AEP-NRC-2014-42, Response to Request for Additional Information on the Application for Amendment to Restore Normal Reactor Coolant System Pressure and Temperature Consistent with Previously Licensed Conditions.

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Response to Request for Additional Information on the Application for Amendment to Restore Normal Reactor Coolant System Pressure and Temperature Consistent with Previously Licensed Conditions.
ML14181A537
Person / Time
Site: Cook American Electric Power icon.png
Issue date: 06/05/2014
From: Gebbie J
Indiana Michigan Power Co
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
AEP-NRC-2014-42, TAC MF2916
Download: ML14181A537 (34)
v • d • e


Text

z INDIANA Indiana Michigan Power MICHIGAN Cook Nuclear Plant POWEMR* One Cook Place Bridgman, MI 49106 A unitofAmerican Electric Power IndianaMichiganPower.com June 5, 2014 AEP-NRC-2014-42 10 CFR 50.90 10 CFR 50.36 Docket No.: 50-315 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC, 20555-0001 Donald C. Cook Nuclear Plant Unit 1 Response to "Request for Additional Information on the Application for Amendment to Restore Normal Reactor Coolant System Pressure and Temperature Consistent with Previously Licensed Conditions (TAC No. MF2916)," Dated May 6, 2014

References:

1. Letter from J. P. Gebbie, Indiana Michigan Power Company (I&M), to U. S. Nuclear Regulatory Commission (NRC) Document Control Desk, "Donald C. Cook Nuclear Plant Unit 1 Docket No. 50-315, License Amendment Request Regarding Restoration of Normal Reactor Coolant System Operating Pressure and Temperature Consistent with Previously Licensed Conditions," dated October 8, 2013, Agencywide Documents Access and Management System (ADAMS) Accession Number ML13283A121.
2. Letter from T. J. Wengert, NRC, to L. J. Weber, I&M, "Donald C. Cook Nuclear Plant, Unit 1, Request for Additional Information on the Application for Amendment to Restore Normal Reactor Coolant System Pressure And Temperature Consistent with Previously Licensed Conditions (TAC No. MF2916)," dated March 31, 2014, ADAMS Accession Number ML14066A311.
3. Letter from J. P. Gebbie, I&M, to NRC Document Control Desk, "Donald C. Cook Nuclear Plant Unit 1, Response to 'Request for Additional Information on the Application for Amendment to Restore Normal Reactor Coolant System Pressure and Temperature Consistent with Previously Licensed Conditions (TAC No. MF2916),"' dated April 29, 2014, ADAMS Accession Number ML14121A422.
4. Letter from T. J. Wengert, NRC, to L. J. Weber, I&M, "Donald C. Cook Nuclear Plant, Unit 1 - Request for Additional Information on the Application for Amendment to Restore Normal Reactor Coolant System Pressure and Temperature Consistent with Previously Licensed Conditions (TAC NO. MF2916)," dated May 6, 2014, ADAMS Accession Number ML14099A450.

PROPRIETARY INFORMATION Enclosure 6 to this Letter contains proprietary information.

Withhold from public disclosure under 10 CFR 2.390.

Upon removal of Enclosure 6, this Letter is decontrolled.

U. S. Nuclear Regulatory Commission AEP-NRC-2014-42 Page 2 By letter dated October 8, 2013 (Reference 1), Indiana Michigan Power Company (I&M) submitted an application for a license amendment to restore the normal reactor coolant system operating pressure and temperature consistent with previously licensed conditions for the Donald C. Cook Nuclear Plant, Unit 1. The U.S. Nuclear Regulatory Commission (NRC) staff provided a Request for Additional Information (RAI) (Reference 2) to complete the review of Reference 1. I&M responded to Reference 2 by Reference 3. By letter dated May 6, 2014, the NRC provided an additional RAI (Reference 4) to complete the review of Reference 1. This submittal provides I&M's response to the RAIs contained in Reference 4, with the exception of RAIs Containment and Ventilation Systems Branch (SCVB) RAI-3(c), SCVB RAI-9(a), and SCVB RAI-9(b). I&M plans to respond to these items by July 3, 2014. to this letter provides an affirmation statement. Enclosure 2 provides a cross reference for I&M's response to the NRC RAI. Enclosure 3 provides responses to RAIs: SCVB RAI-5(b),

SCVB RAI-10(a, b), SCVB RAI-11(a, b, c, d, e), Radiation Protection and Consequence Branch (ARCB) RAI-1(a, b), ARCB RAI-2(a, b), Electrical Engineering Branch (EEEB) RAI-1, EEEB RAI-2, EEEB RAI-3, EEEB RAI-4, and EEEB RAI-5. Enclosure 4 provides an "Application for Withholding Proprietary Information from Public Disclosure." Enclosure 5 provides non proprietary responses to RAIs: SCVB RAI-1 (a, b), SCVB RAI-2(a, b, c), SCVB RAI-3(a, b), SCVB RAI-4 (a, b, c, d, e), SCVB RAI-5(a), SCVB RAI-6, SCVB RAI-7, SCVB RAI-8, SCVB RAI-12, Nuclear Performance and Code Review Branch (SNPB) RAI-1, SNPB RAI-2, and SNPB RAI-3(a, b). Enclosure 6 provides the response to NRC RAI: SCVB RAI-4(f) and includes proprietary information.

This letter contains no new or revised commitments. Should you have any questions, please contact Mr. Michael K. Scarpello, Regulatory Affairs Manager, at (269) 466-2649.

Sincerely, Joel P. Gebbie Site Vice President JMT/amp

Enclosures:

1. Affirmation
2. Donald C. Cook Nuclear Plant Unit 1 Cross

Reference:

Request for Additional Information -

Response Enclosure

3. Responses to SCVB RAI-5(b), SCVBRAI-10(a,b), SCVB RAI-11(a,b,c,d,e), ARCB RAI-1(a,b), ARCB RAI-2(a,b), EEEB RAI-1, EEEB RAI-2, EEEB RAI-3, EEEB RAI-4, and EEEB RAI-5
4. Westinghouse Letter, CAW-14-3954, Application for Withholding Proprietary. Information from Public Disclosure, dated May 28, 2014 PROPRIETARY INFORMATION Enclosure 6 to this Letter contains proprietary information.

Withhold from public disclosure under 10 CFR 2.390.

