W3F1-2004-0073, Supplement to Amendment Request NPF-38-249, Extended Power Uprate

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Supplement to Amendment Request NPF-38-249, Extended Power Uprate
ML042440423
Person / Time
Site: Waterford Entergy icon.png
Issue date: 08/25/2004
From: Peters K
Entergy Nuclear South, Entergy Operations
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
W3F1-2004-0073
Download: ML042440423 (21)
v • d • e


Text

Entergy Nuclear South I'nMcrgy Operations. Inc.

17265 River Road Entergy Killona. LA 70057 Tcl 504 739 6440 Fax 504 739-6698 k neters('cntermv.com Ken Peters I)irector, NucIclr Safety Assurance Waterford 3 W3FI-2004-0073 August 25, 2004 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555

SUBJECT:

Supplement to Amendment Request NPF-38-249, Extended Power Uprate Waterford Steam Electric Station, Unit 3 Docket No. 50-382 License No. NPF-38

REFERENCES:

1. Entergy Letter dated November 13, 2003, "License Amendment Request NPF-38-249 Extended Power Uprate"
2. NRC Letter dated June 21, 2004, "Waterford Steam Electric Station, Unit 3 (Waterford 3) - Request for Additional Information Related to Revision to Facility Operating License and Technical Specifications -

Extended Power Uprate Request (TAC No. MC1355)"

3. Entergy Letter dated July 28, 2004, "Supplement to Amendment Request NPF-38-249 Extended Power Uprate"
4. Entergy Letter dated August 10, 2004, "Supplement to Amendment Request NPF-38-249 Extended Power Uprate"
5. Entergy Letter dated July 14, 2004, "Supplement to Amendment Request NPF-38-249 Extended Power Uprate"

Dear Sir or Madam:

By letter (Reference 1), Entergy Operations, Inc. (Entergy) proposed a change to the Waterford Steam Electric Station, Unit 3 (Waterford 3) Operating License and Technical Specifications to increase the unit's rated thermal power level from 3441 megawatts thermal (MWt) to 3716 MWt.

By letter (Reference 2), the Nuclear Regulatory Commission (NRC) staff requested additional information (RAI) related to reactor systems. By letters (Reference 3 and Reference 4)

Entergy responded to 60 of the 61 questions and committed to provide the remaining response in a future supplement. Entergy's response to the last unanswered question is contained in Attachment 1 to this letter.

The Waterford 3 Extended Power Uprate (EPU) Power Uprate Report (PUR) was submitted as Attachment 5 to Reference 1. The control element assembly (CEA) ejection analysis Izo I

W3Fl-2004-0073 Page 2 of 3 presented in Section 2.13.4.3.2 of the PUR included the results of a peak reactor coolant system (RCS) pressure case. The peak pressure case has been reanalyzed to conservatively prevent actuation of the pressurizer sprays and to model the occurrence of a turbine trip following reactor trip. The revised analysis indicates that peak calculated RCS pressure increased from 2519 psia previously reported in Reference 1to 2613 psia, a value less than the acceptance criterion of 2750 psia (i.e., 110% of RCS design pressure). Revised PUR pages are provided in Attachment 2 and supersede the corresponding pages previously provided in Reference 1.

Following conference calls with members of the NRC staff, Entergy is providing additional information in support of the EPU.

  • Attachment 3 contains additional information regarding the EPU containment analysis.
  • Attachment 4 contains additional information regarding the development of the EPU steam generator pressure - low setpoint and its allowable value. Note that an arithmetic error was identified that results in a steam generator pressure - low allowable value different from that proposed in Reference 1. The error has been entered into Entergy's 10 CFR 50 Appendix B Corrective Action Program and has been corrected in the attached calculations. Revised technical specification mark-ups for technical specification pages 2-3, 3/4 3-19, and 3/4 3-20 will be provided in a future supplement to replace those previously provided.
  • Attachment 5 contains additional information regarding the EPU spent fuel cooling analysis.

The no significant hazards consideration included in Reference 5 is not affected by any information contained in this letter. The submittal includes one new commitment as summarized in Attachment 6.

If you have any questions or require additional information, please contact D. Bryan Miller at 504-739-6692.

I declare under penalty of perjury that the foregoing is true and correct. Executed on August 25, 2004.

