for Catawba Nuclear Station. This license amendment request includes an analysis of the radiological consequences of a design basis Loss of Coolant Accident (LOCA) postulated to occur at Catawba (Reference 5 through Reference 7). In all these analyses the rate of unfiltered inleakage to the control room was set to 100 cfm.

Catawba has a condition in its licensing basis that is currently operable but nonconforming. A non-safety, non-seismic control room air handling unit drain line loop seal exist on both the Ul and U2 sides of the control room. These lines will be seismically supported per plant modifications which are scheduled to be completed in 2004. Continued operation was evaluated per GL 91-18 and determined to be acceptable.

(b)

That the most limiting unfiltered inleakage into your CRE is incorporated into your hazardous chemical assessment. This inleakage may differ from the value assumed in your design basis radiological analyses. Also confirm that the reactor control capability is maintained from either the control room or the alternate shutdown panel in the event of smoke.

The following sections describe the hazardous chemical assessment and smoke evaluation for Catawba.

Hazardous Chemical Assessment An evaluation of the impact of a toxic gas release on control room habitability is documented in Table 6-100 and 6-101 of the Catawba UFSAR. These tables provide comparisons of the Catawba control room design against the guidelines described in Regulatory Guide 1.95 and 1.78. An evaluation of onsite and offsite hazardous chemicals is documented in a site controlled calculation that is periodically reviewed for changes that may require revision. The evaluation accounts for the most limiting unfiltered DATE Page 4 of II inleakage into the Catawba control room envelope, which occurs during normal operation. The hazardous chemical assessment is performed using the progressive screening steps and control room habitability evaluation guidance provided in Regulatory Guide 1.78. The conclusion of the analysis is that on-site and off-site releases of hazardous chemicals do not result in concentrations within the control rooms that are unacceptable.

Smoke Evaluation The Standby Shutdown Facility (SSF) provides an alternate and independent means to achieve and maintain a hot shutdown condition for scenarios in which the control room is unavailable or equipment it controls is unavailable. The SSF was designed for safe shutdown during postulated fire, security, and Station Blackout (SBO) events.

The SSF is housed in a separate structure that is physically removed from the auxiliary building and control room. The SSF has its own ventilation system that is completely independent from any of the plant ventilation systems.

The following sections address smoke design, egress pathways, and smoke control to assure reactor control capability is maintained in the event of smoke.

Smoke Design The strategy used to assure the capability to safely shut the Catawba units down in the event of a fire that results in evacuation of the control room relies on control from the Standby Shutdown Facility (SSF). The Standby Shutdown System provides an alternate means to achieve and maintain a hot shutdown condition in addition to the normal shutdown capabilities available.

Significant physical separation exists both laterally and vertically between the Control Room Ventilation System outside air intakes and the SSF Control Room Ventilation System outside air intake. In addition, there is also significant separation from the point of discharge from the station exhaust points (Turbine Building Exhaust Fan Discharges and Station Ventilation Stack) to the SSF Ventilation System outside air intake.

Egress Pathways Evaluation The Catawba Auxiliary Building has several pathways to traverse from the control room which is located at ground level elevation 594'-O" out of the building to the yard and to the SSF at elevation 594'-O".

Use of the egress pathways from the control room to the SSF is an operator training Job Performance Measure (JPM). JPMs verify that licensed operators can properly man the SSF. This JPM is a required part of both the initial and continuing requalification training.

DATE Page 5 of 11 Based on the numerous available alternate pathways the operators could use to get from the control room to the SSF, a fire in the Auxiliary Building is not expected to significantly challenge the operators' ability to man the SSF, w-hich contains controls needed to accomplish safe shutdown.

Smoke Control If smoke is detected in an outside air intake, an alarm is received in the control room alerting the operators of the condition. Since the VC System can pressurize the control room with only one intake open, and the pressurizing filter trains (1(2)CRA-PFT-1) filter out smoke prior to supplying air to the control room, it is the operator's prerogative whether or not to close the affected intake. The determination to close or keep open an intake should consider issues such as the validity of the alarm and the status of the other intake.

