W3F1-2015-0062, Control Element Assembly Drop Times Submittal Request for Additional Information

Control Element Assembly Drop Times Submittal Request for Additional Information

ML15268A019
Person / Time
Site:	Waterford
Issue date:	09/23/2015
From:	Chisum M Entergy Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
Shared Package
ML15268A013	List: ML15268A020 W3F1-2015-0062, Control Element Assembly Drop Times Submittal Request for Additional Information
References
CAW-15-4272, W3F1-2015-0062
Download: ML15268A019 (32)

Text

Proprietary Information -Withhold From Public Disclosure Under 10 CFR 2.390 The balance of this letter may be considered non-proprietary upon removal of Attachment 4.

--En tergy Entergy Operations, Inc.

17265 River Road Killona, LA 70057 Tel 504 739 6660 Fax 504 739 6678 Michael R. Chisum Site Vice President Waterford 3 W3FI1-2015-0062 September 23, 2015 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555

SUBJECT:

Control Element Assembly Drop Times Submittal Request for Additional Information Waterford Steam Electric Station, Unit 3 Docket No. 50-382 License No. NPF-38

REFERENCES:

1. W3F1-2015-0040, License Amendment Request to Revise Control Element Drop Times, July 2, 2015 [ADAMS Accession Number ML15197A106].

2. W3F1-2015-0061, Supplement to Revise Control Element Assembly Drop Times Associated with Technical Specification 3.1.3.4, August 13, 2015 2015 [ADAMS Accession Number MLI15226A346].

3. NRC CEA Drop Time Submittal Request for Additional Information, August 26, 2015, [ADAMS Accession Number ML15232A275].

Dear Sir or Madam:

On July 2, 2015, Entergy Operations, Inc. (Entergy) requested an amendment to revise the Control Element Assembly (CEA) drop times associated with Technical Specification 3.1.3.4 for Waterford Steam Electric Station Unit 3 (Waterford 3) [Reference 1]. On August 13, 2015, Waterford 3 submitted a supplement [Reference 2] to provide additional accident scenario results to bound a possible delay in the control element assembly holding coil decay time. Subsequently, the Nuclear Regulatory Commission (NRC) has requested additional information to aid in their review [Reference 3]. This letter provides the response to the NRC request for additional information.

This submittal does not alter the no significant hazards consideration or environmental assessment previously submitted by Waterford 3 in letter W3F1 -201 5-0040 [Reference 1].

W3FI1-2015-0062 Page 2 Waterford 3 letter W3F1-2015-0040 [Reference 1] submitted a proprietary fuel thermal conductivity degradation evaluation. This letter provides a non-proprietary and proprietary version of that fuel thermal conductivity degradation evaluation. is proprietary in its entirety, as it contains information that is proprietary to Westinghouse Electric Company (Westinghouse). Attachment 3 contains a redacted non-proprietary version. Attachment 2 contains the Proprietary Information Affidavit. The purpose of Attachment 2 is to withhold the proprietary information contained in Attachment 4 from public disclosure. The Affidavit, signed by Westinghouse as the owner of the information, sets forth the basis for which the information may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in paragraph (b)(4) of § 2.390 of the Commission's regulations. Accordingly, it is respectfully requested that the information proprietary to Westinghouse be withheld from public disclosure in accordance with 10CFR 2.390.

If you have any questions or require additional information, please contact John Jarrell, Regulatory Assurance Manager, at 504-739-6685.

I declare under penalty of perjury that the foregoing is true and correct. Executed on September 23, 2015.

Sincerely, MRC/J PJ/wjs Attachments:

1. Response to NRC Request for Additional Information

2. Fuel Thermal Conductivity Degradation Evaluation Proprietary Affidavit

3. Non-Proprietary - Fuel Thermal Conductivity Degradation Evaluation

4. PROPRIETARY - Fuel Thermal Conductivity Degradation Evaluation cc: Mr. Marc L. Dapas Regional Administrator U. S. NRC, Region IV RidsRgn4MailCenter@nrc.gov NRC Senior Resident Inspector for Waterford 3 Frances.Ramirez@nrc.gov (SRI)

Chris.Speer~nrc.gov (RI)

NRC Project Manager for Waterford 3 Michael.Orenak@nrc.gov Louisiana Department of Environmental Quality Office of Environmental Compliance Surveillance Division Ji.Wiley@LA.gov

Attachment I to W3FI1-2015-0062 Response to NRC Request for Additional Information to W3F1 -201 5-0062 Page 1 of 21

1.0 DESCRIPTION

On April 22, 2015, a Category 1 public meeting was held between the U.S. Nuclear Regulatory Commission (NRC) staff and representatives of Entergy Operations, Inc.

(Entergy) and Westinghouse Electric Company (Westinghouse) at the NRC Headquarters.

The purpose of the meeting was to discuss Entergy's proposed license amendment request (LAR) regarding changes to Technical Specification (TS) 3.1.3.4 (CEA Drop Time) and Updated Final Safety Analysis Report (UFSAR) Chapter 15 (Accident Analyses).

Reference 3.1 provides the meeting summary information and Reference 3.2 provides the meeting presentation information.

As discussed in the public meeting and pursuant to 10 CFR 50.90, Entergy requested an amendment to revise the Control Element Assembly (CEA) drop times associated with Technical Specification 3.1.3.4 for Waterford Steam Electric Station Unit 3 (Waterford 3)

[Reference 3.3]. The submittal identified that additional information would be useful to the NRC during their review. The NRC staff reviewed the Waterford 3 submittal and concluded that this same additional information was necessary to enable the NRC staff to make an independent assessment regarding the acceptability of the proposed amendment

[Reference 3.4].

On August 13, 2015, Waterford 3 submitted a supplement [Reference 3.51 to provide additional accident results due to a possible cause identified in the apparent cause evaluation. Subsequently, the Nuclear Regulatory Commission (NRC) has requested additional information to aid in their review [Reference 3.6]. Section 2 provides the response to the NRC request for additional information.

2.0 REQUEST FOR ADDITIONAL INFORMATION (RAI)

NRC RAI #1 In Attachment 2, page 4, the following statement is made:

The analysis margin for the axial power distribution was reduced from an ASI [axial shape index] of +0.3 to +0.2, which is conservative to the COLR [core operating limits report] limit of +0.16.