Upon removal of Enclosure 6, this Letter is decontrolled.

U. S. Nuclear Regulatory Commission AEP-NRC-2014-42 Page 3

5. Attachment #2 (NP-Attachment) of Westinghouse Letter, LTR-PL-14-22, "Westinghouse Responses to NRC, "Donald C. Cook Nuclear Plant Unit 1 - Request for Additional Information on the Application for Amendment to Restore Normal Reactor Coolant System Pressure and Temperature Consistent with Previously Licensed Conditions (TAC No.

MF2916)," dated May 28, 2014

6. Westinghouse Letter, LTR-PL-14-22 Cover Letter and Attachment #1 (P-Attachment),

Westinghouse Responses to NRC, "Donald C. Cook Nuclear Plant Unit I - Request for Additional Information on the Application for Amendment to Restore Normal Reactor Coolant System Pressure and Temperature Consistent with Previously Licensed Conditions (TAC No. MF2916)," dated May 28, 2014 c: J. T. King, MPSC MDEQ - RMD/RPS NRC Resident Inspector C. D. Pederson, NRC Region III T. J. Wengert, NRC Washington DC A. J. Williamson, AEP Ft. Wayne, w/o enclosure PROPRIETARY INFORMATION Enclosure 6 to this Letter contains proprietary information.

Withhold from public disclosure under 10 CFR 2.390.

Upon removal of Enclosure 6, this Letter is decontrolled.

ENCLOSURE 1 TO AEP-NRC-2014-42 AFFIRMATION I, Joel P. Gebbie, 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 U. S. 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 Joel P. Gebbie Site Vice President SWORN TO AND SUBSCRIBED BEFORE ME My Commission Expires OTýPATRICIA 1A-.P.. ,,.--,. ANN EDIE SO -

C of Umniu Conins~u*

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Enclosure 2 to AEP-NRC-2014-42 Donald C. Cook Nuclear Plant Unit I Cross

Reference:

Request for Additional Information - Response Enclosure The Request for Additional Information (RAI) responses are contained in separate Enclosures.

Table 1 below provides a cross reference between the RAI and the Enclosure providing the RAI Response.

I SCVB RAI-1 a, b Enclosure 5, Non-Proprietary SCVB RAI-2 a, b, c Enclosure 5, Non-Proprietary SCVB RAI-3 a, b Enclosure 5, Non-Proprietary SCVB RAI-3 c July 3, 2014 SCVB RAI-4 a, b, c, d, e Enclosure 5, Non-Proprietary SCVB RAI-4 f Enclosure 6, Proprietary SCVB RAI-5 a Enclosure 5, Non-Proprietary SCVB RAI-5 b Enclosure 3 SCVB RAI-6 Enclosure 5, Non-Proprietary SCVB RAI-7 Enclosure 5, Non-Proprietary SCVB RAI-8 Enclosure 5, Non-Proprietary SCVB RAI-9 a, b July 3, 2014 SCVB RAI-10 a, b Enclosure 3 SCVB RAI-1 1 a, b, c, d, e Enclosure 3 SCVB RAI-12 Enclosure 5, Non-Proprietary ARCB RAI-1 a, b Enclosure 3 ARCB RAI-2 a, b Enclosure 3 EEEB RAn1 B Encoure EEEB RAI-1 Enclosure 3 EEEB RAI-2 Enclosure 3 EEEB RAI-3 Enclosure 3 EEEB RAI-4 Enclosure 3 EEEB RAI-5 Enclosure 3 SNPB RAI-1 Enclosure 5, Non-Proprietary SNPB RAI-2 Enclosure 5, Non-Proprietary SNPB RAI-3 a, b Enclosure 5, Non-Proprietary

ENCLOSURE 3 TO AEP-NRC-2014-42 Responses to SCVB RAI-5(b), SCVB RAI-10(a,b), SCVB RAI-11(a,b,c,d,e),

ARCB RAI-1(a,b), ARCB RAI-2(a,b), EEEB RAI-1, EEEB RAI-2, EEEB RAI-3, EEEB RAI-4, and EEEB RAI-5 to AEP-NRC-2014-42 Page 2 Table of Contents Containment and Ventilation Systems Branch (SCVB) .................................... 3 SCVB RAI-5(b) ............................................................................................. 3 SCVB RAI-10 ............................................................................................... 4 SCVB RAI-11 ............................................................................................... 5 Radiation Protection and Consequence Branch (ARCB) ................................. 9 ARCB RAI-1 ................................................................................................. 9 ARCB RAI-2 .................................................................................................. 9 Electrical Engineering Branch (EEEB) .......................................................... 10 EEEB RAI-1 .............................................................................................. 11 EEEB RAI-2 ............................................................................................... 13 EEEB RAI-3 ............................................................................................... 14 EEEB RAI-4 ............................................................................................... 15 EEEB RAI-5 ............................................................................................... 15 Re fe re n ce s .......................................... ................................................ 2 0 to AEP-NRC-2014-42 Page 3 Containment and Ventilation Systems Branch (SCVB)

SCVB RAI-5(b)

Reference 1, Enclosure 6, Section 5.4.2.6 states, in part The evaluation of the long term LOCA M&E and peak containmentpressure is predicatedupon the continued applicationof the operability assessment supporting NSAL- 11-5 (Reference 3), in conjunction with the AOR.

(b) Describe the operability assessment supportingNSAL-11-5 (Reference 2), including the impact of the corrected M&E release (due to the issues identified in NSAL-1 1-5), on the containment response AOR.

Response

Donald C. Cook Nuclear Plant (CNP) has taken a multi-staged approach to address the issues in Nuclear Safety Advisory Letter (NSAL)-1 1-5. The primary assessment for continued operability is the 6 pounds per square inch (psi) generic margin contained within the Westinghouse WCAP-10325 loss of coolant accident (LOCA) mass and energy (M&E) and Containment response analysis. The generic analysis margin, listed by Westinghouse in NSAL-1 1-5, consists of a conservative non-mechanistic calculation and input I initial condition assumptions. A mechanistic calculation with realistic input assumptions would provide a calculated peak pressure more than 6 psi lower. This is more than enough to offset the 2.31 psi CNP specific penalty due to NSAL-1 1-5. CNP decided that switching to the Westinghouse WCOBRA/TRAC LOCA M/E and Containment response methodology (generic topical WCAP-17721 currently under U. S. Nuclear Regulatory Commission (NRC) review) would provide the best resolution to the NSAL-1 1-5 issues. Indiana Michigan Power Company (I&M) and Westinghouse are currently working to reanalyze the LOCA M/E and Containment response using the WCOBRA/TRAC methodology. After approval of WCAP-17721, the CNP plant specific WCOBRAITRAC analysis will be submitted for review.