Sincerely, Atta hments:

1. Response to Request for Additional Information
2. Revised Control Element Assembly Ejection Peak Pressure Analysis Results
3. Additional Information Related to EPU Containment Analysis
4. Additional Information Regarding EPU Steam Generator Pressure - Low Setpoint Development
5. Additional Information Regarding EPU Spent Fuel Pool Cooling Analysis
6. List of Regulatory Commitments

W3F11-2004-0073 Page 3 of 3 cc: Dr. Bruce S. Mallett U. S. Nuclear Regulatory Commission Region IV 611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011 NRC Senior Resident Inspector Waterford 3 P.O. Box 822 Killona, LA 70057 U.S. Nuclear Regulatory Commission Attn: Mr. Nageswaran Kalyanam MS O-07D1 Washington, DC 20555-0001 Wise, Carter, Child & Caraway Attn: J. Smith P.O. Box 651 Jackson, MS 39205 Winston & Strawn Attn: N.S. Reynolds 1400 L Street, NW Washington, DC 20005-3502 Louisiana Department of Environmental Quality Office of Environmental Compliance Surveillance Division P. 0. Box 4312 Baton Rouge, LA 70821-4312 American Nuclear Insurers Attn: Library Town Center Suite 300S 29th S. Main Street West Hartford, CT 06107-2445

Attachment I To W3Fl-2004-0073 Response to Request for Additional Information to W3F1-2004-0073 Page 1 of 5 Response to Request for Additional Information Question 20:

Please provide a tabulation to indicate that for each event, what specific acceptance criteria are satisfied, to demonstrate that the general acceptance criteria of the event's class are met?

Response 20:

Table RAI 20-1 contains the regulatory limits per the NUREG-0800 Standard Review Plan acceptance criteria to which the events were evaluated and analyzed for the power uprate project.

to W3Fl-2004-0073 Page 2 of 5 TABLE RAI 20-1 EPU Section Event Pressure Criteria Barrier Integrity 2.13.1.1.1 Decrease in Feedwater Temperature Max Pressures

  • 110% of Design No Specified Acceptable Fuel Design Limit (SAFDL) Violation Doses
  • 10% of 10CFR100 2.13.1.1.2 Increase in Feedwater Flow ' Max Pressures 5 110% of Design No SAFDL Violation Doses
  • 110% of Design No SAFDL Violation Doses
  • 10% of 10CFR100 2.13.1.1.4 Inadvertent Opening of a Steam Max Pressures < 110% of Design No SAFDL Violation Generator ADV (IOADV) Doses
  • 110% of Design SAFDL Violation Permitted with Single Active Failure (SAF) ' Doses 5 10% of 10CFR100 2.13.1.2.2 Increase in Feedwater Flow with SAF Max Pressures
  • 110% of Design SAFDL Violation Permitted 1 Doses 5 10% of 10CFRIOO 2.13.1.2.3 Increased Main Steam Flow with Max Pressures 5 110% of Design SAFDL Violation Permitted SAF Doses 5 10% of 10CFR100 2.13.1.2.4 IOADV with Loss of Offsite Max Pressures
  • 110% of Design SAFDL Violation Permitted 2 Power (LOOP) Doses
  • 10% of 10CFR100 2.13.1.3.1 Steam System Piping Failures Post- Max Pressures
  • 110% of Design SAFDL Violation Permitted Trip Analysis Doses5 10CFR100 3 2.13.1.3.2 Mode 3 and 4 All Rods In (ARI) Max Pressures
  • 110 % of Design SAFDL Violation Permitted Return to Power (RTP) Steam Line Doses 5 10CFR100 Break (SLB) 2.13.1.3.3. Steam System Piping Failures Pre- Max Pressures
  • 110% of Design SAFDL Violation Permitted Trip Power Excursion Doses
  • 10CFR100 3 2.13.2.1.1 Loss of External Load ' Max Pressures
  • 110% of Design No SAFDL Violation I I I Doses_ 10% of1 OCFR100 to W3Fl-2004-0073 Page 3 of 5 TABLE RAI 20-1 EPU Section Event Pressure Criteria Barrier Integrity 2.13.2.1.4 Loss of Normal Alternating Max Pressures
  • 110% of Design No SAFDL Violation Current (AC) Power ' Doses *10% of 10CFR100 2.13.2.1.5 Steam Pressure Regulator Failure 1 Max Pressures
  • 110% of Design No SAFDL Violation Doses
  • 10% of 10CFR100 2.13.2.2.1 Loss of External Load with SAF 1 Max Pressures 110% of Design SAFDL Violation Permitted 2

