ML021620348

Final Reponse to Second Nuclear Regulatory Commission Request for Additional Information and Verbal Concerns Regarding License Amendment Request for Control Room Habitability

ML021620348
Person / Time
Site:	Cook
Issue date:	06/05/2002
From:	Joseph E Pollock Indiana Michigan Power Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
AEP:NRC:2075
Download: ML021620348 (31)

Text

Indiana Michigan Power Company 500 Circle Drive Buchanan, Ml 49107 1395 INDIANA MICHIGAN POWER June 5, 2002 AEP:NRC:2075 10 CFR 50.90 Docket No.:

50-315 50-316 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Mail Stop O-P 1-17 Washington, DC 20555-0001 Donald C. Cook Nuclear Plant Units 1 and 2 FINAL RESPONSE TO SECOND NUCLEAR REGULATORY COMMISSION REQUEST FOR ADDITIONAL INFORMATION AND VERBAL CONCERNS REGARDING LICENSE AMENDMENT REQUEST FOR CONTROL ROOM HABITABILITY

References:

1) Letter from R. P. Powers, Indiana Michigan Power Company (I&M), to U. S. Nuclear Regulatory Commission (NRC)

Document Control Desk, "License Amendment Request for Control Room Habitability and Generic Letter 99-02 Requirements," C0600-13, dated June 12, 2000

2) Letter from J. F. Stang (NRC) to R. P. Powers (I&M),

"Donald C. Cook Nuclear Plant, Units 1 and 2 - Request for Additional Information, License Amendment Request for Control Room Habitability, (TAC Nos.

MA9394 and MA9395)," dated March 29, 2001

3) Letter from M. W. Rencheck (I&M) to NRC Document Control Desk, "Partial Response to Nuclear Regulatory Commission Request for Additional Information Regarding License Amendment Request for Control Room Habitability, (TAC Nos. MA9394 and MA9395)," C0601-03, dated June 19, 2001

U. S. Nuclear Regulatory Commission AEP:NRC:2075 Page 2

4) Letter from M. W. Rencheck (I&M) to NRC Document Control Desk, "Final Response to Nuclear Regulatory Commission Request for Additional Information Regarding License Amendment Request for Control Room Habitability (TAC Nos.

MA9394 and MA93 95)," C0801-02, dated August 17, 2001

5) Letter from J. F. Stang (NRC) to R. P. Powers (I&M),

"Donald C. Cook Nuclear Plant, Units 1 and 2 - Request for Additional Information, License Amendment Request for Control Room Habitability (TAC Nos.

MA9394 and MA9395)," dated August 16, 2001

6) Letter from A. C. Bakken (I&M) to NRC Document Control Desk, "Partial Response to Second Nuclear Regulatory Commission Request for Additional Information Regarding License Amendment Request for Control Room Habitability,"

C0102-04, dated January 15, 2002

7) Regulatory Guide (R.G.)

1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," dated July 2000

8) Regulatory Guide DG-1081, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at nuclear Power Reactors," dated December 1999 This letter provides Indiana Michigan Power Company's (I&M) responses to remaining Nuclear Regulatory Commission (NRC) questions and verbal concerns regarding a previously proposed license amendment to address control room habitability issues at Donald C. Cook Nuclear Plant (CNP).

In Reference 1, I&M proposed to amend Facility Operating Licenses DPR-58 and DPR-74 for CNP Unit 1 and Unit 2 to address control room habitability issues. Reference 2 transmitted an NRC Request for Additional Information (RAI) regarding the proposed amendment.

References 3 and 4 transmitted I&M's responses to that RAI. Reference 5 transmitted a second RAI pertaining to the proposed amendment. Reference 6 transmitted a partial response to the second RAI and to NRC concerns identified in phone conferences with members of the NRC staff.

In Reference 6, I&M committed to provide additional information in a subsequent letter.