In the event of a fire in the control room, smoke will be purged from the control room by operating one of the pressurizing filter train fans to pressurize the control room and opening the doors from the control room to the auxiliary building. The smoke from the control room will be exhausted to the atmosphere by the auxiliary building exhaust fans.

Overall Smoke Evaluation Conclusion Based on the fact that Catawba relies on the SSF as an alternate and independent means to achieve and maintain a hot shutdown condition for scenarios in which the control room is unavailable or equipment it controls is unavailable and the physical layout of the site structures and the numerous egress pathways available for operators to reach the SSF, there is adequate assurance that safe shutdown will be achieved during a smoke event originating inside or outside the control room.

(c)

That your technical specifications verify the integrity of the CRE, and the assumed inleakage rates of potentially contaminated air. If you currently have a AP surveillance requirement to demonstrate CRE integrity, provide the basis for your conclusion that st remains adequate to demonstrate CRE integrity in light of the ASTM E741 testing results. If you conclude that your AP surveillance requirement is no longer adequate, provide a schedule for: 1) revising the surveillance requirement in your technical specification to reference an acceptable surveillance methodology (e.g., ASTM E-741), and 2) making any necessary modifications to your CRE so that compliance with your new surveillance requirement can be demonstrated.

If your facility does not currently have a technical specification surveillance requirement for your CRE, explain how and on what frequency you confirm your CRE integrity.

Currently, Catawba Technical Specification Surveillance Requirement SR 3.7.10.4 periodically verifies the positive pressurization of the control room with respect to adjacent areas. Duke plans to implement TSTF-448, "Control Room Habitability," on a DATE Page 6 of I1 reasonable schedule once the TSTF is made available for use. The industry submitted TSTF-448 to the NRC for review on August 19, 2003. This TSTF will follow the standard TSTF process and schedule.

No modifications are necessary for Catawba to demonstrate compliance with TSTF-448 as submitted.

2.

If you currently use compensatory measures to demonstrate CRE habitability, describe the compensatory measures at your facility and the corrective actions needed to retire these compensatory measures Catawba does not rely on compensatory measures to demonstrate control room envelope habitability.

3.

If you believe that your facility is not required to meet either the GDC, the draft GDC, or the "Principle Design Criteria" regarding control room habitability, in addition to responding to items 1 and 2 above, provide the documentation (e.g., Preliminary Safety Analysis Report, Final Safety Analysis Report sections, or correspondence, etc.) of the basis for this conclusion and identify your actual requirements.

The Catawba facility is required to meet the GDC regarding control room habitability. A detailed comparison of the GDC 1, 3, 4, 5, and 19 versus the facility's current design and licensing bases is shown in Table 1.

REFERENCES

1. G. R. Peterson to U.S. Nuclear Regulatory Commission, "Proposed Amendment for Partial Scope Implementation of the Alternate Source Term and Proposed Amendment to Technical Specifications (TS) 3.7.10, Control Room Area Ventilation System, T.S. 3.7.11, Control Room Area Chilled Water System, TS 3.7.13, Fuel Handling Ventilation Exhaust System, and TS 3.9.3, Containment Penetrations,"

December 20, 2001.

2. G. R. Peterson to U.S. Nuclear Regulatory Commission, "Planned Revision to Proposed Amendment for Partial Scope Implementation of the Alternate Source Term," February 14, 2002.

3. G. R. Peterson to U.S. Nuclear Regulatory Commission, "Revision to Proposed Amendment for Partial Scope Implementation of the Alternate Source Term and Proposed Amendment to Technical Specification (TS) 3.7.10, Control Room Area Ventilation System, T.S.3.7.11, Control Room Area Chilled Water System, T.S.

3.7.13, Fuel Handling Ventilation Exhaust System, and 3.9.3, Containment Penetrations," March 26, 2002.

DATE Page 7 of 11

4. Chandu Patel (USNRC) to G.R. Peterson, "Catawba Nuclear Station, Units 1 and 2 Re: Issuance of Amendments (TAC Nos. MB3758 and MB3759)," April 23, 2002.