Table 15.0-4 of the FSAR shows the axial shape index used is: -0.2 :5 ASI s +0.2. Provide the bases for the current margin for the ASI of +0.3 and explain why this is different from the ESAR value.

Waterford 3 RAI #1 Response W3FI-2015-0040 Attachment 2 page 4 is for UFSAR Section 15.1.2.3 (Increased Main Steam Flow). UFSAR Table 15.0-4 is for the general overview of Chapter 15 initial conditions.

UFSAR Section 15.0.3.2 states:

The range of values of each of the principal process variables that were considered in analyses of all incidents discussed in this section are listed in Table 15.0-4. It is to W3F1 -2015-0062 Page 2 of 21 emphasized that no plant operational or safety problems have been identified for operating conditions outside of the range shown in Table 15.0-4. This range merely represents a range of expected normal reactor operation.

UFSAR Table 15.0-4 is based upon the Technical Specifications, Core Operating Limits Report (COLR), and reload design allowed values. The specific analyses may use different values provided that they are conservative and provide bounded results.

Letter W3F1-2003-0074 [Reference 3.9] for the extended power uprate was approved in NRC Technical Specification Amendment 199 [Reference 3.10]. Letter W3F1-2003-0074

[Reference 3.9] Attachment 5 Section 2.13.0.2 described the range of initial conditions (similar to UFSAR Table 15.0-4) as follows:

The range of initial conditions evaluated in the non-LOCA transient analyses is listed in Table 2.13.0-2. Values beyond this range are used in certain analyses to provide additional margin.

This is the same approach used for UFSAR Table 15.0-4. UFSAR Section 15.0.3.2 will be updated to contain similar wording to that contained W3FI-2003-0074 Attachment 5 Section 2.13.0.2 to better describe that more bounding inputs may be used to provide conservative results.

NRC RAI #2 In Attachment 2, pages 4, 6, 16, and 17, the following statement is made:

Analysis margin for the least negative Doppler reactivity was reduced from -0.00113 to -0.0013 Ap/h/°K.

Provide the bases for the least negative Doppler coefficients of -0.00113 to -0.0013 Ap/U/K. Also, justify adequacy of the use of the value of -0.0013 Ap/PPK in the reanalysis used to support the proposed increase in the limits of the CEA drop times in TS 3.1.3.4.

Waterford 3 RAI #2 Response Figure 2-1 documents the fuel temperature coefficient (FTC) ranges from the most negative (-0.0026 Ap/h/°K) to the least negative (-0.0013 Ap/*/°K) for the current Waterford Unit 3 operating fuel cycles. These bounding cycle independent values are confirmed every reload cycle.

A conservative value of -0.00113 Ap/P/°K was selected during the Waterford Unit 3 extended power update analyses with the intention of bounding a wider range of subsequent reload cycles. As part of the reanalysis to support the revised CEA drop time curve for this License Amendment Request, this extra analysis conservatism was removed.

Figure 2-1 shows the FTC in units of IAp/°F for the -0.00113 Ap/h/°K, -0.00130 Ap//°K, and -0.0026 App/a/K curves.

to W3F1 -2015-0062 Page 3 of 21 Figure 2-1 Cycle Independent Fuel Temperature Coefficient Range and Safety Analysis Limits Fuel Temperature, °F

-4.00E-05

-3.50E-05

-3.00E-05 - .F,, ,w.

-2.50E-05

• -2.00E-05

-1.00E-05

-5.00E-06 5' iC00 15b30 20 30 25K30 30 0.00E+00 ~___

-- -- -0.00260 rho/(SQRT-K) ....- 0.00130 rho/(SQRT-K) - -0.00113 rho/(SQRT-K)

NRC RAI #3 In Attachment 2, page 7, the following statement is made:

Plant changes since the extended power uprate have been incorporated into the analysis under the 10 CFR [Title 10 of the Code of FederalRegulations] 50.59 process. The analysis has been updated to account for the replacement steam generators (SGs) and the NGF [next generation fuel] DNBR [departure from nucleate boiling ratio] correlation. The revised evaluation started with the analysis of record and only revised CEA drop time to determine the impact.

Clarify the difference between "analysis" and "analysis of record." Does the revised evaluation include the replacement steam generators and NGF DNBR correlation?

Waterford 3 RAI #3 Response "Analysis" and "Analysis of Record" are used synonymously in this instance. The revised analysis does include the replacement steam generators and Next Generation Fuel DNBR correlation.

to W3F1-201 5-0062 Page 4 of 21 NRC RAI #-4 In Attachment 4, page 4, the Figure 2, "Cycle 15 through Cycle 20 CEA Insertion Times,"

shows the repeated measured CEA insertion times for Cycle 20. The measured insertion times for Cycle 20-2 are significantly shorter than that of Cycle 20-1.

Discuss the measurement methods for the Cycle 20-1 and Cycle 20-2 data, including uncertainties, identifying the causes for the reduction of the insertion times from measurements for Cycle 20-1 to Cycle 20-2, and justify the adequacy of the Cycle 20-2 data by showing that the reduced insertion times are not contributed from random measurement errors. Was anything done to improve the drop time between Cycle 20-1 and Cycle 20-2, or was the test simply repeated?

Waterford 3 RAI #4 Response No changes were made between performance of the Cycle 20-1 and Cycle 20-2 CEA drop time tests.

The CEA Drop Time Test (CDTT) software records all the CEA positions every 50 milliseconds [CENPSD-388 Revision 01 - CEA Drop Time Test Software User's Manual for WSES-3]. The CDTT software will always round time up to the next 50 millisecond interval ensuring the rod drop times delivered will be conservative. This results in the CDTT software having a CEA drop time uncertainty of 100 milliseconds (0.1 seconds).

The Cycle 20-1 initial test was performed with the average time being 3.024 seconds. The Cycle 20-2 second test was performed with the average time being 2.967 seconds. The difference between the Cycle 20-1 and Cycle 20-2 test results were within the CDTT software uncertainty band.

NRC RAI #5 Table 4.0-1 in Attachment 1, page 4, provides the new CEA drop times. Other than the change at 90% CEA insertion (from 3.0 to 3.2 seconds), what is the basis for the other changes to the CEA drop time curve? Was any plant data used to determine the new curve? If so, provide plant data and demonstrate that the new curve is conservative. Provide the reactivity vs. time (or CEA position) curve that was presented during the pre-application public meeting on April 22, 2015.