U. S. Nuclear Regulatory Commission AEP:NRC:2075 Page 3 This letter fulfills I&M's commitment made in Reference 6 to provide additional information. to this letter provides the remainder of the information requested by the second RAI (Reference 5). Attachment 2 addresses the remainder of the NRC concerns identified in phone conferences with members of the NRC staff as documented in Reference 6. Attachments 3 and 4 provide supporting information for Attachment 2 regarding the analysis of a locked rotor event. contains information proprietary to Westinghouse Electric Company, LLC. provides an affidavit setting forth the basis on which the proprietary information contained in may be withheld from public disclosure pursuant to 10 CFR 2.790. contains a non-proprietary version of Attachment 4. There are no new commitments in this letter.

The information provided in this letter consists of supporting information for the amendment request previously submitted by References 1 and 4.

The information provided in this letter does not. alter the requested amendment and does not affect the validity of the original evaluation of significant hazards considerations performed in accordance with 10 CFR 50.92 as documented in to Reference 1.

The environmental assessment provided in to Reference 1 also remains valid.

Should you have any questions, please contact Mr. Gordon P. Arent, Manager of Regulatory Affairs, at (616) 697-5553.

Sincerely, J. E. Pollock Site Vice President

/dmb Attachments:

1. Final Response to Second RAI

2. Final Response to Verbal Concerns

3. Locked Rotor Rods-in-DNB Transient Analysis Information

4. DNBR Margins and Allocations (Proprietary)

5. Affidavit for Withholding Attachment 4 from Public Disclosure

6. Non-Proprietary Version of Attachment 4

U. S. Nuclear Regulatory Commission Page 4 c:

K. D. Curry, w/o attachments J. E. Dyer MDEQ - DW & RPD, w/o attachments NRC Resident Inspector R. Whale, w/o attachments AEP:NRC:2075

U. S. Nuclear Regulatory Commission AEP:NRC:2075 Page 5 AFFIRMATION I, Joseph E. Pollock, being duly sworn, state that I am Site Vice President of Indiana Michigan Power Company (I&M), that I am authorized to sign and file this request with the Nuclear Regulatory Commission on behalf of I&M, and that the statements made and the matters set forth herein pertaining to I&M are true and correct to the best of my knowledge, information, and belief.

Indiana Michigan Power Company J. E. Pollock Site Vice President SWORN TO AND SUBSCRIBED BEFORE ME THIS DAY OF jurj 2002 My Commission Expires

,.:!0 JERNNER L KERNOSKY NotMy Public, Berrien County, Mchigan MY Commission Expires May 26,2005

ATTACHMENT 1 TO AEP:NRC:2075 FINAL RESPONSE TO SECOND NUCLEAR REGULATORY COMMISSION REQUEST FOR ADDITIONAL INFORMATION REGARDING PROPOSED CONTROL ROOM HABITABILITY AMENDMENT The documents referenced below are identified in the cover letter for this transmittal.

This attachment provides Indiana Michigan Power Company's (I&M) final response to a Nuclear Regulatory Commission (NRC) request for additional information (RAI) transmitted by Reference 5. I&M provided a partial response to NRC Question 1 and full responses to all other NRC questions from the RAI in Reference 6. The remainder of the response to NRC Question 1 is provided below.

NRC Question 1 The meteorological data set (ARCON96 format) provided by I&M as an attachment to the June 19, 2001, letter appears to contains data which are questionable. For example:

" For the year 1996, stability class A was reported for 4912 hours0.0569 days <br />1.364 hours <br />0.00812 weeks <br />0.00187 months <br /> out of the available 8760 hours0.101 days <br />2.433 hours <br />0.0145 weeks <br />0.00333 months <br />; 4404 hours0.051 days <br />1.223 hours <br />0.00728 weeks <br />0.00168 months <br /> in 1997; and 4653 hours0.0539 days <br />1.293 hours <br />0.00769 weeks <br />0.00177 months <br /> in 1998. These appear to be unusually large fractions and are inconsistent with historic data reported in Table 2.2-4 of the Updated Final Safety Analysis [Report] (UFSAR).

"* There are periods in the data set in which the reported stability class did not change for numerous hours; 56 hours6.481481e-4 days <br />0.0156 hours <br />9.259259e-5 weeks <br />2.1308e-5 months <br /> in one case. Given that this encompasses two diurnal cycles, the constant stability class suggests a potential instrumentation or data processing problem resulting in invalid data that perhaps should have been flagged as such.