._ Doses

  • 10% of IOCFR100 2.13.2.2.2 Turbine Trip with SAF 1 Max Pressures
  • 110% of Design SAFDL Violation Permitted 2 Doses
  • 110% of Design SAFDL Violation Permitted 2 SAF
  • Doses 10% of 10CFR100 2.13.2.2.4 Loss of Normal AC Power with SAF Max Pressures
  • 110% of Design SAFDL Violation Permitted Doses *10% of 10CFR100 2.13.2.2.5 Loss of Normal Feedwater Flow Max Pressures 110% of Design No SAFDL Violation Doses
  • 120% of Design SAFDL Violation Permitted 2 Large Breaks GIS Doses
  • 10% of 10CFR100 (Max Pressure < 110% of Design for PIS Doses
  • 10% of 10CFR1 00 5 small FWLB with offsite power available) 2.13.2.3.2 Loss of Normal Feedwater Flow with Max Pressures
  • 110% of Design SAFDL Violation Permitted 2 SAF Doses *10% of IOCFRIOO 2.13.3.1.1 Partial Loss of Forced Reactor Max Pressures
  • 110% of Design No SAFDL Violation Coolant Flow ' Doses
  • 10% of 10CFR100 2.13.3.2.1 Total Loss of Forced Reactor Max Pressures
  • 110% of Design No SAFDL Violation Coolant Flow Doses *10% of 10CFR100 2.13.3.2.2 Partial Loss of Forced Reactor Max Pressures
  • 110% of Design SAFDL Violation Permitted Coolant Flow with SAF ' Doses
  • 110% of Design SAFDL Violation Permitted Shaft Seizure/Sheared Shaft Doses
  • 110% of Design No SAFDL Violation Assembly (CEA) Withdrawal from Doses
  • 10% of 10CFR100 Subcritical 2.13.4.1.2 Uncontrolled CEA Withdrawal from Max Pressures
  • 110% of Design No SAFDL Violation Low Power Doses *10% of 10CFR100 2.13.4.1.3 Uncontrolled CEA Withdrawal at Max Pressures
  • 110% of Design No SAFDL Violation Power Doses
  • 10% of 10CFR100 2.13.4.1.4 CEA Misoperation Max Pressures < 110% of Design No SAFDL Violation Doses : 10% of 10CFR100 2.13.4.1.5 Inadvertent Boron Dilution Max Pressures
  • 110% of Design No SAFDL Violation Doses < 10% of 10CFR100 Time Between Alarm and Loss of Shutdown Margin:

> 30 Minutes, Mode 6

_ 15 Minutes, other Modes 2.13.4.1.6 Startup of an Inactive Reactor Max Pressures

  • 110% of Design No SAFDL Violation Coolant Pump Shutdown Margin 2.13.4.1.7 CEA Withdrawal Modes 3,4 and 5 Max Pressures
  • 110% of Design No SAFDL Violation ARI Doses < 10% of 10CFR100 2.13.4.3.1 Inadvertent Loading of a Fuel Max Pressures
  • 110% of Design SAFDL Violation Permitted 2 Assembly into an Improper Position Doses
  • 10% of 10CFR100 2.13.4.3.2 Control Element Assembly Ejection Max Pressures
  • 110% of Design SAFDL Violation Permitted Doses < IOCFR100 2.13.5.1.1 Chemical & Volume Control System Max Pressures
  • 110% of Design No SAFDL Violation (CVCS) Malfunction Doses
  • 10% of 10CFR100 2.13.5.1.2 Inadvertent Emergency Core Cooling Max Pressures
  • 110% of Design No SAFDL Violation System (ECCS) Doses
  • 10% of 10CFR100 2.13.5.2.1 CVCS Malfunction with SAF Max Pressures
  • 110% of Design No SAFDL Violation I I_ Doses* _10% of IOCFR100 to W3F1 -2004-0073 Page 5 of 5 TABLE RAI 20-1 EPU Section Event Pressure Criteria Barrier Integrity 2.13.6.3.1 Small Primary Line Break Outside Max Pressures
  • 110% of Design SAFDL Violation Permitted 2 Containment GIS Doses
  • 110% of Design No SAFDL Violation

_ Doses

As documented in Power Uprate Report (PUR) Table 2.13.0-1, this event is bounded by a different Final Safety Analysis Report (FSAR) event.

2 Waterford 3 PUR analyses meet the radiological acceptance criteria by demonstrating no fuel failure for these events.

3 The radiological consequences for the Return to Power Main Steam Line Break (MSLB) and Pre Trip Power Excursion MSLB are added together and compared to the acceptance limits of 10CFR1 00.

4 Applicability of 10CFR100 criteria to Letdown Line Break Doses with pre-existing Iodine spike (PIS) was approved via January 8, 2003, letter from NRC to Waterford 3. (TAC No. MB3231). NUREG-0800 does not address the PIS case for this event.