Waterford 3 RAI #5 Response The basis for the revised rod drop position versus time was based on plant data and a conservative analysis approach. Reviewing the time it takes the rods to reach 90%

inserted for each cycle reveals that for Cycles 5 through 17, the average drop time is 2800 milliseconds while the average drop time for Cycles 19 and 20 is 2950 milliseconds. To bound the step change, the analysis rod drop time was shifted by 200 milliseconds or 0.2 seconds for all rod positions from 10% to 90% inserted. To maintain continuity in the curve, the 5% insertion position time was increased by 0.15 second.

to W3F 1-2015-0062 Page 5 of 21 The shape of the revised drop time curve is the same as that utilized in the current UFSAR Chapter 15 Non-LOCA safety analysis, but shifted to the right (slightly slower) to account for the impact of major plant modifications on the reactor coolant system over time. Therefore, the shape of the rod drop time curve remains constant in the revised analyses supporting the Technical Specification change request. Confirmation of the revised rod drop position versus time curve will occur as part of the Cycle 21 rod drop testing.

Figure 5-1 provides the rod drop position (% withdrawn) versus time for the current safety analysis and the revised rod drop time.

Figure 5-1 Rod Drop Position versus Time 120 100 80~

"C 60~

40~

20 0

0 1 2 3 4 Ttu., seconds

-- -- Curnte Rod DropTiuw -.. R--ied Rod Drop Time Per the response to Question 1, the actual full power axial shape index (ASI) range is +/-0.2.

The normalized reactivity versus time value of +0.3 ASI was conservatively used during the Waterford Unit 3 extended power uprate analyses with the intent of bounding a wider range of subsequent reload cycles. As part of the reanalysis due to the revised rod drop curve, this extra analysis conservatism was removed. Figure 5-2 provides normalized reactivity versus time for both a +0.3 and +0.2 ASI.

to W3F 1-2015-0062 Page 6 of 21 Figure 5-2 Normalized Reactivity vs. Time 0.2

-O.2 4 .0.4

-1

-1.2 0123 4 Timne, seconds

--- .O.3 M Ciurnt Drp Time -.-. *0*.3 ASS ul Domp Time .0,.2 ASS Rte~isel Drop Time NRC RAI #6 , page 12, calls out references 7.40, 7.41 and 7.42, however, these three references are not provided in the References section (pages 13-15). It appears that some of the references are improperly numbered. For example, page 12 of Attachment 1 states "The NRC approved this request in NRC Technical Specification Amendment 158

[Reference 7.42]." However, this appears to be Reference 7.37 on page 15. Update the reference numbers in text so they are consistent with the reference list.

Waterford 3 RAI #6 Response The reference issue is limited to letter W3F1-2015-0040 Section 6. All references are included in the reference section. The following is the correct reference numbers.

Waterford 3 letter W3P89-3094 was listed as reference 7.37 and should have been reference 7.32.

Waterford 3 Technical Specification Amendment 58 was listed as reference 7.38 and should have been reference 7.33.

ANO letter 2CAN080701 was listed as reference 7.39 and should have been reference 7.34.

ANO Technical Specification Amendment 275 was listed as reference 7.40 and should have been reference 7.35.

to W3F1 -2015-0062 Page 7 of 21 St. Lucie letter L-2009-127 was listed as reference 7.41 and should have been reference 7.36.

St. Lucie Technical Specification Amendment 158 was listed as reference 7.42 and should have been reference 7.37.

The corrected W3FI-2015-0040 Section 6 is provided.

6.0 PRECEDENCE Waterford 3 previously requested a Technical Specification change to the CEA drop methodology in letter W3P89-3094 [Reference 7.32]. This submittal was approved in NRC Technical Specification Amendment 58 [Reference 7.33]. This change request is similar in scope to that previously requested and approved.

Arkansas Unit 2 requested a Technical Specification change to increase the individual CEA drop time in letter 2CAN080701 [Reference 7.34]. The NRC approved this request in NRC Technical Specification Amendment 275 [Reference 7.35]. This change was necessitated by the transition to Next Generation Fuel similar to Waterford 3's submittal request.

St. Lucie Unit 2 requested a Technical Specification change to increase the CEA drop time in letter L-2009-127 [Reference 7.36]. The NRC approved this request in NRC Technical Specification Amendment 158 [Reference 7.37].

NRC RAI #7 , pages 2 and 3, discuss apparent and potential causes for increased rod drop times. Provide the cycles as to when plant modifications were made (i.e.,

replacement steam generators were installed between Cycles XX and YY).

Waterford 3 RAI #7 Response Appendix K Power Uprate was implemented in Cycle 12.

Extended Power Uprate was implemented in Cycle 14.

Alternate Source Term was implemented in Cycle 14.

Next Generation Fuel (NGF) - Cycle 16 implemented a partial core of NGF and Cycle 17 implemented a full core of NGF.

Control Element Assembly Replacement - The outage prior to Cycle 18 (between Cycle 17 and Cycle 18) replaced the CEAs.

Steam Generator Replacement - The outage prior to Cycle 19 (between Cycle 18 and Cycle 19) replaced the steam generators.

Reactor Head Replacement - The outage prior to Cycle 19 (between Cycle 18 and Cycle

19) replaced the reactor vessel head.

Control Element Drive Mechanism Replacement - The outage prior to Cycle 19 (between Cycle 18 and Cycle 19) replaced the CEDMs.

to W3F1 -2015-0062 Page 8 of 21 NRC RAI #8 , Page 6, Paragraph 3 states that:

In order to aid the NRC review, the relevant Waterford 3 licensing basis history is provided.

The last major plant change that submitted the accident and transient analyses to the NRC for review was the extended power uprate [Reference 7.15] and alternate source term implementation [Reference 7.18]. Two additional major plant changes that have been implemented since the extended power uprate are the use of Next Generation Fuel (NGF) and Steam Generator replacement. These changes were addressed by NRC approvals (Table 4.0-2) and the 10CFR50.59 process. The Westinghouse reload analysis methodology has been applied to implement these Waterford 3 changes.

Please confirm that the stated "Westinghouse reload analysis methodology" was previously approved by the NRC and that no changes were made to the approved methodology since it was used to implement the use of NGF and Steam Generator replacement. If changes were made to the NRC-approved reload analysis methodology, identify and justify the changes for the use in support of the proposed TS regarding the CEA drop times.