"* Over 25 percent of the observations of stability class A for the 3 years were reported between the evening hours of 1900 to 0700. This appears to be an untypically large fraction.

The I&M response to Question 9 indicates that the data were validated by a meteorologist on I&M's contractor's staff to ensure that the wind speed and direction were within normal operating ranges. The response also states that invalid data were not used. The response does not explicitly state that a similar validation was performed on the stability class data. (The staff did determine that a wind rose prepared using the submitted wind speed and wind direction data showed a good correlation to the 1992 data reported in the UFSAR.)

Although the staff recognizes that local temporal meteorological conditions can often result in observations that appear askew, the large quantity of stability class A observations in the D. C.

Cook data set raises a question regarding the representativeness of the reported data. Since to AEP:NRC:2075 stability class A is generally more favorable with regard to dispersion than the other classes, the reported z/Q values may not be adequately conservative.

Please provide a suitable explanation of the conditions identified above. If the conditions described above cannot be reasonably explained, or are deemed to be the result of instrumentation or processing problems, please provide a justification of why these data are appropriate for use in determining short-term dispersion estimates for design-basis calculations.

I&M Response to NRC Question 1 I&M's partial response to this question, provided in Reference 6, described two errors that were identified in the processing of the meteorological data transmitted to the NRC. I&M provided a computer disc containing new data and provided new atmospheric dispersion (X/Q) values calculated from that data. Using the new X/Q values, I&M determined bounding doses for the events identified in the original amendment request, except for the dose from a large break loss of coolant accident (LOCA). I&M committed to provide the recalculated dose from a large break LOCA in a subsequent letter. The large break LOCA dose has since been recalculated by revising the originalanalysis documented in Reference 1 to address the X/Q errors, to address the NRC concern described in Attachment 2 to this letter regarding the value assumed for the fraction of emergency core cooling system (ECCS) leakage that becomes airborne, and to modify other assumptions for consistency with Regulatory Guide (RG) 1.183 (Reference 7).

The specific changes are as follows:

"* New X/Q values are used as described in Reference 6.

" The fraction of iodine in ECCS leakage that becomes airborne during the recirculation phase is assumed to be 0.1. In the original analysis, a value of 0.0001 was assumed. However, the guidance in RG 1.183 (Reference 7), Appendix A, Paragraph 5.5, states that a value of 0.1 should be assumed. Therefore, the analysis was revised accordingly.

" The procedurally controlled ECCS effective leak rate limit is assumed to be 0.1 gallons per minute. In the original analysis, the procedurally controlled effective leak rate was assumed to be 0.2 gallons per minute. However, this did not provide the factor of two conservatism that RG 1.183, Appendix A, Paragraph 5.2 states should be included in the value used in the dose calculation. Since the value of 0.2 gallons per minute used in the dose calculation has not been changed, the procedurally controlled ECCS effective leak rate limit has been reduced to 0.1 gallons per minute to provide the factor of two conservatism specified in RG 1.183.

"* The failure of a containment spray pump is no longer postulated. In the original analysis, a simultaneous, independent failure of both a containment spray pump and a control room Page 2 to AEP:NRC:2075 emergency ventilation system (CREVS) emergency inlet damper was assumed. However, RG 1.183, Paragraph 5.1.2, states that only the single active failure that results in the most limiting radiological consequences should be assumed. I&M has determined that a failure of a CREVS emergency inlet damper is the most limiting single active failure, and has revised the analysis accordingly.

" The passive failure of an ECCS recirculation piping pressure boundary is no longer assumed.

In the original analysis, a passive failure resulting in a leak rate of 50 gallons per minute for half an hour starting 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> into the event was assumed in accordance with draft RG DG-1081 (Reference 8), Appendix A, Paragraph 5.3. However, the assumption of a passive failure in ECCS piping is not required by RG 1.183, which superceded DG-1081. Therefore, the analysis was revised accordingly.

The radioactive release from the fuel gap is assumed to be a linear release of 5 percent of the noble gases, halogens, and alkali metals in the core over the duration of the release phase of thirty minutes, starting at 30 seconds after accident initiation. In the original analysis, the fuel gap release was assumed to be a release of 3 percent of these radionuclides 30 seconds after accident initiation, followed by a linear release of 2 percent of these radionuclides over 30 minutes. The timing of the release in the revised analysis was changed to be consistent with the guidance in RG 1.183, Table 2, Paragraph 3.3, and Table 5.