Additionally, to supplement the generic operability evaluation, described above, Westinghouse performed a CNP-specific LOCA containment peak pressure sensitivity analysis. The sensitivity analysis found that sufficient ice is already contained in the ice condenser to also offset the errors. This ice is above the Technical Specification minimum considered in the analysis of record (AOR). SCVB RAI-5(b), Table 1, provides the calculated peak pressure results of the Westinghouse sensitivities. The ice mass is tracked using procedural guidance to ensure that it is available each cycle.

to AEP-NRC-2014-42 Page 4 SCVB RAI-5(b), Table 1: CNP Calculated Peak Pressure Sensitivities Peak Pressure (pounds per square inch gauge (psig))

AOR 11.75 AOR + NSAL-11-5 Errors 14.06 AOR + NSAL-11-5 w/Additional Ice Mass 11.96 Recently, Westinghouse issued NSAL-14-2 which identified additional errors in the LOCA M/E and Containment response analysis relative to the modeling of the specific heat of the steam generator (SG) thick metal mass. The penalty has been assessed against the generic margin communicated in NSAL-1 1-5 and excess ice contained within the condenser and both sources of margin continue to be sufficient to offset the combined NSAL-14-2 and NSAL-1 1-5 penalties.

SCVB RAI-1O Please describe the impact of the changes in M&E release during LOCA and main steam line break (MSLB) accident to the following containment analyses:

(a) Sump water temperature response (b) Net Positive Suction Head (NPSH) analysis for Emergency Core Cooling System (ECCS) and the containment heat removal pumps that draw suction from the containment sump in the post-accidentrecirculationmode.

Response (a):

The maximum recirculation sump water temperature value used as a design input to other CNP Unit 1 accident analyses is not itself based on an analysis. Instead, the temperature used as design input reflects measurements taken during Waltz Mill Facility ice condenser testing performed in 1974, and as such, is not impacted by specific changes in M&E releases associated with returning the Unit 1 reactor coolant system (RCS) to normal operating pressure/normal operation temperature (NOP/NOT) conditions.

CNP Unit 1 Updated Final Safety Analysis Report (UFSAR), Revision 25, Chapter 14, Section 14.3.4.1.3.1.3, Peak Containment Pressure Transient, Item 4, identifies an ice condenser drain temperature of 190 degrees (0) Fahrenheit (F) as the sump temperature value used for containment pressure analysis, citing WCAP-8282 as a reference. WCAP-8282 (Reference 1) and WCAP-8282, Addendum 1 (Reference 2) document test values measured at the Waltz Mill Facility. Per Page 23 of WCAP-8282, Addendum 1:

to AEP-NRC-2014-42 Page 5 "The measured temperature of the condensate and drain at the end of blowdown is 220°F, but additional water drains from the ice bed after blowdown is complete. The resultant condensate and drain temperature after this drainage is 190'F."

Because the initiation of switchover of ECCS and containment spray (CTS) pumps from the refueling water storage tank to the recirculation sump occurs after biowdown is complete, typical ice condenser design basis containment pressure response analysis (post-blowdown calculation) assume a maximum sump temperature of 190'F.

However, subsequent calculations of sump temperature transients show sump temperature at the initiation of the switchover to cold leg recirculation to be <170 0 F, which confirms that the assumption of 1 90°F maximum sump temperature is conservative.

Response (b):

To ensure the calculated NPSH-required value is conservatively maximized under all conditions, the CNP Unit 1 NPSH AOR (Reference 3) for the ECCS pumps and CTS pumps uses bounding design inputs rather than relying on specific containment conditions associated with specific M&E releases or pump flow rates associated with specific flow requirement scenarios. These bounding design inputs include use of negative containment pressure (12.9 pounds per square inch absolute (psia)), maximum recirculation sump temperature (190°F), and conservative flow values during the cold leg recirculation and hot leg recirculation phases of the design basis accident.

Consequently, the ECCS and CTS pumps NPSH AOR is not affected by changes in M&E release associated with returning the Unit 1 RCS to NOP/NOT conditions.

SCVB RAI-1I Considering the changes (due to change in NOP/NOT) in the M&E release during LOCA and MSLB accident, please provide the following information for NRC staff review. Refer to SECY- 11-00 14 for more information.

(a) The value of the required NPSH (i.e., NPSHR) for the ECCS and the containment heat removal pumps that draw suction from the containment sump in the post-accident recirculationmode.

(b) The basis for the NPSHR, including the standardon which it is based. As an example, the NPSHR for ECCS and containment heat removal system pumps is commonly based on the Hydraulic Institute standard, to which the NPSHR is equal to the available NPSH determined in a factory test at the pump design flow with a three-percent drop in the total dynamic head.

(c) The uncertaintyin the factory tested value of NPSHR based on the actual site conditions.

(d) The minimum NPSH available in the proposed analysis at each pump inlet based on maximizing the sump water temperature along with maximizing the suction strainerhead loss based on Generic Safety Issue 191 resolution and maximizing piping head loss.

to AEP-NRC-2014-42 Page 6 (e) The minimum NPSH margin for each pump and its percentage, based on NPSHR with uncertainty added.

Response

As indicated in the response to SCVB RAI-10(b), because of the previous use of a conservative analysis approach, the NPSH AOR for the ECCS and CTS pumps is not impacted by specific changes in M&E releases associated with returning the Unit 1 RCS to NOP/NOT conditions.

Consequently, the current AOR (Reference 3) remains bounding. The responses to SCVB RAI-11 (a) through SCVB RAI-11 (e) reflect the current AOR (Reference 3).