Waterford 3 RAI #8 Response The CEA drop time analyses did not require any changes to the methodologies described in the Update Final Safety Analysis Report [Reference 3.8].

Westinghouse reload analysis methodology is not a specific topical that was approved by the NRC. The Westinghouse reload analysis methodology is a term used to describe the accumulation of those methodologies that have been approved by the NRC and used in the reload process.

The UFSAR methodologies were submitted in the extended power uprate request

[W~aterford 3 letter W3F1-2003-0074 Reference 3.9]. Letter W3FI-2003-0074 Attachment 5 Section 2.6 (Reactor Systems), Section 2.12.3.1 (LBLOCA Methodology), Section 2.12.4.1 (SBLOCA Methodology), Section 2.12.5.1 (LOCA Long Term Cooling Methodology), and Section 2.13.0.1 (Non-LOCA Methodology) provided the analysis methodologies utilized by Westinghouse. Methodology changes since extended power uprate that could be applicable to this change were listed in letter W3FI-2015-0040

[Reference 3.3] Table 4.0-2 (NRC Amendments of Interest).

to W3F 1-2015-0062 Page 9 of 21 N RC RAI #9 , Page 5, "Section FSAR 15.1.2.3 Increased Main Steam Flow," states that:

The increase in peak secondary pressure is based on the loss of condenser vacuum results which showed an increase in peak secondary pressure of less than 1 psi.

ESAR Section 15.1.2.3 does not discuss a loss of condenser vacuum. Explain why a loss of condenser vacuum was used for this case.

Waterford 3 RAI #9 Response The limiting events with respect to peak primary and secondary pressures are the loss of condenser vacuum and feedwater line break accidents. These are the limiting events because they have the closest approach to the Technical Specification Section 2.1.2 (Reactor Coolant System Pressure) safety limit of 2750 psia and secondary pressure limit of 1210 psia (110% of design pressure). UFSAR Section 15.1 contains the increase in heat removal by the secondary system events. These events are less adverse than the loss of condenser vacuum with respect to peak primary and secondary pressure.

Specifically, loss of condenser vacuum peak secondary pressure is 1181 psia [UFSAR Section 15.2.1.3.3.3] whereas the increased main steam flow event peak secondary pressure is 1102 psia [UFSAR Table 15.1-8A].

The increased CEA drop time impacts the power reduction post-trip and the amount of energy deposited into the primary coolant system. The slightly longer drop time means a slight increase in the amount of energy added to the system. The assessment provided in W3FI-2015-0040 used the loss of condenser vacuum event because it produced the largest post-trip primary and secondary pressure spike due to losing its secondary heat removal capability. Since, loss of condenser vacuum event produces a more adverse pressure transient, the slight energy deposition increase would be expected to have the most adverse impact on this event. Thus, taking the loss of condenser vacuum event results and applying them to the increased main steam flow event was considered a conservative approach to ensure the change was bounded.

to W3FI1-2015-0062 Page 10 of 21 NRC RAI #10 , page 6, states:

The increase in peak primary pressure was based on the loss of condenser vacuum results which showed an increase in peak primary pressure of less than 1 psi

[pounds per square inchJ when the pressurizer pressure exceeded the pressurizer safety valve opening setpoints (2575 psia [pounds per square inch absolute]). The increase in peak secondary pressure was based on the loss of condenser vacuum results which showed an increase in peak secondary pressure of less than 1 psi.

FSAR, Section 15.1.2.4 does not discuss loss of condenser vacuum. Explain why loss of condenser vacuum was used for this case.

Waterford 3 RAI #10 Response The limiting events with respect to peak primary and secondary pressures are the loss of condenser vacuum and feedwater line break accidents. These are the limiting events because they have the closest approach to the Technical Specification Section 2.1.2 (Reactor Coolant System Pressure) safety limit of 2750 psia and secondary pressure limit of 1210 psia (110% of design pressure). UFSAR Section 15.1 contains the increase in heat removal by the secondary system events. These events are less adverse than the loss of condenser vacuum with respect to peak primary and secondary pressure.

Specifically, loss of condenser vacuum peak primary pressure is 2711 psia and secondary pressure is 1181 psia [UFSAR Section 15.2.1.3.3.3] whereas the inadvertent operating of an atmospheric dump valve event peak primary pressure is 2583 psia and secondary pressure is 1117 psia [UFSAR Table 15.1-8C].

The increased CEA drop time impacts the power reduction post-trip and the amount of energy deposited into the primary coolant system. The slightly longer drop time means a slight increase in the amount of energy added to the system. The assessment provided in W3F1-2015-0040 used the loss of condenser vacuum event because it produced the largest post-trip primary and secondary pressure spike due to losing its secondary heat removal capability. Since, the loss of condenser vacuum event produces a more adverse pressure transient, the slight energy deposition increase would be expected to have the most adverse impact on this event. Thus, taking the loss of condenser vacuum event results and applying them to the inadvertent operating of an atmospheric dump valve event was considered a conservative approach to ensure the change was bounded.

to W3F1 -2015-0062 Page 11 of 21 NRC RAI #11 , pages 4 and 5, FSAR, Section 15.1.2.3 considers two cases: (1) typical case and (2) worst departure from nucleate boiling performance case. This section does not discuss which case was analyzed. Clarify if both cases were considered in the updated analysis.

Waterford 3 RAI #11 Response UFSAR Section 15.1.2.3 describes the increased main steam flow with a concurrent loss of offsite power event. The event analyzed was the worst DNB performance case. Only the worst DNB performance case was evaluated because it bounds the typical case in terms of fuel failure.

NRC RAI #12 The table in Attachment 2, page 6, contains a "*" after 2584 psia. Provide the significance of the "*".

Waterford 3 RAI #12 Response The "*" should have not have been in the table. It has no significance.

NRC RAI #13 , page 26, states, in part, that changes to the updated analysis include:

..the use of the actual initial thermal margin reserved in the LCO.

Provide the parameter that was changed and the original and updated values.

Waterford 3 RAI #13 Response U FSAR Section 15.9.1.1 describes the Asymmetric Steam Generator Transient (ASGT).