With the changes described above, the revised analysis determined that the dose to control room operators from a large break LOCA would be less than or equal to 4.2 rem total effective dose equivalent (TEDE). This is below the 10 CFR 50, Appendix A, General Design Criterion 19 limit of 5 rem TEDE.

Page 3

ATTACHMENT 2 TO AEP:NRC:2075 FINAL RESPONSES TO NUCLEAR REGULATORY COMMISSION CONCERNS IDENTIFIED IN PHONE CONFERENCES The documents referenced below are identified in the cover letter for this transmittal.

The following provides Indiana Michigan Power Company's (I&M) final responses to Nuclear Regulatory Commission (NRC) verbal concerns regarding I&M's response (Reference 4) to the first NRC Request for Additional Information (Reference 2). These concerns were identified in telephone discussions conducted August 28, 2001, October 12, 2001, and November 13, 2001, between members of the NRC staff and I&M personnel.

Responses to all identified NRC concerns, except those regarding I&M's responses to NRC Questions 10 and 21, were transmitted by Reference 6. The responses to NRC concerns regarding I&M's responses to NRC Questions 10 and 21 are provided below.

NRC Concern Regarding I&M Response to Question 10 I&M's response provided justification for use of an iodine airborne fraction of 10-4 in the proposed large break LOCA analysis to determine the iodine released from ECCS leakage, rather than the 10-1 value given in RG 1.183. I&M's justification was based on laboratory experiments described in the original CNP FSAR and a theoretical study of the same vintage.

The NRC considers thatjustification for use of an iodine airborne fraction that differs so greatly from that given from the value given in RG 1.183 requires a numeric argument linked to the actual conditions in the plant.

I&M Response to NRC Concern As described in Attachment 1 to this letter, the dose from a large break loss of coolant accident has been recalculated using a value of 10' for the iodine airborne fraction in accordance with Regulatory Guide 1.183 (Reference 7). The recalculated dose was determined to be below the limit stated in 10 CFR 50, Appendix A, General Design Criterion 19.

NRC Concern Regarding I&M Response to Question 21 In order for the staff to continue its review of the locked rotor event described in your application, the staff will need the following additional information: (1) a description of and justification for the initial assumptions used in the new analysis, (2) a comparison of the differences in assumption between the previous and new analyses, (3) the sequence of events for the new analysis, and (4) the results of the analysis including plots of important parameters to show plant response and minimum DNBR.

to AEP:NRC:2075 I&M Response to NRC Concern The specific items identified in the NRC concern are addressed below.

(1) A description of and justification for the initial assumptions used in the new analysis.

The initial assumptions are shown in Table 1 of Attachment 3 to this letter. They are based on and bound values in the Updated Final Safety Analysis Report (UFSAR).

(2) A comparison of the differences in assumption between the previous and new analyses.

As noted in Reference 4, the previous UFSAR analyses for both units were based on a power shape that bounded all expected core reloads.

The new (cycle-specific) analyses use cycle-specific power shape rather than a bounding power shape.

The acceptance criterion for the number of rods experiencing departure from nucleate boiling (DNB) also changed.

The acceptance criteria in previous analyses allowed some rods to experience DNB. The acceptance criterion for the new analysis is that no rods experience DNB.

Unused DNB ratio (DNBR) margin may be used to meet the new acceptance criterion.

All other assumptions used in the new analyses are the same as those used in the previous analyses.

(3) The sequence of events for the new analysis.

Table 3 in Attachment 3 to this letter provides a sequence of events for the new locked rotor DNB analyses. The sequence of events is not cycle-specific. It is based on the system transient analysis, which is not repeated on a cycle-specific basis.

(4) The results of the analysis including plots of important parameters to show plant response and minimum DNBR.

Figures 1 through 5 in Attachment 3 to this letter show the plant response for a Unit 2 locked rotor event. These figures are also representative of the Unit 1 plant response.