5. G. R. Peterson (Duke Energy) to U.S. Nuclear Regulatory Commission, "Proposed Technical Specification and Bases Amendment: Technical Specification and Bases 3.6.10, Annulus Ventilation System (AVS); Technical Specification and Bases 3.6.16, Reactor Building; Technical Specification and Bases 3.7.10, Control Room Area Ventilation System (CRAVS); Technical Specification and Bases 3.7.12, Auxiliary Building Filtered Ventilation Exhaust System (ABFVES); Technical Specification and Bases 3.7.13, Fuel Handling Ventilation Exhaust System (FHVES);

Technical Specification and Bases 3.9.3, Containment Penetrations; Technical Specification 5.5.11, Ventilation Filter Testing Program (VFTP)," November 25, 2002.

6. R. E. Martin (USNRC) to D. M. Jamil, "Catawba Nuclear Station, Units 1 and 2 Re:

Request for Additional Information (TAC Nos. MB7014 and MB70i5),"

September 11, 2003.

7. D. M. Jamil to U.S. Nuclear Regulatory Commission, "Proposed Technical Specifications and Bases Amendment: Technical Specification and Bases 3.6.10, Annulus Ventilation System (AVS); Technical Specification and Bases 3.6.16, Reactor Building; Technical Specification and Bases 3.7.10, Control Room Area Ventilation System (CRAVS); Technical Specification and Bases 3.7.12, Auxiliary Building Ventilation Filtered Exhaust System (ABFVES); Technical Specification and Bases 3.7.13, Fuel Handling Ventilation Exhaust System (FHVES); Technical Specification and Bases 3.9.3, Containment Penetrations; Technical Specification and Bases 5.5.11, Ventilation Filter Testing Program (VFTP); TAC Numbers MB7014 and MB7015," November 13, 2003.

DATE Page 8 of 11 Table 1 Comparison between GDC and Catawba Licensing Basis 10 CFR 50 Appendix A Criteria Catawba Current Licensing Basis Criterion I - Quality standards and records Structures, systems, and components important to safety shall be designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety functions to be performed. Where generally recognized codes and standards are used, they shall be identified and evaluated to determine their applicability, adequacy, and sufficiency and shall be supplemented or modified as necessary to assure a quality product in keeping with the required safety function. A quality assurance program shall be established and implemented in order to provide adequate assurance that these structures, systems, and components will satisfactorily perform their safety functions. Appropriate records of the design, fabrication, erection, and testing of structures, systems, and components important to safety shall be maintained by or under the control of the nuclear power unit licensee throughout the life of the unit Per the Chapter 3 of the Catawba UFSAR, the structures, systems, and components of this facility are classified, as defined in ANS N 18.2, according to their importance in the prevention and mitigation of accidents using generally recognized engineering codes and standards. The Auxiliary Building, which contains the control room, is designed as a Category "I" system with seismic and tornado protection.

The Control Room Area Ventilation System (CRAVS) and the Control Room Area Chilled Water System (CRACWS) are QA Condition 1 systems. The control room area air conditioning-systems are engineered safety features. Each redundant train (100 percent capacity) of air handling units, water chillers, pumps, pressurizing filter trains and fans, and outside air intake isolation valves is served from separate trains of the Emergency Class IE Power System. All essential air conditioning and ventilating equipment, ductwork and supports are designed to withstand the safe shutdown earthquake. This assures the integrity and availability of one train of the Control Room Area Ventilation System in the event of any single active failure. Design of the Control Room Area Ventilation System is such that the maximum radiation dose received by the control room personnel under accident conditions is within the limits of General Design Criterion 19 of Appendix A to IOCFR50. The Control Room Area Ventilation System is designed to maintain temperature, cleanliness and pressurization in the areas served during normal plant operation, shut-down, post-accident conditions, and in all feasible weather conditions. Two isolation valves are provided on each outside air intake to ensure the capability of manual closure of the intake on high radiation, high smoke concentration, or high chlorine concentration.

Duke's quality assurance program conforms with the requirements of IOCFR50, Appendix B, Quality Assurance Criteria for Nuclear Power Plants. This Quality Assurance program is described in Chapter 17 of the Catawba UFSAR. Included in this quality assurance program is specific direction for the maintenance of appropriate records.