The results of NPSH analyses for CNP were previously submitted to address Generic Letter (GL) 97-04, "Assurance of Sufficient Net Positive Suction Head for Emergency Core Cooling and Containment Heat Removal Pumps," and GL 2004-02, "Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors." The following public documents in Agencywide Documents Access Management System provide historical context and relevant technical information regarding CNP NPSH analysis:

" ML003681807 (Reference 5), AEP revised response to GL 97-04

" ML003710891 (Reference 6), Nuclear Reactor Regulation Safety Evaluation related to GL 97-04 response for CNP

  • ML101960128 (Reference 8), NRC Staff comments on GL 2004-02 response for CNP The stated purpose of SECY-11-0014 is to "resolve issues regarding the use of containment accident pressure (CAP) in analyzing pump performance in emergency core cooling systems and containment heat removal systems during postulated accidents." These issues are not relevant for CNP because the AOR does not credit CAP, but rather conservatively assumes a negative pressure (12.9 psia) during all phases of a design basis accident. The acceptance criteria in the AOR are consistent with the requirements of NRC Regulatory Guide 1.1, "Net Positive Suction Head for Emergency Core Cooling and Containment Heat Removal System Pumps" (Reference 4), specifically stating that, "Results of this analysis are acceptable if adequate NPSH (NPSHa > NPSHr) is provided to system pumps, assuming maximum expected temperatures of pumped fluids and no increase in containment pressure from that present prior to postulated loss of coolant accidents."

Response (a):

See Table RAI-SCVB-1 1-1 for the requested information taken from the current AOR.

to AEP-NRC-2014-42 Page 7 Response (b):

A response to this question was previously provided by AEP in a submittal related to resolution of GL 2004-02 (Reference 7.3, NRC Information Item 3.g.3, p. 233). The text of the previous response, which remains valid, is provided below:

"The required NPSH values are specified by the individual pump manufacturers, and are shown on the certified pump performance curves. Pump manufacturers typically set the required NPSH using a test loop where pressure is lowered in the suction source until a drop of 3% of the total head is measured. The available NPSH at which the 3% head drop occurs is then defined as the required NPSH for that flow. This method of required NPSH determination is used unless an alternate method was specified by the customer. Since I&M did not specify an alternate method to calculate required NPSH, it is assumed that the 3% head drop method was utilized."

Response (c):

SECY-11-0014, Section 6.0, provides draft Staff Guidance regarding treatment of uncertainties for the use of CAP in analyzing pump performance in ECCS and containment heat removal systems during postulated accidents. Included in the draft guidance is a qualitative discussion of potential uncertainty differences in NPSHr of a pump installed in the field versus a pump tested at the pump vendor's facility.

The CNP NPSH AOR is not impacted by the proposed restoration of NOP/NOT conditions.

Additionally, the AOR does not credit CAP. The AOR uses conservative (i.e., bounding) inputs and assumptions for the design basis accident case to produce bounding results, but does not quantify specific uncertainties.

Response (d):

See Table RAI-SCVB-1 1-1 for the requested information taken from the current AOR.

Response (e):

See Table RAI-SCVB-11-1 for the requested information taken from the current AOR.

to AEP-NRC-2014-42 Page 8 Table RAI-SCVB-11-1: Cook Unit 1 ECCS and CTS Pumps NPSH Margin Minimum NSPS Margin Note 1 Limiting Recirculation Case from MD-01-ECCS-004-N Note 2 Pump Designation NPSHr NPSHa Margin Description (ft.) (ft.) (ft.-) Ratio Case No.

1-PP-50E East Centrifugal Charging 13.6 184.2 170.6 13.54 Rev 4, Case 10 Dual train with East RHR pump failure 1-PP-50W West Centrifugal Charging 13.6 184.3 170.7 13.55 Rev 4, Case 10 Dual train with East RHR pump failure 1-PP-26N North Safety Injection 16.2 159.1 142.9 9.82 Rev 4, Case 11 Dual train with West RHR pump failure 1-PP-26S South Safety Injection 16.6 157.9 141.3 9.51 Rev 4, Case 11 Dual train with West RHR pump failure 1-PP-35E East RHR 16.3 30.7 14.4 1.88 Rev 4, Case 11 Dual train with West RHR pump failure 1-PP-35W West RHR 16.6 27.6 11.0 1.66 Rev 4, Case 10 Dual train with East RHR pump failure 1-PP-9E East Containment Spray 15.0 26.9 11.9 1.79 Rev 6, Case 23 Dual train with West RHR pump failure 1-PP-9W West Containment Spray 14.8 27.8 13.0 1.88 Rev 6, Case 8 Dual train with East RHR pump failure Note 1 Slight differences between values in this table and similar information tabulated in Ref. 7 reflect recent updates to the analysis-of-record associated with planned replacement of the CTS heat exchangers.

Revision 2 (1/20/00) - Provides basis for information submitted in Ref. 5 Revision 3 (2/29/08) - Calculation Impact Addendum to assess the impact resulting from EC-47743 (additional check valves downstream of RHR pump minimum flow connection and reconfigured cross-tie downstream of the RHR heat exchangers)

Revision 4 (8/22/08) - Calculation Impact Addendum to incorporate revised recirculation sump level per EC-48234 (new Note 2 Containment Recirculation Sump Licensing Basis per GL 2004-02); provides basis for information previously submitted in Ref. 7 and also the ECCS pump information included in this RAI response Revision 5 (6/25/13) - Calculation Impact Addendum to reflect planned replacement of Unit I West CTS heat exchanger Revision 6 (11/11/13) - Calculation Impact Addendum to reflect planned replacement of Unit 1 East CTS heat exchanger; provides basis for CTS pump information in this RAI response to AEP-NRC-2014-42 Page 9 Radiation Protection and Consequence Branch (ARCB)

ARCB RAI-1 On page 1 of Enclosure 8 to the October 8, 2013, application, Section 2.0 lists the proposed revised parameters for the proposed amendment. Please provide additional Information regarding the effect these revisions will have on each of the radiologicaldesign basis accident analyses.

(a) For the revised analyses, were any changes made to the methodologies that are in the current analyses of record?

(b) What are the revised calculated dose values for the Exclusion Area Boundary, Low PopulationZone, and Control Room?

Response (a):

No methodology changes were made to the current analyses of record.