5 NUREG-0800 Section 15.2.8 acceptance criteria for Feedwater Line Break (FWLB) are a small fraction of 10CFR100 based on an accident generated iodine spike (GIS). The case of a Pre-existing iodine spike is not addressed.

Waterford 3, via PUR Section 2.13.2.3.1.6, applied the small fraction acceptance criteria for the case of a Pre-existing Iodine spike.

Attachment 2 To W3Fl-2004-0073 Revised Control Element Assembly Ejection Peak Pressure Analysis Results to W3Fl-2004-0073 Revised Control Element Assembly Ejection Peak Pressure Analysis Results None of the discussion included in Section 2.13.4.3.2 of the Power Uprate Report (PUR)

(Reference 1, Attachment 5) or its subsections (Sections 2.13.4.3.2.1 through 2.13.4.3.2.6) are impacted by the revised peak Reactor Coolant System (RCS) pressure case. The discussion included in these sections does not directly report the Control Element Assembly (CEA)

Ejection peak RCS pressure case results. Details of the peak RCS pressure case are only discussed in the tables and figures included with Section 2.13.4.3.2 of the PUR. Hence, only the PUR Section 2.13.4.3.2 tables and figures that report peak pressure case results are impacted. PUR tables and figures reporting "full power" CEA Ejection case results remain unchanged. Moreover, the initial assumptions for the CEA Ejection peak pressure case reported in PUR Table 2.13.4.3.2-3 were not changed for the new case performed. Hence this PUR table remains unchanged.

Changes to the PUR tables and figures are summarized below.

  • PUR Tables 2.13.4.3.2-1, 2.13.4.3.2-2 and 2.13.4.3.2-3 are UNCHANGED.
  • The information in PUR Table 2.13.4.3.2-4, the RCS Peak Pressure Sequence of Events, is replaced.
  • PUR Table 2.13.4.3.2-5 is UNCHANGED.
  • PUR Figures 2.13.4.3.2-1 through 2.13.4.3.2-6 are UNCHANGED.
  • The core power vs. time response in PUR Figure 2.13.4.3.2-7 is replaced.
  • The core heat flux vs. time response in PUR Figure 2.13.4.3.2-8 is replaced.
  • The core coolant temperatures vs. time response in PUR Figure 2.13.4.3.2-9 is replaced.
  • The RCS pressure vs. time response in PUR Figure 2.13.4.3.2-10 is replaced.
  • The SG pressure vs. time response in PUR Figure 2.13.4.3.2-11 is replaced.
  • The reactivity components vs. time response in PUR Figure 2.13.4.3.2-12 is replaced.

Waterford 3 Extended Power Uprate Table 2.13.4.3.2.4 CEA Ejection Peak RCS Pressure Sequence of Events 3716 MWt Event 3716 MWt EPU EPU Time Setpoint I Value (sec) 0.00 Mechanical Failure of CEDM causes CEA to eject.

0.05 CEA fully ejected 0.07 CPC VOPT, % of Full Power_ 163 0.08 Maximum core power occurs, % 187.0 of full power 0.699 __ Trp a.kers open _

1.299 CEA's begin to drop into core -

2.9 Maximum RCS pressure, PSIA 2613*

4.8 CEA ully inserted, core power -

reduced to below 10% power

  • 2597 PSIA for BOC Cycle 1 HFP CEA Ejection.

2.13-341

Waterford 3 Extended Power Uprate Ii ------ - i I

I f

200%

I 180% i 3:

0 160%

IL i i

-J Us IL 140% iI IL 0 120%

C., 100% i W I 80%

i w Ii a- 60% II 0 40%

IL, iI 20% i zi i

0%

i 0 5 10 15 20 25 30 i TIME, SECONDS i

Figuro 2.13.4.3.2-7 CEA Ejection Core Power vs. Time for Peak RCS Pressure 2.13-349

Waterford 3 Extended Power Uprate 140%

120%

0 U. AL 0 0~ 100%/o

-I I- -J 0 80%

IL x

'X 600/%

-J

-J LU AL LL I'- I-40%/o II II 0 LU 20%

U-0%/0 0 5 10 15 20 25 30 TlIME SECONDS Figure 2.13.4.3.2-8 CEA Ejection Core Heat Flux vs. Time for Peak RCS Pressure 2.13-350

Waterford 3 Extended Power Uprate Tout =aveagcore et Tag = a,-e aerage erzcahxe TM oow mfr teni~eab Tin oormeirfetbupma 620 Tout wu600 re. 580 Iy-v LJs

~560 052 C.)