The revised CEA drop time analysis changed the AOR requirement for the actual initial thermal margin reserved. The limiting AOR scenario evaluated was the 95% power closure of the main steam isolation valve #2. The AOR initial thermal margin is defined as Required Over Power Margin (ROPM) and was 119.9% whereas the revised CEA drop time analysis ROPM was 120.23%. The ROPM that was already preserved in COLSS at 95% power is 123%, thus the ASGT event used additional initial thermal margin but was within the initial thermal margin already preserved so no COLSS changes were required.

The ASGT analysis demonstrated that the initial margins were adequate to ensure that the ASGT event does not violate the DNBR (> 1.24) and Fuel Centerline Temperature SAFDLs.

to W3F1 -2015-0062 Page 12 of 21 NRC RAI #14 In FSAR 15.9.1.1, "Asymmetric Steam Generator Transient," there are four events considered including: Loss of Load to One Steam Generator (LLI1SG), Excess Load to One Steam Generator (EL/i1SG), Loss of Feedwater to One Steam Generator (LF/1 SO) and Excess Feedwater to One Steam Generator (EF/1SG). In the updated analysis described in Attachment 2, pages 25-26, verify if all four cases were analyzed or if it was assumed the limiting case was still limiting after the rod drop time change.

Waterford 3 RAI #14 Response The limiting initiating event for the Asymmetric Steam Generator Transient analysis is the closure of the Main Steam Isolation Valve (MSIV) to steam generator #2 (LL/1SG). Only the limiting case was evaluated with the revised CEA drop time.

NRC RAI #15 In Attachment 2, the results of the analysis for a loss of normal feedwater (LONE) event were used to support the adequacy of the dose releases for the following events:

Page 10 - Steam System Piping Failures: Pre-Trip Power Excursion Analysis Page 18 - Single Reactor Coolant Pump (RCP) Shaft Seizure / Sheared Shaft Page 22 - CEA Ejection Page 23 - Primary Sample or Instrument Line Break Page 24 - Steam Generator Tube Rupture Page 24 - Loss of Coolant Accident Provide justification for each event above for the use of the LONE results to support the radiological releases.

Waterford 3 RAI #15 Response Water-ford 3 letter W3F1-2015-0040 [Reference 3.3] generically described that the W3F1-2015-0040 Attachment 2 Loss of Normal Feedwater Flow event (ESAR 15.2.3.2) was chosen to evaluate the transient characteristics with respect to energy deposition and associated steam releases which would be applicable to all the events. The analysis showed that the differences in primary and secondary system energy after reactor trip are insignificant. As time increases farther past the time of CEA rod insertion, the differences of the impact of the revised CEA drop time become negligible. The radiological releases due to steam release and break flow would remain the same. In addition, the W3FI-2015-0040 Attachment 2 results for each of the individual events demonstrate that the fuel failure limits remain unchanged which mean the radiological source terms remain the same. The W3F1-2015-0040 conclusion was there is no change to the radiological results.

The increased CEA drop time impacts the power reduction post-trip and the amount of energy deposited into the primary coolant system. The slightly longer drop time means a slight increase in the amount of energy added to the system. For the loss of normal feedwater event, Figure 15-1 shows the energy added to the reactor coolant system from time of trip (43.7 seconds) to 50 seconds for both the AOR and revised CEA drop time to W3F 1-2015-0062 Page 13 of 21 analysis. Figure 15-2 shows the loss of normal feedwater event AOR and revised CEA drop time analysis energy deposition from the time of trip to 1800 seconds. The deposited energy between the reactor trip time and 1800 seconds is 161485.7 Mwt-sec for the AOR and 162228.1 Mwt-sec for the revised CEA drop time analysis. This equates to an energy change of 0.0046 or 0.46% [(162228.1 - 161485.7)1/161485.7] which would have a negligible change on the radiological consequences.

Figure 15-1. Deposited Energy 4000 3500 r 3000.,

2500 I2000 1500 1000 500 --

0-43 45 46 47 48 49 50 44 Time (seconds)

-I,-Revised CEA Drop Time Analysis

-4Analysis of Record to W3F1 -2015-0062 Page 14 of 21 Figure 15-2. Deposited Energy 3500 i 3000 2500 i 1500 1000-5001 0

0 L 200 400 600 800 1000 1200 10 1400 60 1600 10 1800 Time (seconds)

-4.-Analysis of Record -U1-Revised CEA Drop Time Analysis Waterford 3 letters W3F1-2004-0053 [Reference 3.13] and W3F1-2004-0071 [Reference 3.16] submitted the Alternate Source Term (AST) analyses to the NRC. NRC Operating License amendment 198 [Reference 3.14] approved the use of AST for Waterford 3.

W3FI1-2004-0053, W3F1 -2004-0071, and Amendment 198 provide additional details for each of the radiological events. The Waterford 3 AST analyses follow the guidance provided in NRC Regulatory Guide 1.183 [Reference 3.15]. The AST analyses are evaluated for the Exclusion Area Boundary (EAB) dose for a 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> duration [Regulatory Guide 1.183 Section 4.1.5], Low Population Zone (LPZ) dose for the accident duration

[Regulatory Guide 1.183 Section 4.1.6], and control room dose for the accident duration

[Regulatory Guide 1.183 Section 4.2.6]. Since, the EAB, LPZ, and control room doses are calculated for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and beyond, the use of the loss of normal feedwater event information is appropriate for the time duration. In addition, Figure 15-2 shows a negligible change (0.46%) in deposited energy at 1800 seconds. As time increases to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and beyond, the deposited energy difference will continue to decrease. Thus, the radiological consequences are not adversely impacted by this change.

The NRC had requested a more detailed description of the release pathways. The following provides the references to more detailed release information for each of the events requested.

to W3FI-201 5-0062 Page 15 of 21 Steam System Pipingi Failures: Pre-Trip Power Excursion Analysis W3FI-2004-0053 [Reference 3.13] Attachment 2 Section 7.0 describes the pre-trip steam line break event. The NRC Alternate Source Term Safety Evaluation Report

[Reference 3.14] Section 2.1.2 also describes the pre-trip steam line break event.

Single Reactor Coolant Pump (RCP) Shaft Seizure I Sheared Shaft W3F1-2004-0071 [Reference 3.16] Attachment 1 Section 4.0 describes the reactor coolant pump sheared shaft I seized rotor event. The NRC Alternate Source Term Safety Evaluation Report [Reference 3.14] Section 2.1.4 also describes the reactor coolant pump sheared shaft I seized rotor event.