Response (b):

The results of the revised Large Break (LB) LOCA radiological dose analyses, which were the only analyses that required sensitivity runs du- to NOP/NOT-related input parameter changes, are found below:

Control Room: 4.26 rem total effective dose equivalent (TEDE)

Offsite (exclusion area boundary (EAB)): 23.6 rem / 2.64 rem (Thyroid / Whole Body)

Offsite (low-population zone (LPZ)): 176 rem / 0.864 rem (Thyroid I Whole Body)

The above results maintain greater than 90 percent (%) of the margin to the applicable acceptance criteria:

Control Room: 5 rem TEDE (10 CFR 50.67)

EAB: 300 rem / 25 rem (Thyroid I Whole Body) (RG 1.195)

LPZ: 300 rem 1 25 rem (Thyroid I Whole Body) (RG 1.195)

ARCB RAI-2 In Section 3.0 of Enclosure 8 to the October 8, 2013, application, it states that the proposed revised parameters will affect the previous maximum and minimum RCS liquid mass values. It also states that these values are used in same of the dose consequence analyses.

(a) What are the revised values for the maximum and minimum RCS liquid mass?

to AEP-NRC-2014-42 Page 10 (b) Which dose consequence analyses are being referred to in the above statement and how does that affect the resulting dose values for those analyses?

Response (a):

The Unit 1 and Unit 2 RCS liquid masses utilized in the current analyses of record are bounding when compared to the revised Unit 1 liquid mass at NOP/NOT conditions. The dose consequence analyses are common to both units. For the AOR, the lower-bound liquid mass of 2.2649E+08 grams is representative of Unit 2, and the upper-bound liquid mass of 2.3874E+08 grams is representative of Unit 1. At NOP/NOT conditions, the Unit 1 liquid mass is reduced to approximately 2.3696E+08 grams due to the decrease in liquid density caused by NOP/NOT.

Since this value is greater than the lower-bound liquid mass (Unit 2) and less than the upper-bound liquid mass (pre-NOP/NOT Unit 1) utilized in the AOR, the previous values remain bounding.

Response (b):

As noted in ARCB RAI-2(a) above, the lower-bound and upper-bound liquid mass values utilized in the dose consequence analyses of record remain bounding for the CNP Unit 1 Return to NOP/NOT Program. Therefore, no analyses require revision as a consequence of this parameter. The upper-bound liquid mass value is used in the LOCA dose consequence analyses to conservatively increase the RCS fission product inventory available for release to the environment through containment purge. The lower-bound liquid mass is utilized to determine conservative RCS radionuclide concentrations and is used in a wide range of dose consequence analyses such as the following scenarios: SG Tube Rupture, Main Steam Line Break (MSLB), Control Rod Ejection, and Locked Rotor. Again, all of these analyses are unaffected as the RCS masses utilized in the analyses of record remain bounding.

Electrical Engineering Branch (EEEB)

Background

The licensee proposes to implement a return to RCS NOP/NOT conditions for CNP Unit I by increasingthe current operating nominal full-power pressurizerpressure from 2100 psia to 2250 psia and increasing the current operating nominal full-power Tavg from 556 OF to 571 °F.

Implementation of the program is proposed to occur prior to CNP Unit I Cycle 26 startup (October2014).

The licensee states that the proposed change also revises the Containment Air Recirculation/Hydrogen Skimmer (CEQ) fan start time from "108 seconds < CEO fan delay <

132 seconds" to "270 seconds < CEO fan delay < 300 seconds" and revise containment spray system (CTS) actuation time delay from 115 seconds to 315 seconds.

The CTS design at CNP includes a time delay relay in the CTS pump start circuitry that is presently used to properly sequence the pump onto the emergency diesel generator (EDG) bus and prevent overloading of the diesel. To offset the adverse effects of the proposed increase in full power average RCS temperature on best estimate loss-of-coolant accident peak cladding to AEP-NRC-2014-42 Page 11 temperature (BELOCA-PCT), the setting of this time delay relay is increased to further delay CTS actuation following a Hi-Hi containment pressure signal. The new time delay setting continues to support proper EDG bus loading, but also results in a higher containmentpressure during large break loss of coolant accident RCS blowdown, which limits the rate of RCS mass release to containment and improves BELOCA-PCT results.

In order for the NRC staff to verify that there are no adverse effects on the EDGs and no environmental qualification changes of electrical equipment due to the proposed changes, please provide the following information:

EEEB RAI-1 Provide the loading sequence for each EDG at CNP Unit 1. In your response, please describe the changes that have been made to the EDG loading sequence and any changes in motor loads as a consequence of higher containmentpressure.

Response

The automatic start time for individual safety-related loads to be sequenced onto an emergency electrical bus following a loss of offsite power (LOOP) is determined by safety-related logic circuitry and may vary depending on the type of LOOP scenario that has occurred, i.e., LOOP without LOCA, LOOP with concurrent safety injection (SI), or LOOP with concurrent CTS.

The automatic starting sequence and nominal start times for affected equipment for each LOOP scenario are summarized in Table RAI-EEEB-1-1. The requested license amendment to allow restoring the Unit 1 RCS to NOP/NOT conditions does not affect the "LOOP without LOCA" and "LOOP with Safety Injection" columns because the actuation of safety-related equipment required for these scenarios remains unchanged. The "LOOP with CTS - Proposed" column of the table reflects the impact of the requested license amendment. Note that the "LOOP with SI" and "LOOP with CTS - Current" columns of the table reflect information provided in Section 8.4 of the CNP UFSAR, Revision 25.0.

to AEP-NRC-2014-42 Page 12 Table RAI-EEEB-1-1: CNP EDG Automatic Loading Sequence (Train A and Train B)

Nominal Start Time After LOOP Signal (seconds)

Equipment LOOP LOOP with CTS without without LOOP with LOPwtCT S

LOCA SI Current Proposed 600 Volt Safety-Related Load (block) 10 10 10 10 Centrifugal Charging Pump Note 1 13 13 13 SI Pump Note 1 17 17 17 RHR Pump Note 1 21 21 21 Component Cooling Pump 13 25 25 25 Essential Service Water Pump 17 30 30 30 Auxiliary Feed Water Pump 21 35 35 35 CTS Pump Note 1 Note 1 41 225 Non-Essential Service Water (NESW) 25 47 Note 2 Note 2 Pump Note I - Not Automatically Started Note 2 -The breaker for the NESW pump is sequenced closed but is then automatically tripped due to a load conservation interlock associated with the CTS signal. In either the Current or Proposed scenario, the NESW pump breaker would be closed at a nominal 47 seconds, but would have no impact on bus loading since the NESW pump logic would prevent it from starting. The CTS pump is the last load to start under either the current or the proposed scenario.