500, 0 5 10 15 20 25 30j TIiiSEaNDS Figure 2.13.4.3.2-9 CEA Ejactlon Cora Coolant Tcrnporaturcs vs. Timo for Peak RCS Pressure 2.13-351

Waterford 3 Extended Power Uprate 2700 i

i 2600 I 2500 I I

II Cl) i

,, 2400 i i

i gnn 2300 ----------

i Cl, I 0y 0 2200 i

i 2100 i

I 2000 0 5 10 15 20 25 30 i II TIME SECONDS Figure 2.13.4.3.2.10 CEA Ejection Peak RCS Pressure vs. Time for Peak RCS Pressure 2.13-352

Waterford 3 Extended Power Uprate

1 1100

< 1050 Cl) a._ 1000 M 950 Cl)

Cl) il 900 a 8 50 0

i° 800 cX 750 z

W 700 2 650 U- 600 I___

Ul)

ID 5 10 15 20 25 30 i TIRE, SBEONDi Figure 2.13.4.3.2-11 CEA Ejection SG Pressure vs. Time for Peak RCS Pressure 2.13-353

Waterford 3 Extended Power Uprate 1 DOPPLER 0~  : MODERATOR z

Wt a.

I-3 w-4 TOTAL aI 5 \CEA's AFTER SCRAM 0 10 20 30 TIME, SECONDS Figure 2.13.4.3.2-12 CEA Ejection Reactivity Components vs. Time for Peak RCS Pressure 2.13-354

Attachment 3 To W3FI -2004-0073 Additional Information Related to EPU Containment Analysis to W3F1-2004-0073 Page 1 of 2 Additional Information Related to EPU Containment Analysis Question 1:

An October 28, 1999 Waterford Unit 3 letter to the NRC provides a resolution to the penetration overpressurization issue of GL 96-06. A December 22, 1997 Waterford Unit 3 letter states that a containment atmosphere temperature of 260 F was assumed for the Waterford Unit 3 analyses. Please explain why this resolution is still valid for the power uprate.

In particular, please explain why the pre-power uprate analyses remain valid when the peak temperature from the main steam line break is greater than 260 F.

Response 1:

The 260'F temperature reported in the December 22, 1997, letter was used to justify operability of the penetrations and continued operation during the period while penetrations susceptible to the overpressurization condition described in Generic Letter (GL) 96-06 could be reconfigured, modified, or drained to eliminate the potential for overpressurization. These changes, as committed to in response to GL 96-06, necessary to eliminate the potential for penetration overpressurization due to a temperature excursion from a DBA inside containment have been completed. Extended Power Uprate (EPU) does not propose to operate the containment penetrations differently or modify containment environmental parameters such that a containment penetration would become susceptible to conditions described in GL 96-06.

EPU does not invalidate the measures taken to address the potential containment penetration overpressurization concern.

Question 2:

The power uprate submittal proposes using EQ temperature envelopes as the criteria for containment temperature for the main steam line break and LOCA analyses. (i) Explain why the containment design temperature of 263 'F is not used for the criterion. (ii)Since both the LOCA and main steam line break temperatures are greater than 263 'F, how is the containment design temperature satisfied?

Response 2:

(i) The originally determined peak loss of coolant accident (LOCA) and main steam line break (MSLB) temperatures (i.e., 269 'F and 413 'F respectively) were used in establishing the EQ temperature envelope as discussed in Section 3.11.3.3.1 of the original safety evaluation report. Thus the original licensing basis established the maximum allowed temperatures which were adopted as the acceptance criteria for the LOCA and MSLB. The adoption of the limits (i.e., 269.3 'F for LOCA and 413.5 'F for MSLB) are discussed in the Bases for Technical Specification 3/4.6.1.5, "Air Temperature." As discussed below, the containment steel vessel does not exceed the 263-F containment design temperature during the LOCA and MSLB events.

(ii) A calculation was done for Waterford 3 in 1985 to determine the maximum steel containment vessel wall temperature post-LOCA and post-MSLB. The "Maximum Steel Vessel (Containment) Temperature Estimation" calculation shows:

to W3F1-2004-0073 Page 2 of 2

- For a peak LOCA temperature of 269.1 F the maximum containment vessel steel temperature would be 257.4-F, and

- for a peak MSLB temperature of 413.5-F the maximum containment vessel steel temperature would be 233.5-F.

The calculation notes: 'The duration of the MSLB is very much smaller compared to LOCA. Therefore, LOCA conditions determine the maximum steel vessel temperature

'l Therefore, since the EPU peak post-LOCA temperature (254.4-F) and the peak post-MSLB temperature (394.7TF) are less than those assumed in the maximum containment vessel steel temperature calculation discussed above, the maximum containment vessel steel temperature will remain below the "design temperature" of 263-F.

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