CEA Ejection W3FI-2004-0053 [Reference 3.13] Attachment 2 Section 10.0 describes the CEA Ejection event. The NRC Alternate Source Term Safety Evaluation Report

[Reference 3.14] Section 2.1.5 also describes the CEA ejection event.

Primary Sample or Instrument Line Break W3FI-2004-0071 [Reference 3.16] Attachment 1 Section 7.0 describes the letdown line break event. The NRC Alternate Source Term Safety Evaluation Report

[Reference 3.14] Section 2.1.6 also describes the letdown line break event.

Steam Generator Tube Rupture W3F1-2004-0053 [Reference 3.13] Attachment 2 Section 8.0 describes the CEA Ejection event. The NRC Alternate Source Term Safety Evaluation Report

[Reference 3.14] Section 2.1.7 also describes the CEA ejection event.

Loss of Coolant Accident W3FI-2004-0053 [Reference 3.13] Attachment 2 Section 5.0 and 6.0 describes the CEA Ejection event. The NRC Alternate Source Term Safety Evaluation Report

[Reference 3.14] Section 2.1.8 and 2.1.9 also describes the CEA ejection event.

to W3F1 -2015-0062 Page 16 of 21 NRC RAI #16 , page 17, discussed the analysis of the single reactor coolant pump (RCP) shaft seizure/sheared shaft events. The reanalysis showed that the peak primary pressure was increased by 20 pounds per square inch (psi) to 2442 pounds per square inch absolute (psia) and the peak secondary pressure was increased by 1 psi to 1118 psia. The increase in the peak primary pressure was based on the LONE long term results and the increase in the peak secondary pressure was based on the loss of condenser vacuum (LOCV) results.

Please provide justification for the use of two different events results, LONF and LOCV, to derive the peak primary and secondary pressure, respectively, for the RCP shaft seizure/sheared shaft events.

Waterford 3 RAI #16 Response The limiting events with respect to peak primary and secondary pressures are the loss of condenser vacuum (LOCV) and feedwater line break accidents. These are the limiting events because they have the closest approach to the Technical Specification Section 2.1.2 (Reactor Coolant System Pressure) safety limit of 2750 psia and secondary pressure limit of 1210 psia (110% of design pressure). Since the LOCV peak primary pressure is greater than the opening setpoint of the pressurizer safety valves (PSVs), this is only applicable to the events that open the PSVs. The loss of normal feedwater event results in a primary pressure increase of less than 20 psi and has peak primary pressure that is less than the opening pressure of the PSVs. Thus, the loss of normal feedwater event would provide bounding primary pressure results for those less bounding events that do not exceed the PSV setpoints.

The single reactor coolant pump (RCP) shaft seizure/sheared shaft event (SR/SS)

(UFSAR Table 15.3-3) has a peak primary pressure less than the PSV opening setpoints.

The SR/SS event peak primary pressure was increased by 20 psi.

Peak secondary pressure increases are based on the LOCV events because it provides the limiting secondary pressure results. The LOCV event results in a secondary pressure increase of less than I psi for all secondary pressure cases. The SR/SS event peak secondary pressure was increased by I psi.

to W3F1 -2015-0062 Page 17 of 21 NRC RAI #17 The fourth paragraph in Attachment 2, page 7 states, in part, that:

The steam generator blowdown is not impacted by the revised CEA drop time...

The steam generator (SG) blowdown rate was dependent on the SG pressure, however, the impact of the revised CEA drop time on the SG pressure response was not sufficiently discussed in the LAR. Provide a more detailed discussion on the impact of the revised CEA drop time on the SG pressure response.

Waterford 3 RAt #17 Response The steamline break events are analyzed at hot full power (HFP), hot zero power (HZP),

and Mode 3 and 4 conditions. The main steam system pipe break causes an increase in steam flow resulting in an increase in energy removal from the affected steam generator (SG). The increased affected SG energy removal causes a decrease in the overall reactor coolant system (ROS) temperatures and pressure. In the presence of a negative moderator temperature coefficient (MTC), the RCS cooldown causes positive reactivity to be added to the core.

The initial SG pressures are based upon the beginning operating conditions. The post steam line break SG pressures fall together until a Main Steam Isolation Signal (MSIS) is actuated and the Main Steam Isolation Valves (MSIVs) close. Once the MSIVs close, the affected SG continues to blowdown whereas the unaffected SG pressure recovers or maintains (refer to UFSAR Figure 15.1-37 as an example). The affected SG blowdown duration is dominated by the SG water inventory, initial temperatures, and break size. The increased CEA drop time impacts the power reduction post-trip and the amount of energy deposited into the primary coolant system. The slightly longer drop time means a slight increase in the amount of energy added to the system. The slight increase in energy would have a minimal impact on the SG pressure and no change to the event characteristics.

NRC RAI #18 The second to last paragraph in Attachment 2, page 9, states:

The loss of feedwater flow event demonstrated that the impact of the revised CEA drop time on long term parameters is insignificant. Hence there is an insignificant impact on the plant cooldown to shutdown cooling conditions post-trip.

Provide justification for the use of the results of the loss of feedwater flow event to the steam line break long term cooldown.

Waterford 3 RAI #18 Response The steamline break events are analyzed at hot full power (HFP), hot zero power (HZP),

and Mode 3 and 4 conditions. The main steam system pipe break causes an increase in steam flow resulting in an increase in energy removal from the affected steam generator (SG). The increased affected SG energy removal causes a decrease in the overall reactor to W3FI-201 5-0062 Page 18 of 21 coolant system (RCS) temperatures and pressure. The post steam line break SG pressures fall together until a Main Steam Isolation Signal (MSIS) is actuated and the Main Steam Isolation Valves (MSIVs) are closed. Once the MSIVs are closed, the affected SG continues to blowdown whereas the unaffected SG pressure recovers or maintains (refer to UFSAR Figure 15.1-37 as an example).

The increased CEA drop time impacts the power reduction post-trip and the amount of energy deposited into the primary coolant system. The slightly longer drop time means a slight increase in the amount of energy added to the system. The Loss of Normal Feedwater Flow event (FSAR 15.2.3.2) was chosen to evaluate the transient characteristics with ,respect to energy deposition and associated steam releases which would be applicable to all the events. The analysis showed that the differences in primary and secondary system energy after reactor trip are insignificant. As time increases farther past the time of CEA rod insertion, the differences of the impact of the revised CEA drop time become negligible. NRC RAI#15 Figure 15-2 shows the energy deposition from reactor trip time to 1800 seconds for the loss of normal feedwater event. For the long term response, the energy deposition change between the AOR and the revised CEA drop time analysis was 0.46% for the loss of normal feedwater event which is considered negligible in the long term response.