The only change to the current loading scenarios involves the "LOOP with CTS," in which the last load automatically started, i.e., the CTS pump, moves from a nominal 41 seconds out to a nominal 225 seconds.

To clarify the question regarding "changes in motor loads as a consequence of higher containment pressure," the containment design pressure is not affected by the proposed changes. The delay in CTS pump and hydrogen skimmer fan (CEQ) actuation allows crediting a higher value for minimum containment pressure during the early phases of a LBLOCA, which provides a benefit to the Best Estimate LOCA Peak Clad Temperature analysis.

Figure RAI-EEEB-1-1 shows the increase in credited backpressure in the time interval from 50 to 500 seconds. From the plot, the credited backpressure at the peak of this interval increases from 18.3 psia (3.6 psig) in the AOR to 19.5 psia (4.8 psig) in the proposed analysis. The credited minimum backpressure is well below the containment design pressure of 12 psig. As presented in Section 5.4.2.6 of Enclosure 6 of the original License Amendment Request (LAR),

the long-term peak containment pressure remains within the original containment design to AEP-NRC-2014-42 Page 13 pressure. Safety-related motor loads used in the CNP EDG evaluations bound maximum loading conditions consistent with maximum design performance of the driven equipment. The motor loads used in the analysis and evaluations are not dependent on the transient response of driven equipment. Therefore, there are no changes to analyzed motor loads as a consequence of the proposed changes.

EEEB RAI-2 Please provide a summary of your analysis and tests performed to show that the EDGs will continue to perform within the voltage and frequency requirements during the load sequencing and steady state operation after implementing the delayed CTS pump start. Please confirm whether the proposed delay in CTS pump start can result in overlapping of loads during EDG load sequencing if the containment Hi-Hi signal is generated after the permissive from the sequencertimer during LOCAs (for any break sizes).

Response

The adequacy of the CNP EDGs for automatic load sequencing of safety-related loads during LOOP events is demonstrated by dynamic simulation of EDG response based on a functional computer model that was benchmarked and tuned against EDG performance tests. The AOR takes into account not only the nominal load sequence starting times for each equipment load on an EDG, but also considers worst case stack-up of timer tolerances (i.e., evaluates the effect of combinations of latest and earliest start times for the various components). The calculation provides a bounding analysis that applies to the two Unit 1 EDGs and the two Unit 2 EDGs.

Based on the output of the simulation, each EDG will be able to start and accelerate its required loads during accident conditions with an assumed worst case LOOP scenario. Maximum dips in voltage or frequency during transient loading are demonstrated to be acceptable.

As discussed in the response to NRC RAI EEEB RAI-1, the only change to the current automatic loading scenarios involves the LOOP with concurrent CTS, in which the last load automatically started, i.e., the CTS pump, moves from a nominal 41 seconds out to a nominal 225 seconds. The calculation identifies that the worst case loading transient occurs when the time interval between two motor starts is minimized. Therefore, delaying the start of the last load is conservative and the results of the AOR remain valid.

The AOR considers that the containment pressure "Hi-Hi" signal is generated simultaneously with a SI signal, and as noted above, the CTS pump is the last load automatically added to the emergency bus (under both the current and proposed time delays). Having the containment pressure "Hi-Hi" signal generated at any time subsequent to the SI signal serves only to move the CTS pump start farther away in time from the previously started load. Consequently, the proposed delay in CTS pump start cannot result in overlapping of loads during EDG load sequencing.

Closeout activities for the engineering change package that implements NOP/NOT conditions for Unit 1 will update the AOR to acknowledge delayed actuation of CTS for Unit 1.

Although not directly included in EEEB RAI-2, in addition to extending the start time of the CTS pumps, the proposed change also extends the start time of the 600-volt Containment Air to AEP-NRC-2014-42 Page 14 Recirculation/CEQ fan motors. In the AOR for EDG loading, the 600-volt bus is energized upon EDG start and loads that are not process-driven are added immediately. Process-driven 600-volt motor loads, such as the CEQ fan motors that are actuated following a containment pressure "Hi" signal, are conservatively assumed to start simultaneously as a block load at the most critical time (based on lowest voltage) in the automatic sequence. This point occurs prior to starting the CTS pump. Therefore, the delay in automatically starting the CEQ fans has no impact on the EDG AOR.

EEEB RAI-3 Updated Final Safety Analysis Report Section 8.1.2 states:

The engineered safety feature (ESF) Loads are sequenced onto the Reserve Auxiliary Transformers (RA Ts), under accident conditions, using the same timing relays and sequence as used for the EDG sequencing. The RATs supply the reserve auxiliary power for both units, Under certain plant conditions and grid loading conditions, and with proper precautions and limitations, it is possible for either transformerNo. 4 or transformerNo. 5 to feed the entire plant auxiliary load.

Please provide a summary of your analysis to show that the RATs can provide the required voltage to ESF loads without actuating protective devices and meet the accident analysis assumptions.

Response

The Load Flow analysis for CNP is performed using Operation Technology, Inc.'s Electrical Transient Analyzer Program (ETAP). The Block Motor Start LOCA Sequence analysis is run in multiple cases with the plant fed from 34.5kV transformer TR4, TR5, and split bus configuration (both TR4 and TR5).

The calculation identifies the Degraded Voltage Relay (DGR) voltage setpoint and tolerances associated with DGR time sequence energization and time delay setting and reset voltage with tolerances.

The ETAP run analysis includes verification that each individual LOCA Sequence run has steady state voltages that meet the acceptance criteria for the connected equipment and a LOCA Sequence output voltage vs. time plot on the 4.16kV T-Buses (T11A, T11D, T21A, T21 D), where the degraded voltage relays are located, that does not trip the degraded voltage relays.

During Mode 1, ESF buses are fed from the main generator via UATs and the RATs remain unloaded in standby. The RATs load tap changer controller maintains the tap position at 132.4V (or 4634V at 4.16kV buses) when in standby. This allows for compensation of voltage drop in TR4, TR5 and the RATs when the ESF buses are transferred to offsite power during a LOCA. After the 4.16kV buses are transferred to the RATs, the controller setting is changed to 119.2V (or 4172V at 4.16kV buses) to maintain bus voltage at 100%. The analysis shows that required grid voltages at 345kV and 765kV are less than 95% during normal split bus alignment (Train A on TR4 and Train B on TR5).

to AEP-NRC-2014-42 Page 15 EEEB RAI-4 to AEP-NRC-2013-79, "UFSAR Section 6.3.2 Pages Marked To Show Proposed Changes, Containment Spray Systems, System Design, paragraphA states:

Delaying actuation of the Containment Spray Pumps for a period of time following a containment pressure Hi Hi signal, for example, has the effect of allowing a higher containment backpressure during the initial phases of a loss of coolant accident, which provides a benefit to the Peak Clad Temperature analysis.