Criterion 3 - Fire protection Structures, systems, and components important to safety The Fire Protection System provides fire detection equipment for areas where potential for fire is shall be designed and located to minimize, consistent with greatest or areas not normally occupied by personnel. Also, reliable sources of either water or other safety requirements, the probability and effect of carbon dioxide are provided to appropriate parts of the station.

fires and explosions. Noncombustible and heat resistant materials shall be used wherever practical throughout the The station is designed to utilize non-combustible and heat-resistant materials, wherever practical.

unit, particularly in locations such as the containment and Duplication and physical separation of components to provide redundancy against other hazards DATE Page 9 of 11 Table 1 Comparison between GDC and Catawba Licensing Basis 10 CFR 50 Appendix A Criteria Catawba Current Licensing Basis control room. Fire detection and fighting systems of also protects against simultaneous failures due to local fires.

appropriate capacity and capability shall be provided and designed to minimize the adverse effects of fires on Inadvertent operation of or a crack in a fire suppression system would not preclude safe shutdown structures, systems, and components important to safety.

of the plant since redundant trains of safety-related equipment required for safe shutdown are Firefighting systems shall be designed to assure that their located in separate rooms, have adequate spatial separation, or have appropriate water spray rupture or inadvertent operation does not significantly shielding. In general, pressure-retaining equipment or piping is not permitted in the control impair the safety capability of these structures, systems, complex enclosure. Several small hand-held fire extinguishers are located within the area for local and components.

fire control. Areas of the Auxiliary Building and the Service Building which contain high-pressure equipment or piping have no direct interface with the control room enclosure.

Criterion 4 - Environmental and dynamic effects design basis Structures, systems, and components important to safety Structures, systems and components important to safety are designed to function in a manner which shall be designed to accommodate the effects of and to be assures public safety at all times. These structures, systems and components are protected for all compatible with the environmental conditions associated worst-case conditions by appropriate missile barriers, pipe restraints, and station layout. The with normal operation, maintenance, testing, and control room is designed to withstand such missiles as may be directed toward it and still maintain postulated accidents, including loss-of-coolant accidents.

the capability of controlling the units. Class IE electrical equipment is designed and qualified to perform its safety function(s) under the harsh environmental conditions applicable to its location.

These structures, systems, and components shall be appropriately protected against dynamic effects, including The control room habitability system is comprised of the equipment, components, and the building the effects of missiles, pipe whipping, and discharging enclosure that are provided to ensure a suitable environment is maintained for personnel and fluids, that may result from equipment failures and from equipment in the control room for safe, long-term occupancy during both normal and emergency events and conditions outside the nuclear power unit.

operation of the plant. The design bases of the habitability system for the control room includes:

1.

The capability to withstand the safe shutdown earthquake, However, dynamic effects associated with postulated pipe

2.

The capability to function properly following any single active failure, ruptures in nuclear power units may be excluded from the

3.

The capability to function during a design basis tornado, design basis when analyses reviewed and approved by the

4.

The capability to detect and limit concentrations of chlorine gas as specified in Regulatory Commission demonstrate that the probability of fluid Guide 1.95 or products of combustion entering the control room, system piping rupture is extremely low under conditions

5.

The capability to shield control room personnel from radiation sources, such that exposure consistent with the design basis for the piping.

to personnel will not exceed the limits specified in General Design Criterion 19 of Appendix A to IOCFR50,

6.

The capability to detect and limit the introduction of airborne radioactive contamination into the control room such that exposure to personnel will not exceed the limits specified in General Design Criterion 19 of Appendix A to 10CFR50, and

7.

The capability to permit safe shutdown of the plant from the control room following a loss-of-coolant accident (LOCA).

DATE Page lOof II Table 1 Comparison between GDC and Catawba Licensing Basis 10 CFR 50 Appendix A Criteria Catawba Current Licensing Basis The control room is located on the top floor of the Auxiliary Building and is bounded on the north and south sides by electrical penetration rooms which contain no piping. The east side of the control room is bounded by the equipment area housing the control room ventilation equipment.