Table 2 in Attachment 3 to this letter provides the state points for the minimum DNBR. to this letter provides tables showing available DNBR margins, allocations for the locked rotor analyses, other allocations, and remaining margins for the current Unit 1 and Unit 2 fuel cycles, Cycle 18 and Cycle 13, respectively. Attachment 4 contains information proprietary to Westinghouse Electric Company, LLC. Attachment 5 provides an affidavit setting forth the basis on which the proprietary information contained in Attachment 4 may be withheld from Page 2 to AEP:NRC:2075 Page 3 public disclosure pursuant to 10 CFR 2.790. Attachment 6 contains a non-proprietary version of.

ATTACHMENT 3 TO AEP:NRC:2075 LOCKED ROTOR RODS-IN-DNB TRANSIENT ANALYSIS INFORMATION FROM WESTINGHOUSE LETTER LTR-TA-02-157, DATED MAY 28, 2002 to AEP:NRC:2075 Introduction The non-loss-of-coolant-accident Locked Rotor Rods-in-DNB (departure from nucleate boiling) reload limit is revised to establish, for the locked rotor event, that no rods are in a DNB condition (0% Rods-in-DNB). The core coolant flow transient, overpressure, and overtemperature portion of the analysis (transient analysis) has not changed.

The DNB ratio (DNBR) calculation confirms, on a cycle-specific basis, that the 0% Rods-in-DNB limit continues to be met. The following information, pertaining to the Locked Rotor Rods-in-DNB transient analysis of record is provided.

Locked Rotor Rods-in-DNB Transient Analysis The DNB acceptance criterion for a locked rotor event at Donald C. Cook Nuclear Plant (CNP) is that no rods are in a DNB condition (no rods-in-DNB).

DNBR calculations confirm, on a cycle-specific basis, that the no-rods-in-DNB limit is met. Results of the cycle specific analysis for CNP Unit 1 Cycle 18 and Unit 2 Cycle 13 are provided in Attachment 4 to letter AEP:NRC:2075.

The Nuclear Regulatory Commission has requested additional information pertaining to the locked rotor rods-in-DNB transient analysis of record. Specifically, information on the transient analysis done to generate the state points for the DNB analysis that is not provided in the Updated Final Safety Analysis Report for CNP is requested.

The locked rotor transient analysis is performed using the LOFTRAN and FACTRAN computer programs. Several cases are analyzed to generate limiting overpressure, overtemperature (heat flux), and core coolant flow transient conditions during the event.

The core coolant flow conditions and peak heat flux are captured as state points in the Reload Safety Analysis Checklist and do not normally change every cycle. A Reload Safety Evaluation report is generated for each cycle. As part of the Reload Safety Evaluation, locked rotor transient analysis state points and cycle-specific core design parameters are input into the THINC computer program. THINC calculates a cycle-specific minimum DNBR for comparison with the DNB acceptance criterion.

The DNB analysis is performed using the Revised Thermal Design Procedure (RTDP) described in WCAP-1 1397-P-A.

The initial conditions used as input to LOFTRAN and FACTRAN are provided in Table 1 for each unit. Performing the Unit 2 analysis at an Nuclear Steam Supply System thermal power of 3608 MWt is conservative relative to its licensed power of 3411 MWt.

The state points calculated by LOFTRAN and FACTRAN, including core coolant flow conditions and peak core heat flux during the transient, are provided in Table 2. The sequence of events for the locked rotor transient analysis is presented in Table 3. Figures 1 through 5 illustrate the transient response during the Locked Rotor DNB analyzed event, which include power, pressure, temperature, and flow conditions versus time. These figures from the Unit 2 analysis of record are also representative of the transient conditions generated in the Unit 1 locked rotor analysis of record.

Page I

Table 1 Summary of Locked Rotor - Rods-in-DNB Analysis Initial Conditions Initial NSSS Initial Core Temperature Reactivity Coefficients Assumed Thermal Average Heat Reactor Vessel Pressurizer Vessel Core Moderator Doppler Power Power Output Flux Coolant Flow Pressure Average Inlet Temp Coef Coef (Mwt)

(Btu/hr-ft)

(gpm)

(psia)

(-F (LF)

(pcm/OF)

(pcm/%power)**

Unit 1 3270 207827 339,100 2100 576.3 544.0 Maximum Maximum FSAR Table 14.1-3

(+5)

(-19.4 + 0.065Q)

Unit 2 3608 207414 366,400.