NRC RAI #19 The fourth paragraph in Attachment 2, page 16, states, in part, that:

The increase in peak primary pressure was based on the loss of feedwater long term results which showed an increase in peak primary pressure of less than 20 psi...

Provide justification for the use of the loss of feedwater long term results to the total loss of forced reactor coolant flow event in determining the peak primary pressure.

Waterford 3 RAI #19 Response The limiting events with respect to peak primary and secondary pressures are the loss of condenser vacuum (LOCV) and feedwater line break accidents. These are the limiting events because they have the closest approach to the Technical Specification Section 2.1.2 (Reactor Coolant System Pressure) safety limit of 2750 psia and secondary pressure limit of 1210 psia (110% of design pressure). Since the LOCV peak primary pressure is greater than the opening setpoint of the pressurizer safety valves (PSVs), this is only applicable to the events that open the PSVs. The loss of normal feedwater (LOFW) event results in a primary pressure increase of less than 20 psi and has peak primary pressure that is less than the opening pressure of the PSVs. Thus, the LOFW event would provide bounding primary pressure results for those less bounding events that do not exceed the PSV setpoints.

The total loss of forced reactor coolant flow event (LOF) (UFSAR Table 15.3-1) has a peak primary pressure less than the PSV opening setpoints. Thus, the LOF event peak primary pressure was increased by less than 20 psi.

Attachment 1 to W3F1 -2015-0062 Page 19 of 21 NRC RAI #20 The second to last paragraph in Attachment 2, page 18 states, in part, that:

The loss of condenser vacuum event peak primary and secondary pressure increases would bound that expected for the uncontrolled CEA withdrawal from subcritical conditions.

Further, the second to last paragraph in Attachment 2, page 19, states, in part:

The loss of condenser vacuum event peak primary and secondary pressure increases would bound that expected for the uncontrolled CEA withdrawal from low power condition.

/Provide justification for the use of the loss of condenser vacuum results to both cases of the CEA withdrawal events in determining the peak primary and secondary pressure increases due to the CEA drop time changes.

Waterford 3 RAI #20 Response The limiting events with respect to peak primary and secondary pressures are the loss of condenser vacuum and feedwater line break accidents. These are the limiting events because they have the closest approach to the Technical Specification Section 2.1.2 (Reactor Coolant System Pressure) safety limit of 2750 psia and secondary pressure limit of 1210 psia (110% of design pressure). UFSAR Section 15.4 contains the reactivity and power distribution anomalies. These events are less adverse than the loss of condenser

• vacuum with respect to peak primary and secondary pressure. Specifically, loss of condenser vacuum peak primary pressure is 2711 psia and secondary pressure is 1181 psia [UFSAR Section 15.2.1.3.3.3] whereas the CEA withdrawal event peak primary pressure is 2287 psia [UFSAR Table 15.4-7] and secondary pressure is 1085 psia [UFSAR Table 15.4-3].

The increased CEA drop time impacts the power reduction post-trip and the amount of energy deposited into the primary coolant system. The slightly longer drop time means a slight increase in the amount of energy added to the system. The assessment provided in W3F1-2015-0040 used the loss of condenser vacuum event because it produced the largest post-trip primary and secondary pressure spike due to losing its secondary heat removal capability. Since, loss of condenser vacuum event produces a more adverse pressure transient, the slight energy deposition increase would be expected to have the most adverse impact on this event. Thus, the loss of condenser vacuum event results would continue to bound the CEA withdrawal event pressure transients.

NRC RAI #21 Attachment 1, page 2, of the August 14, 2015, supplement provides the CEA drop time curve for the 0.8 second holding coil decay time (Curve 3). This curve has the initial CEA motion delayed 0.2 seconds over the revised curve {Curve 2). However, by 1.15 seconds, both curves are identical. Given that the CEAs insert via gravity, justify how the CEAs that to W3F1 -2015-0062 Page 20 of 21 start dropping at 0.8 seconds (Curve 3) are at the same position as CEAs that started dropping at 0.6 seconds (Curve 2) by 1.15 seconds (1 0 percent inserted).

Waterford 3 RAI #21 Response Letter W3FI-2015-0061 [Reference 3.5] presented the following CEA drop time curves.

W3F1-2015-0061 Table 3.0-1. CEA Drop Times CEA Insertion Curve 1 Curve 2 Curve 3

(%) AOR Time Revised Time Increased Holding (Seconds) (Seconds) Coil Time (Seconds) 0 0.00 0.00 0.00 0 0.60 0.60 08 5 0.80 0.95 1.00_______

10 0.95 1.15 11 20 1.25 1.45 14 30 1.55 1.75 17 40 1.80 2.00 2.00______

50 2.05 2.25 22 60 2.3 2.50 2.50 70 2.535 2.75 2.75 80 2.75 2.95 2.95 90 3.0 3.20 3.20 100 3.5 3.50 3.50 Curve 1 was submitted to the NRC in letter W3P89-3094 [Reference 3.11] and subsequently approved in NRC Technical Specification Amendment 58 [Reference 3.12].

Curve 1 was based upon letter W3P89-3094 Figure A1-2.3 which was generated to bound the measured data from the CEA drop time testing performed at the beginning of Cycle 3.

Letter W3P89-3094 Figure A1-3.1 and Appendix Al-A provided the Cycle 3 CEA drop time test data.

Curve 2 and Curve 3 are both based upon the Curve 1 shape. Curve 2 was generated assuming a failure mode that increased the CEA insertion time with no change to the initial CEA drop start time. Curve 3 was generated assuming a failure mode that increased the CEA holding coil decay time which delayed the initial CEA drop start time. Curve 2 and Curve 3 both reach the same CEA drop time at 10% insertion. This is because Curve 2 conservatively assumed the CEA slowed down by 0.2 seconds over Curve 1 by the 10%

CEA insertion and Curve 3 conservatively used an additional initial delay of 0.2 seconds over Curve 1.