Piping in this area consists of low pressure, low volume chilled water and low pressure, low volume heating steam. On the west side, the control room is bounded by the computer room and supporting areas. Piping in this area consists of sanitary waste and vent piping, drinking water and instrument air. None of these systems are high-energy systems. Immediately below the control room is the cable room containing no piping. Based on the above physical parameters, the control room is structurally isolated from areas containing high-energy systems; therefore, there are no unacceptable consequences to the control room from the postulated break of high-energy piping systems.

Criterion 5 - Sharing of systems, structures, and components Structures, systems, and components important to safety Structures, systems, and components which are either shared (a) between the two units or (b) shall not be shared among nuclear power units unless it among systems within a unit are designed such that there is no interference with basic function and can be shown that such sharing will not significantly operability of these systems due to sharing. This design protects the ability of shared structures, impair their ability to perform their safety functions, systems, and components to perform all safety furctions properly.

including, in the event of an accident in one unit, an orderly shutdown and cooldown of the remaining units Criterion 19 - Control rocm A control room shall be provided from which actions can The station is provided with a control room located in the Auxiliary Building where the nuclear be taken to operate the nuclear power unit safely under power unit is operated under normal and accident conditions. The control room is designed and normal conditions and to maintain it in a safe condition equipped to minimize the possibility of events which might preclude occupancy. In addition, under accident conditions, including loss-of-coolant provisions have been made for bringing both units to and maintaining them in a hot shutdown accidents. Adequate radiation protection shall be provided condition for an extended period of time from locations outside the main control room. If to permit access and occupancy of the control room under necessary, the reactor may subsequently be placed in the cold shutdown condition.

accident conditions without personnel receiving radiation exposures in excess of 5 rem whole body, or its equivalent The employment of non-combustible and fire retardant materials in the construction of the control to any part of the body, for the duration of the accident.

room, the limitation of combustible supplies, the location of fire fighting equipment, and the Equipment at appropriate locations outside the control continuous presence of a highly trained operator minimizes the possibility that the control room room shall be provided (1) with a design capability for will become uninhabitable. Additionally, the Control Room Area Ventilation System is designed prompt hot shutdown of the reactor, including necessary to maintain the control room at a positive pressure to minimize airborne radioactivity in-leakage.

instrumentation and controls to maintain the unit in a safe Under high radiation conditions, makeup air is recycled through a system of filters. The control condition during hot shutdown, and (2) with a potential room enclosure is virtually insensitive to wind effects since only a small portion of wall on the west capability for subsequent cold shutdown of the reactor side and the control room roof is exposed to the outside.

DATE Page 11 of II Table 1 Comparison between GDC and Catawba Licensing Basis 10 CFR 50 Annendix A Criteria Catawba Current Licensine Basis 4~~~~~~~~~~~~~----

through the use of suitable procedures.

Applicants for and holders of construction permits and operating licenses under this part who apply on or after January 10, 1997, applicants for design certifications under part 52 of this chapter who apply on or after January 10, 1997, applicants for and holders of combined licenses under part 52 of this chapter who do not reference a standard design certification, or holders of operating licenses using an alternative source term under §50.67, shall meet the requirements of this criterion, except that with regard to control room access and occupancy, adequate radiation protection shall be provided to ensure that radiation exposures shall not exceed 0.05 Sv (5 rem) total effective dose equivalent (TEDE) as defined in §50.2 for the duration of the accident.

Sufficient shielding, distance, and containment integrity are provided to assure that control room personnel are not subjected to doses under postulated accident conditions which would exceed 5 rem whole body or 5 rem TEDE, as applicable. Control room layouts provide the necessary controls to start, operate, and shut down the units with sufficient information display and alarm monitoring to assure safe and reliable operation under normal and accident conditions. Special emphasis is given to maintaining control during accident conditions. The layout of the Engineered Safety Feature devices of the control board is designed to minimize the time required for the operator to evaluate system performance under accident conditions.

A separate plant subsystem has been incorporated into the Catawba design to allow a means of limited plant shutdown, independent from the control room and auxiliary shutdown panels. This system, known as the Standby Shutdown System, provides an alternate means to achieve and maintain a hot shutdown condition following postulated fire and sabotage events. This system is in addition to the normal shutdown capabilities available. The Standby Shutdown System (except for interfaces to existing safety related systems) is designed in accordance with accepted fire protection and security requirements and is not designed as a safety related system.