2100 581.3 548.6 MaxImum Maximum

(+5)

(-19.4+0.07176Q)

FSAR Table 14.1.0-2*

Notes:

• Similar to Peak Pressure case without uncertainty application (RTDP Methodology)

• • Q Is In % power

Table 2 Summaryof Locked Rotor - Rods-in-DNB Analysis Statepoint Conditions Unit 1 Statepoint Results Initial Conditions Core power level, MWt 3250 Initial core heat flux, Btu/hr-ftz 207827 Initial core mass flow rate, gpm 339100 Statepoint Conditions Statepoint time, seconds 2.6 Core heat flux, fraction of initial 1.0280 Core mass flow rate, fraction of initial 0.6921 Pressurizer pressure, psia 2100 Core inlet temperature, 'F 544.0 Unit 2 Statepoint Results Initial Conditions Core power level, MWt 3588 Initial core heat flux, Btu/hr-ft2 207414 Initial core mass flow rate, gpm 366400 Statepoint Conditions Statepoint time, seconds 2.2 Core heat flux, fraction of initial 1.0288 Core mass flow rate, fraction of initial 0.70544 Pressurizer pressure, psia 2100 Core inlet temperature, OF 548.6

Table 3 Summary of Locked Rotor Rods-in-DNB Analysis Sequence of Events (includes time of maximum pressure and temperature)

Event SRotor in one pump locks Time (seconds)

Low reactor coolant flow trip setpoint reached 0.04 in faulted loop Rods begin to drop 1.04 Maximum percentage of rods-in-DNB 2.6 predicted Maximum RCS pressure occurs (Max 3.20 pressure/temp case)

Maximum clad temperature occurs(Max pressure/temp case) 3.49 I

• 1-Rotor in one oumo locks Low reactor coolant flow trip setpoint reached 0.02 in faulted loop Rods begin to drop 1.02 Maximum percentage of rods-in-DNB 2.2 predicted Maximum RCS pressure occurs (Max 3.10 pressure/temp case)

Maximum clad temperature occurs(Max pressure/temp case) 3.60 r

4-L __________________________

1. _________

Unit 1 Unit 2 0.0 N NO

Figure 1 Nuclear Power and Pressurizer Pressure vs. Time For the Locked Rotor Event 02 3

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Figure 2 Total Core Flow and Faulted Loop (Loop 4) Flow vs. Time For the Locked Rotor Event 1.4 1.2 t

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Figure 3 Core Average Temperature vs. Time For the Locked Rotor Event 2

3 4

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700 680 660 640 620 600 S80 660 S40 S20 Soo

Figure 4 Non-Faulted (Loop 1) and Faulted (Loop 4) Loop Temperature vs. Time For the Locked Rotor Event 700 680 660 640 620 600 680 S60 540 620 600 700 680 660 640 620 600 S80 660 S40 S20 S00 0

1 2

3 4

S 6

TIME ISEC)

S 6

7 8

9 10

[SEC) 7 8

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Figure 5 Average Channel Heat Flux vs. Time For the Locked Rotor Event (taken from the overtemperature case) 1.4 1.2 a:g z

0 z

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ATTACHMENT 5 TO AEP:NRC:2075 AFFIDAVIT SETTING FORTH THE BASIS ON WHICH INFORMATION CONTAINED IN ATTACHMENT 4 MAY BE WITHHELD FROM PUBLIC DISCLOSURE

CAW-02-1530 AFFIDAVIT COMMONWEALTH OF PENNSYLVANIA:

ss COUNTY OF ALLEGHENY:

Before me, the undersigned authority, personally appeared H. A. Sepp, who, being by me duly sworn according to law, deposes and says that he is authorized to execute this Affidavit on behalf of Westinghouse Electric Company LLC ("Westinghouse"), and that the averments of fact set forth in this Affidavit are true and correct to the best of his knowledge, information, and belief:

I.O

,. 4*

4U2N H. A. Sepp, Manager Regulatory and Licensing Engineering Sworn to and subscribed before me this Oq day of _

_,2002 K ay E. Gongaware, Notar Publi Monroeville soo rey Ii Mycomssion oep'ires' Feb. 7, 200.5 member, Pe-nn~svivaniq soitno oalsA P0-0 AgEP-32-109 S.