Curve 2 and Curve 3 are not intended to match the expected test data curves but to bound the test data to ensure conservative analysis results.

to W3F1 -201 5-0062 Page 21 of 21

3.0 REFERENCES

3.1 Entergy Pre-submittal Meeting Summary for the Revised Control Element Assembly Drop Times [ADAMS Accession Number ML15117A503].

3.2 Entergy Pre-Submittal Meeting Revised Presentation Slides for Control Element Assembly Drop Times [ADAMS Accession Number ML15113A787].

3.3 W3F1-2015-0040, License Amendment Request to Revise Control Element Drop Times, July 2, 2015 [ADAMS Accession Number ML15197A106].

3.4 NRC Letter, Regarding License Amendment Request to Revise Control Element Assembly Drop Times, Unacceptable with Opportunity to Supplement, August 3, 2015 [ADAMS Accession Number ML15205A306].

3.5 W3F1-2015-0061, Supplement to Revise Control Element Assembly Drop Times Associated with Technical Specification 3.1.3.4, August 13, 2015 2015 [ADAMS Accession Number ML15226A346].

3.6 NRC CEA Drop Time Submittal Request for Additional Information, August 26, 2015, [ADAMS Accession Number ML15232A275].

3.7 Waterford Nuclear Generator Station Unit 3, Technical Specifications, Through Amendment 245.

3.8 Waterford Nuclear Generator Station Unit 3, Update Final Safety Analysis Report, Revision 308.

3.9 W3F1-2003-0074, Extended Power Uprate License Amendment Request, November 13, 2003 [NRC ADAMS Accession Number ML040260317].

3.10 NRC Safety Evaluation Report, Waterford Steam Electric Station, Unit 3 -

Issuance of Amendment 199 RE: Extended Power Uprate, April 15, 2005 [NRC ADAMS Accession Number ML05I1030082].

3.11 Waterford 3 Letter W3P89-3094, Technical Specification Change Request, Average Control Element Assembly Drop Time, August 14, 1989.

3.12 NRC Waterford 3 Technical Specification Amendment 58, Average Control Element Assembly Drop Time, October 31, 1989 [ADAMS Accession Number ML021760257].

3.13 W3F1-2004-0053, Alternate Source Term License Amendment Request, July 15, 2004 [NRC ADAMS Accession Number ML042020294].

3.14 NRC Safety Evaluation Report, Waterford 3 Amendment 198, Full Scope Implementation of an Alternative Accident Source Term, March 29, 2005 [NRC ADAMS Accession Number ML050890248].

3.15 NRC Regulatory Guide 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors.

3.16 W3F1-2004-0071, Supplement to Alternate Source Term Submittal, August 19, 2004 [NRC ADAMS Accession Number ML042360712].

Attachment 2 to W3F1 -2015-0062 Fuel Thermal Conductivity Degradation Evaluation Affidavit Affidavit to Withhold from Public Disclosure Proprietary Information Under 10 CFR 2.390 As Attachment 4 contains information proprietary to Westinghouse Electric Company LLC, it is supported by an Affidavit signed by Westinghouse, the owner of the information. The Affidavit 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 Section 2.390 of the Commission's regulations.

to W3F1 -2015-0062 Page 1 of 7 W estin houseWestinghouse Electric Company Whouse1000 estin Westinghouse Drive Cranbenry Township, Pennsylvania 16086 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 CAW-15-4272 September 11,2015 APPLICATION FOR WITHHOLDING PROPRIETARY INFORMATION FROM PUBLIC DISCLOSURE

Subject:

CE-15-284-P, Revision 1, Attachment 2, "Fuel Management Adjustment to [Radial Fall-offf*c to Reserve Margin for Thermal Conductivity Degradation." (Proprietary)

The Application for Withholding Proprietary Information from Public Disclosure is submitted by Westinghouse Electric Company LLC (Westinghouse), pursuant to the provisions of paragraph (b)( 1) of Section 2.390 of the Commission's regulations. It contains commercial strategic information proprietary to Westinghouse and customarily held in confidence.

The proprietary information for which withholding is being requested in the above-referenced report is further identified in Affidavit CAW-15-4272 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 Conmmission's regulations.

Accordingly, this letter authorizes the utilization of the accompanying Affidavit by Entergy Operations, Inc.

Correspondence with respect to the proprietary aspects of the Application for Withholding or the Westinghouse Affidavit should reference CAW-I15-4272, 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.

amnes A. Gresham, Manager Regulatory Compliance to W3F1 -2015-0062 Page 2 of 7 CAW- 15-4272 September 11, 2015 AFFIDAVIT COMMONWEALTH OF PENNSYLVANIA:

SS COUNTY OF BUTLER:

I, James A. Gresham, am 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 my knowledge, information, and belief.

"/James A. Gresham, Manager Regulatory Compliance to W3F1 -2015-0062 Page 3 of 7 2 ~CAW-1 5-4272 (1) 1am 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) 1am making this Affidavit in conformance with the provisions of 10 CFR Section 2.3 90 of the Commission's regulations and in conjunction with the Westinghouse Application for Withholding Proprietary Information from Public Disclosure accompanying this Affidavit.

(3) 1 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 paragr'aph (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 to W3F1 -201 5-0062 Page 4 of 7 3 ~CAW-] 5-4272 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 fu~nded development plans and programs of potential commercial value to Westinghouse.

(1) 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. Ittis, 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.

to W3F1 -2015-0062 Page 5 of 7 4 CAW-l 5-4272 (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 CE-15-284-P, Revision 1, Attachment 2, "Fuel Management Adjustment to [Radial Fall-off~c' to Reserve Margin for Thermal Conductivity Degradation" (Proprietary), for submittal to the Commission, being transmitted by Entergy Operations, Inc. 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 Fuel Performance, Safety Analysis and the associated Thermal Conductivity Degradation methodologies, and may be used only for that purpose.

to W3F1 -2015-0062 Page 6 of 7 5 CAW-15-4272 (a) This information is part of that which will enable Westinghouse to:

(i) Perform Reload Fuel and Safety analyses.

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

(i) Westinghouse plans to sell the use of similar information to its customers for the purpose of performing reload fuel and safety analyses.

(ii) Westinghouse can sell support and defense of industry guidelines and acceptance criteria for plant-specific applications.

(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.

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.

to W3F1 -2015-0062 Page 7 of 7 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 2.3 90 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 thle 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.