W3F1-2004-0014, License Amendment Request, Deletion of Pressurizer Heatup and Cooldown Limits

License Amendment Request, Deletion of Pressurizer Heatup and Cooldown Limits

ML040780691
Person / Time
Site:	Waterford
Issue date:	03/15/2004
From:	Venable J Entergy Nuclear South, Entergy Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NPF-38-253, W3F1-2004-0014
Download: ML040780691 (15)

Text

Entergy Nuclear South Entergy Operations, Inc.

17265 River Road Milona, LA 70066 Tel 504 739 6660 OPf Fax 504 739 6678 ivenabl~entergy.com Joseph E. Venable Vice President, Operatbons Waterford 3 W3Fl -2004-0014 March 15, 2004 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555

SUBJECT:

License Amendment Request NPF-38-253 Deletion of Pressurizer Heatup and Cooldown Limits Waterford Steam Electric Station, Unit 3 Docket No. 50-382 License No. NPF-38

REFERENCES:

Entergy letter to the NRC dated November 13, 2003, License Amendment Request NPF-38-249 Extended Power Uprate (W3Fl-2003-0074)

Dear Sir or Madam:

Pursuant to 10 CFR 50.90, Entergy Operations, Inc. (Entergy) hereby requests the following amendment for Waterford Steam Electric Station, Unit 3 (Waterford 3). Entergy proposes to relocate the pressurizer heatup and cooldown limits, Waterford 3 Technical Specification (TS) 3.4.8.2 and the associated surveillance requirements and bases, to the Technical Requirements Manual. The proposed amendment is consistent with the NRC approved Standard Technical Specifications, Combustion Engineering Plants (NUREG-1432).

Waterford 3 submitted to the NRC, by letter dated November 13, 2003 (Reference 1), a proposed change for an extended power uprate, which in part included the deletion of TS 3.4.8.2.c, associated Action b, Table 5.7-1, and SRs 4.4.8.2.2 and 4.4.8.2.3. The approval of the proposed extended power uprate along with the approval of the proposed change in this letter will result in the complete deletion of TS 3.4.8.2.

The proposed change has been evaluated in accordance with 10 CFR 50.91 (a)(1) using criteria in 10 CFR 50.92(c) and it has been determined that this change involves no significant hazards consideration. The bases for these determinations are included in the attached submittal.

The proposed change includes a new commitment as summarized in Attachment 4.

Entergy requests approval of the proposed amendment by December 15, 2004. Once approved, the amendment shall be implemented within 60 days. Although this request is neither exigent nor emergency, your prompt review is requested.

(Do1

W3F1 -2004-0014 Page 2 of 3 If you have any questions or require additional information, please contact Dana Millar at 601-368-5445.

I declare under penalty of perjury that the foregoing is true and correct. Executed on March 15, 2004.

Sincerely, JEV/DM/cbh Attachments:

1. Analysis of Proposed Technical Specification Change

2. Proposed Technical Specification Changes (mark-up)

3. Changes to Technical Specification Bases Pages - For Information Only

4. List of Regulatory Commitments

W3Fl-2004-0014 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 Steam Electric Station, Unit 3 P.O. Box 822 Killona, LA 70066-0751 U.S. Nuclear Regulatory Commission Attn: Mr. Nageswaran Kalyanam Mail Stop 0-7D1 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 W3FI-2004-0014 Analysis of Proposed Technical Specification Change

Attachment I to W3F1 -2004-0014 Page 1 of 4

1.0 DESCRIPTION

This letter is a request to amend Operating License NPF-38 for Waterford Steam Electric Station, Unit 3 (Waterford 3).

The proposed change will revise the Operating License as follows:

Technical Specification (TS) 3.4.8.2, "Pressurizer Heatup/Cooldown" will be re-titled "Pressurizer."

TS 3.4.8.2 Limiting Condition for Operation (LCO), a., the maximum heatup rate, and b., the maximum cooldown rate, will be deleted and moved to the Technical Requirements Manual (TRM).

TS 3.4.8.2 Action a. will be deleted and moved to the TRM.

TS Surveillance Requirement (SR) 4.4.8.2.1 will be deleted and moved to the TRM.

Index page VI will be revised to reflect the revised title for TS 3.4.8.2.

The proposed amendment is consistent with the NRC approved Standard Technical Specifications, Combustion Engineering Plants (NUREG-1432).

TS 3.4.8.2 LCO c., the maximum spray nozzle usage factor, Action b. and SRs 4.4.8.2.2 and 4.4.8.2.3 are proposed for deletion by letter to the NRC dated November 13, 2003, License Amendment Request NPF-38-249, Extended Power Uprate. The combination of the approval of the power uprate along with this proposed amendment will result in the complete deletion of TS 3/4.4.8.2.