Please explain the impact of additionalloading on the RATs and EDGs and potential changes in the fuel oil consumption as a result of the changes in backpressure that ECCS loads could be subjected to during loss of coolant accidents.

Response

As discussed in the response to RAI EEEB RAI-1, the containment design pressure is not affected by the proposed changes. Motor loads used in CNP electrical system analyses and evaluations bound the maximum loading conditions consistent with maximum design performance of the driven equipment. The loads used in the analysis and evaluations are not dependent on the transient response of driven equipment. There are no changes to analyzed motor loads as a consequence of the proposed changes, and therefore no impact on the analyzed performance of the RATs or EDGs.

Similarly, the transient containment pressure response is not a factor in determining design basis EDG fuel oil consumption, so there is no impact on fuel oil consumption or credited fuel oil storage.

EEEB RAI-5 Please provide a discussion of the impact (e.g., changes to environmental qualification parameters such as temperature and pressure) on Environmental Qualification (EQ) of equipment affected by the proposed change. To demonstrate that the equipment remained qualified for service in the revised pressure and temperature environment, please provide a comparison of the EQ overlays with containment temperature and pressure profiles including how margins are being maintained for qualified equipment in accordance with Title 10 of the Code of FederalRegulations (10 CFR) Part50.49.

Response

For CNP, the EQ limiting condition for containment temperature occurs following a main steam line break (MSLB). As explained in Sections 5.5.1.1 and 5.5.2.1 of Enclosure 6 of the original LAR, the MSLB AOR for Unit 1 is a bounding analysis that encompasses both Unit 1 and Unit 2 at a higher-than-licensed power level. Because evaluating the impact of the proposed timing changes to CTS and CEQ fan actuation on this bounding analysis was not straightforward, a to AEP-NRC-2014-42 Page 16 decision was made to prepare a Unit 1-specific full spectrum MSLB analysis to support returning to NOP/NOT conditions. Among other refinements, the new analysis incorporated the Unit 1 licensed power level, modeling of the replacement SGs installed in 2000, the proposed timing change for CTS and CEQ fan actuation, and Westinghouse updates to standard analysis inputs.

The results of the Unit 1 MSLB reanalysis show that the peak post-accident temperature value considered in the EQ program (324.7°F) is not affected by the proposed change to restore NOP/NOT conditions. However, the reanalysis did affect the shape of the post-MSLB temperature profile, as illustrated in Figures RAI-EEEB-5-1 and RAI-EEEB-5-2. The change to the transient temperature profile is being evaluated for impact on the EQ program in accordance with 10CFR 50.49. In conjunction with implementation of NOP/NOT conditions for Unit 1, EQ program documentation will be updated to acknowledge the longer transient. A bounding temperature profile will be used for the update. Based on preliminary assessments and engineering judgment, the change will not adversely impact qualification of components.

The EQ limiting condition for containment pressure occurs following a LOCA. As presented in Section 5.4.2.6 of Enclosure 6 of the original LAR, the long-term peak containment pressure under NOP/NOT conditions remains within the original containment design pressure. The EQ envelope with regard to pressure for equipment in containment is not affected.

The current qualification criteria for other EQ parameters for equipment inside containment, specifically containment post-accident radiation levels, flood level, and pH, are unaffected by implementation of NOP/NOT conditions on Unit 1.

to AEP-NRC-2014-42 Page 17 Original TCD Evaluation

.. . Return to HOP/NOT Evaluation 26 24 0 100 200 300 400 500 Time After Break (s)

Figure RAI-EEEB-1-1: Comparison of Credited Containment Backpressure to AEP-NRC-2014-42 Page 18 II

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ElfRTRE Figure RAI-EEEB-5-1: Post-MSLB Containment Temperature Profile - CNP AOR to AEP-NRC-2014-42 Pagle 19 Lower Compartment Temperature Upper Compartment Temperature 350 300 25 ... . . _. * . . .

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Figure RAI-EEEB-5-2: Post-MSLB Containment Temperature Profile - CNP Unit I Reanalysis to AEP-NRC-2014-42 Page 20 References

1. WCAP-8282, "Final Report Ice Condenser Full Scale Section Test at the Waltz Mill Facility,"

Westinghouse Electric Company, February 1974 (Proprietary)

2. WCAP-8282, Addendum 1, "Answers to AEC Questions on Report WCAP-8282,"

Westinghouse Electric Company, May 1974 (Proprietary)

3. Calculation MD-01-ECCS-004-N, Rev. 6, "Unit 1 ECCS Pumps NPSH Analysis," American Electric Power, November 11, 2013
4. NRC Regulatory Guide 1.1, "Net Positive Suction Head for Emergency Core Cooling and Containment Heat Removal System Pumps," November 2, 1970
5. I&M to NRC letter C0200-09, "Donald C. Cook Nuclear Plants Units 1 and 2, Requested information, Generic Letter (GL) 97-04, Assurance of Sufficient Net Positive Suction Head for Emergency Core Cooling and Containment Heat Removal Pumps," February 4, 2000 (ADAMS ML003681807)
6. NRC to I&M letter, "Donald C. Cook (DC Cook) Units 1 and 2 - Revised Response to NRC Generic Letter 97-04, 'Assurance of Sufficient Net Positive Suction Head for Emergency Core Cooling and Containment Heat Removal Pumps' (TAC Nos. MA8265 and MA8266)",

May 3, 2000, (ADAMS ML003710891)

7. I&M to NRC letter AEP:NRC:8054-02, "Donald C. Cook Nuclear Plant Units 1 & 2, Supplemental Response to Nuclear Regulatory Commission Generic Letter 2004-02:

Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized Water Reactors, February 29, 2008 (ADAMS ML080770407, package) 7.1. AEP:NRC:8054-02, cover letter (ADAMS ML080770394) 7.2. AEP:NRC:8054-02, Attachment 1, "References," through Attachment 3, "Supplemental Response to GL 2004-02 and Request for Additional Information," February 29, 2008 (ADAMS ML080770395) 7.3. AEP:NRC:8054-02, Attachment 3, "I&M Response to Information Item 3.f.4," to "NRC Information Item 3, "Conclusions," February 29, 2008 (ADAMS ML080770396) 7.4. AEP:NRC:8054-02, Attachment 4, Figure A4-1 to Attachment 5, Figure A5-40 (ADAMS ML080770400) 7.5. AEP:NRC:8054-02, Attachment 5, Figure A5-41 to Attachment 7, "Regulatory Commitments," February 29, 2008 (ADAMS ML080770404)

8. NRC to I&M letter, "Donald C. Cook Nuclear Plant, Units 1 and 2 (DCCNP 1&2) - NRC Staff Comments on Licensee's Supplemental Responses to Generic Letter 2004-02 (TAC Nos.

MC4679 and MC4680), July 27, 2010 (ADAMS ML101960128)

ENCLOSURE 4 TO AEP-NRC-2014-42 Westinghouse Letter, CAW-14-3954, Application for Withholding Proprietary Information from Public Disclosure

Westinghouse Electric Company Engineering, Equipment and Major Projects 1000 Westinghouse Drive, Building 3 Cranberry Township, Pennsylvania 16066 USA U.S. Nuclear Regulatory Commission Direct tel: (412) 374-4643 Document Control Desk Direct fax: (724) 940-8560 11555 Rockville Pike e-mail: , greshaja@westinghouse.com Rockville, MD 20852 Proj letter: AEP-14-23 CAW-14-3954 May 28, 2014 APPLICATION FOR WITHHOLDING PROPRIETARY INFORMATION FROM PUBLIC DISCLOSURE

Subject:

Westinghouse Responses to NRC, "Donald C. Cook Nuclear Plant Unit 1 - Request for Additional Information on the Application for Amendment to Restore Normal Reactor Coolant System Pressure and Temperature Consistent with Previously Licensed Conditions (TAC No.

MF2916)," Set #2, SCVB RAI-4f (Proprietary)

The proprietary information for which withholding is being requested in the above-referenced report is further identified in Affidavit CAW-14-3954 signed by the owner of the proprietary information, Westinghouse Electric Company LLC. The Affidavit, which accompanies this letter, sets forth the basis on which the information may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in paragraph (b)(4) of 10 CFR Section 2.390 of the Commission's regulations.

Accordingly, this letter authorizes the utilization of the accompanying Affidavit by Indiana Michigan Power Company.

Correspondence with respect to the proprietary aspects of the application for withholding or the Westinghouse Affidavit should reference CAW-14-3 954, and should be addressed to James A. Gresham, Manager, Regulatory Compliance, Westinghouse Electric Company, 1000 Westinghouse Drive, Building 3 Suite 310, Cranberry Township, Pennsylvania 16066.

Very truly yours,

'James A. Gresham, Manager Regulatory Compliance Enclosures

CAW-14-3954 AFFIDAVIT COMMONWEALTH OF PENNSYLVANIA:

ss COUNTY OF BUTLER:

Before me, the undersigned authority, personally appeared James A. Gresham, 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:

P"imes A. Gresham, Manager Regulatory Compliance Sworn to and subscribed before me this ý 2 J Ngoary of 2 2014 Noay[ bi COMMONWEA ,OF NNSM IA F[ NotarhlSealI

/ Renee Glampole, Notary Public I l Penn Twp., WestmWeland County i My Commiss Expires Sept. 2O25, 2017

- -OFTWAR[E,*T'O P- -..-NYLVJi MEMJBER,

2 CAW- 14-3954 (1) 1 am Manager, Regulatory Compliance, 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 Westinghouse.

(2) I am making this Affidavit in conformance with the provisions of 10 CFR Section 2.390 of the Commission's regulations and in conjunction with the Westinghouse Application for Withholding Proprietary Information from Public Disclosure accompanying this Affidavit.

(3) I have personal knowledge of the criteria and procedures utilized by Westinghouse 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 2.390 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 constitute Westinghouse policy and provide 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:

(a) The information reveals the distinguishing aspects of a process (or component, structure, tool, method, etc.) where prevention of its use by any of

3 CAW-14-3954 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.

(iii) 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.

(b) It is information that 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.

4 CAW-14-3954 (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.

(iv) The information is being transmitted to the Commission in confidence and, under the provisions of 10 CFR Section 2.390, it is to be received in confidence by the Commission.

(v) 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.

(vi) The proprietary information sought to be withheld in this submittal is that which is appropriately marked in Westinghouse Responses to NRC, "Donald C. Cook Nuclear Plant Unit 1 - Request for Additional Information on the Application for Amendment to Restore Normal Reactor Coolant System Pressure and Temperature Consistent with Previously Licensed Conditions (TAC No. MF2916)" Set #2, SCVB RAI-4f (Proprietary), for submittal to the Commission, being transmitted by Indiana Michigan Power Company letter and Application for Withholding Proprietary Information from Public Disclosure, to the Document Control Desk. The proprietary information as submitted by Westinghouse is that associated with NRC approval of WCAP- 17762-NP, and may be used only for that purpose.

5 CAW-14-3954 (a) This information is part of that which will enable Westinghouse to:

(i) Provide input to Indiana Michigan Power Company for input to the U.S.

Nuclear Regulatory Commission in response for Additional Information regarding Restoration of Normal Operating Pressure and Normal Operating Temperature.

(ii) Provide licensing support for customer submittal.

(b) Further this information has substantial commercial value as follows:

(i) Westinghouse plans to sell the use of the information to its customers for the purpose of obtaining license changes for a Westinghouse pressurized water reactor (PWR).

(ii) Westinghouse can sell support and defense of the technology to tis customer in the licensing process.

(iii) The information requested to be withheld reveals the distinguishing aspects of a methodology which was developed by Westinghouse.

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 technical evaluation justifications 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.

6 CAW-14-3954 In order for competitors of Westinghouse to duplicate this information, similar technical programs would have to be performed and a significant manpower effort, having the requisite talent and experience, would have to be expended.

Further the deponent sayeth not.

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 2.390 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) 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 2.390(b)(1).

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 2.390 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.

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