Page 30 of 371

/cm/0253S.doc

CAW-02-1530 (1)

I am Manager, Regulatory and Licensing Engineering, in Nuclear Services, Westinghouse Electric Company LLC ("Westinghouse"), and as such, I have been specifically delegated the function of reviewing the proprietary information sought to be withheld from public disclosure in connection with nuclear power plant licensing and rule making proceedings, and am authorized to apply for its withholding on behalf of the Westinghouse Electric Company LLC.

(2)

I am making this Affidavit in conformance with the provisions of 10CFR Section 2.790 of the Commission's regulations and in conjunction with the Westinghouse application for withholding accompanying this Affidavit.

(3)

I have personal knowledge of the criteria and procedures utilized by the Westinghouse Electric Company LLC in designating information as a trade secret, privileged or as confidential commercial or financial information.

(4)

Pursuant to the provisions of paragraph (b)(4) of Section 2.790 of the Commission's regulations, the following is furnished for consideration by the Commission in determining whether the information sought to be withheld from public disclosure should be withheld.

(i)

The information sought to be withheld from public disclosure is owned and has been held in confidence by Westinghouse.

(ii)

The information is of a type customarily held in confidence by Westinghouse and not customarily disclosed to the public. Westinghouse has a rational basis for determining the types of information customarily held in confidence by it and, in that connection, utilizes a system to determine when and whether to hold certain types of information in confidence. The application of that system and the substance of that system constitutes Westinghouse policy and provides the rational basis required.

Under that system, information is held in confidence if it falls in one or more of several types, the release of which might result in the loss of an existing or potential competitive advantage, as follows:

/cm/0253S.doc AEP-02-109 Page 31 of 37 2

CAW-02-1530 (a)

The information reveals the distinguishing aspects of a process (or component, structure, tool, method, etc.) where prevention of its use by any of Westinghouse's competitors without license from Westinghouse constitutes a competitive economic advantage over other companies.

(b)

It consists of supporting data, including test data, relative to a process (or component, structure, tool, method, etc.), the application of which data secures a competitive economic advantage, e.g., by optimization or improved marketability.

(c)

Its use by a competitor would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing a similar product.

(d)

It reveals cost or price information, production capacities, budget levels, or commercial strategies of Westinghouse, its customers or suppliers.

(e)

It reveals aspects of past, present, or future Westinghouse or customer funded development plans and programs of potential commercial value to Westinghouse.

(f)

It contains patentable ideas, for which patent protection may be desirable.

There are sound policy reasons behind the Westinghouse system which include the following:

(a)

The use of such information by Westinghouse gives Westinghouse a competitive advantage over its competitors. It is, therefore, withheld from disclosure to protect the Westinghouse competitive position.

/cm/0253S.doc AEP-02-109 Page 32 of 37 3

CAW-02-1530 (b)

It is information which is marketable in many ways. The extent to which such information is available to competitors diminishes the Westinghouse ability to sell products and services involving the use of the information.

(c)

Use by our competitor would put Westinghouse at a competitive disadvantage by reducing his expenditure of resources at our expense.

(d)

Each component of proprietary information pertinent to a particular competitive advantage is potentially as valuable as the total competitive advantage. If competitors acquire components of proprietary information, any one component may be the key to the entire puzzle, thereby depriving Westinghouse of a competitive advantage.

(e)

Unrestricted disclosure would jeopardize the position of prominence of Westinghouse in the world market, and thereby give a market advantage to the competition of those countries.

(f)

The Westinghouse capacity to invest corporate assets in research and development depends upon the success in obtaining and maintaining a competitive advantage.

(iii)

The information is being transmitted to the Commission in confidence and, under the provisions of 10CFR Section 2.790, it is to be received in confidence by the Commission.

(iv)

The information sought to be protected is not available in public sources or available information has not been previously employed in the same original manner or method to the best of our knowledge and belief.