2.0 PROPOSED CHANGE

The proposed change will delete TS 3.4.8.2 LCO a. and b., the pressurizer temperature maximum heatup and cooldown rate limits. The associated Action a, SR 4.4.8.2.1 and related TS bases will also be deleted. The pressurizer heatup and cooldown rates will be placed in the TRM which is maintained in accordance with the 10 CFR 50.59 review process.

This approach provides an effective level of regulatory control and provides an appropriate change control process.

3.0 BACKGROUND

The pressurizer is a cylindrical carbon steel vessel with stainless steel internal surfaces. A spray nozzle on the top of the head is used in conjunction with heaters in the bottom head to provide level and pressure control. Overpressure protection is provided by two safety valves.

The pressurizer is supported by a cylindrical skirt welded to the bottom head.

The pressurizer is designed and fabricated in accordance with the ASME Code requirements.

Final Safety Analysis Report (FSAR) Table 5.2-2 provides a listing of the applicable code requirements. The interior surface of the cylindrical shell and upper head is clad with weld deposited stainless steel. The lower head is clad with a Ni-Cr-Fe alloy to facilitate welding of the Ni-Cr-Fe alloy heater sleeves to the shell. A stainless steel safe end is provided on the pressurizer nozzles, after vessel final stress relief, to facilitate field welds to the stainless steel piping. The structural integrity of the pressurizer is assessed by performing inservice inspections in accordance with the ASME Code, Section Xl requirements.

to W3F1 -2004-0014 Page 2 of 4 Five hundred heatup and cooldown cycles were considered in the fatigue analysis during the design life of the components in the reactor coolant system with heating and cooling at a rate of 2000 F/hour between 701F and 6531F for the pressurizer. The heatup and cooldown rate of the system is administratively limited to assure that these limits will not be exceeded. This is based on consideration of both historical plant transient history and projections of transient lifetime occurrences for the components.

A detailed description of the operation of the pressurizer is included in section 5.4.10 of the Waterford 3, FSAR.

Chapter 15 Accident Analysis Review A review of the FSAR Chapter 15 accident analysis concluded that the pressurizer heatup and cooldown rates are not credited in the mitigation or prevention of any accidents and therefore, do not meet the criteria set forth in 10 CFR 50.36 (c) (2) (ii) for inclusion in the TSs.

The pressurizer pressure high and low trips are included as reactor trip set points. The trip set points have been selected to ensure that the reactor core and reactor coolant system are prevented from exceeding their safety limits during normal operation and design basis anticipated operational occurrences and to assist the engineered safety features actuation system in mitigating the consequences of accidents. The pressurizer heatup and cooldown limits do not affect the pressurizer high and low pressure reactor trip set points.

4.0 TECHNICAL ANALYSIS

The proposed amendment is consistent with the NRC approved Standard Technical Specifications, Combustion Engineering Plants (NUREG-1432). The pressurizer heatup and cooldown rates are placed on the pressurizer to prevent non-ductile failure and assure compatibility of operation with the fatigue analysis performed. An engineering evaluation of the continued structural integrity of the pressurizer is required if these limits are exceeded.

The following provides a review of the criteria set forth in 10 CFR 50.36 for TS limiting condition for operations to justify the removal of the TS.

Criterion I - Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

The pressurizer heatup and cooldown rates are not used as an instrumentation system used to detect a significant abnormal degradation of the reactor coolant pressure boundary.

Criterion 2 - A process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

Pressurizer heatup and cooldown rates are not a process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or challenge to the integrity of a fission product barrier.

Criterion 3 - A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

to W3Fl-2004-0014 Page 3 of 4 Pressurizer heatup and cooldown rates are not a structure, system, or component that is part of the primary success path which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

Criterion 4 - A structure, system, or component which operating experience or probabilistic risk assessment has shown to be significant to public health and safety.

The pressurizer heatup and cooldown rates are considered to be non-risk contributors to the core damage frequency and offsite dose assessment models and as such are not part of the Waterford 3 probabilistic risk assessment.

5.0 REGULATORY ANALYSIS

5.1 Applicable Regulatory Requirements/Criteria The proposed changes have been evaluated to determine whether applicable regulations and requirements continue to be met. Entergy has determined that the proposed changes do not require any exemptions or relief from regulatory requirements, other than the Technical Specifications, and do not affect conformance with any General Design Criterion differently than described in the Final Safety Analysis Report.

5.2 No Significant Hazards Consideration Entergy Operations, Inc. (Entergy) proposes to move the Waterford Steam Electric Station, Unit 3 (Waterford 3) Technical Specification (TS) 3.4.8.2, Pressurizer, maximum heatup and cooldown limits to the Technical Requirements Manual (TRM), which is reviewed in accordance with 10 CFR 50.59, "Changes, Test, and Experiments." The associated action statement, surveillance requirement and bases are also proposed for relocation.