(v)

The proprietary information sought to be withheld in this submittal is that which is appropriately marked in letter CAA-02-96, Revision 2 (Proprietary), May 2002 for D. C. Cook Units 1 and 2 being transmitted by the American Electric Company letter and Application for Withholding Proprietary Information from Public Disclosure, to the Document Control Desk, Attention Mr. Samuel J. Collins. The

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CAW-02-1530 proprietary information as submitted for use by American Electric Company for D. C. Cook Units 1 and 2 is expected to be applicable in other licensee submittals in response to certain NRC requests for information to support the locked rotor rods-in-DNB analysis for D. C. Cook Unit 1 or 2.

This information is part of that which will enable Westinghouse to:

(a)

Justify no rods-in-DNB for the locked rotor analysis.

(b)

Assist the customer to respond to NRC requests for information.

Further this information has substantial commercial value as follows:

(a)

Westinghouse plans to sell the use of similar information to its customers for purposes of meeting NRC requirements for licensing documentation.

(b)

Westinghouse can sell support and justification for no rods-in-DNB for the locked rotor analysis.

Public disclosure of this proprietary information is likely to cause substantial harm to the competitive position of Westinghouse because it would enhance the ability of competitors to provide similar support documentation and licensing defense services for commercial power reactors without commensurate expenses. Also, public disclosure of the information would enable others to use the information to meet NRC requirements for licensing documentation without purchasing the right to use the information.

The development of the technology described in part by the information is the result of applying the results of many years of experience in an intensive Westinghouse effort and the expenditure of a considerable sum of money.

In order for competitors of Westinghouse to duplicate this information, similar technical programs would have to be performed and a significant manpower AEP-02-109 Page 34 of 37

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CAW-02-1530 effort, having the requisite talent and experience, would have to be expended for developing testing and analytical methods and performing tests.

Further the deponent sayeth not.

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PROPRIETARY INFORMATION NOTICE Transmitted herewith are proprietary and/or non-proprietary versions of documents furnished to the NRC in connection with requests for generic and/or plant-specific review and approval.

In order to conform to the requirements of 10 CFR 2.790 of the Commission's regulations concerning the protection of proprietary information so submitted to the NRC, the information which is proprietary in the proprietary versions is contained within brackets, and where the proprietary information has been deleted in the non-proprietary versions, only the brackets remain (the information that was contained within the brackets in the proprietary versions having been deleted). The justification for claiming the information so designated as proprietary is indicated in both versions by means of lower case letters (a) through (f) contained within parentheses located as a superscript immediately following the brackets enclosing each item of information being identified as proprietary or in the margin opposite such information. These lower case letters refer to the types of information Westinghouse customarily holds in confidence identified in Sections (4)(ii)(a) through (4)(ii)(f) of the affidavit accompanying this transmittal pursuant to 10 CFR 2.790(b)(1).

AEP-02-109 Page 36 of 37

COPYRIGHT NOTICE The reports transmitted herewith each bear a Westinghouse copyright notice. The NRC is permitted to make the number of copies of the information contained in these reports which are necessary for its internal use in connection with generic and plant-specific reviews and approvals as well as the issuance, denial, amendment, transfer, renewal, modification, suspension, revocation, or violation of a license, permit, order, or regulation subject to the requirements of 10 CFR 2.790 regarding restrictions on public disclosure to the extent such information has been identified as proprietary by Westinghouse, copyright protection notwithstanding. With respect to the non-proprietary versions of these reports, the NRC is permitted to make the number of copies beyond those necessary for its internal use which are necessary in order to have one copy available for public viewing in the appropriate docket files in the public document room in Washington, DC and in local public document rooms as may be required by NRC regulations if the number of copies submitted is insufficient for this purpose.

Copies made by the NRC must include the copyright notice in all instances and the proprietary notice if the original was identified as proprietary.

AEP-02-109 Page 37 of 37

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ATTACHMENT 6 TO AEP:NRC:2075 WESTINGHOUSE PROPRIETARY CLASS 3 Donald C. Cook Nuclear Plant Departure From Nucleate Boiling Ratio (DNBR)

Margins and Allocations for Revised Thermal Design Procedure Analyses (Locked Rotor) from Westinghouse Letter CAA-02-96, Revision 2, dated May 29, 2002 (a, b, c) 4 4

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