Entergy has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92,

'Issuance of amendment," as discussed below:

1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The probability of an accident is unchanged as a result of the proposed change to delete the Waterford 3 pressurizer heatup and cooldown rates and associated action, surveillance requirement, and bases from the TS. The cooldown and heatup rates are not initiators to any accidents or pressurizer transients discussed in the Waterford 3 Final Safety Analysis Report (FSAR). Therefore, the probability of an accident is not changed.

The purpose of the pressurizer heatup and cooldown limits is to ensure that given transient events will not negatively affect the pressurizer structural integrity beyond Code allowables. These limits will be maintained within ASME Code allowables in the

Attachment I to W3Fl -2004-0014 Page 4 of 4 TRM in accordance with 10 CFR 50.59. Therefore, the consequences of an accident are not increased.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The limitations imposed on the pressurizer heatup and cooldown rates are provided to assure that the pressurizer is operated within the design criteria assumed for the flaw evaluation and fatigue analysis performed in accordance with the ASME Code Section Xl, subsection IWB-3600 requirements. The Waterford 3 FSAR has analyzed the conditions that would result from a thermal or pressurization transient on the Waterford 3 pressurizer. The proposed deletion of the pressurizer heatup and cooldown rates and relocation of the limits to the TRM does not change the way that the pressurizer is designed or operated.

Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

3. Does the proposed change involve a significant reduction in a margin of safety?

Response: No.

The margin of safety is established by the rules contained in the ASME Section III Code. Any future changes to the cooldown or heatup rates will be evaluated using 10 CFR 50.59 and are required to meet the ASME Code margins.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, Entergy concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of 'no significant hazards consideration" is justified.

5.3 Environmental Considerations The proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

Attachment 2 W3FI -2004-0014 Proposed Technical Specification Changes (mark-up) to W3Fl -2004-0014 Page 1 of 2 INDEX LIMITING CONDITION FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION STARTUP AND POWER OPERATION .3/4 4-1 HOT STANDBY .3/4 4-2 HOT SHUTDOWN .3/4 4-3 COLD SHUTDOWN - LOOPS FILLED .............................. 3/4 4-5 COLD SHUTDOWN - LOOPS NOT FILLED. 3/4 4-6 3/4.4.2 SAFETY VALVES SHUTDOWN..................................................................... 3/4 4-7 OPERATING..................................................................... 3/4 4-8 3/4.4.3 PRESSURIZER PRESSURIZER................................................................. 3/4 4-9 AUXILIARY SPRAY .............. ,........................... 3/4 4-9a 3/4.4.4 STEAM GENERATORS ............................................. 3/4 4-10 3/4.4.5 REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION SYSTEMS .................................... 3/4 4-17 OPERATIONAL LEAKAGE .............................................. 3/4 4-18 3/4.4.6 CHEMISTRY .............................................. 3/4 4-21 3/4.4.7 SPECIFIC ACTIVITY .............................................. 3/4 4-24 3/4.4.8 PRESSURE/TEMPERATURE LIMITS REACTOR COOLANT SYSTEM .3/4 4-28 PRESSURIZE ZE...... .................... 3/4 4-33 OVERPRESSURE PROTECTION SYSTEMS .3/4 4-34 3/4.4.9 DELETED ....... . . . . . . . . . . . .............. 3/4 4-36 3/4.4.10 REACTOR COOLANT SYSTEM VENTS ............................... 3/4 4-37 3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 314.5.1 SAFETY INJECTION TANKS .3/4 5-1 3/4.5.2 ECCS SUBSYSTEMS - Modes 1, 2, and 3 .3/4 5-3 3/4.5.3 ECCS SUBSYSTEMS - Modes 3 and 4 ................................ 3/4 5-8 3/4.5.4 REFUELING WATER STORAGE POOL.3/4 5-9 WATERFORD - UNIT 3 VI AMENDMENT NO. 22, 34, 18-8,

Attachment 2 to W3F1 -2004-0014 Page 2 of 2 REACTOR COOLANT SYSTEM l PRESSURIZER 4EATUP/COOLDOWN LIMITING CONDITION FOR OPERATION 3.4.8.2 The pressurizer shall be limited to:

a. A}ximum-heatuprate-of 200°F-pe-heudeletdr

b. A-maxmum-oeodlwn-rate of 200FpeF-hGuFdeIeted, and

c. A maximum spray nozzle usage factor of 0.65.

APPLICABILITY: At all times.

ACTION:

a. With-thepressuer4emperatufe~rits4iRexoess of any of-the above4imtsestorethetempeFature-within thelimits within 30 miutes%-perform aRengfReefing-evaluation-to-detemine the effeGtSefthe- out-fimit-oofitinon Whe stFuoGratergInrityevf the pressufizeFdeteFmine that-thpessuriz ernains-ac6eptable ffe-Gontinued-operation or ben-iat4ea st-HOT STAN OY-vithin 4he iext-6heurs-and-reduce the pFessurize pressur4lessthan 500-pskwitin4he-following 30 housdeleted.

b. With the spray nozzle usage factor > 0.65, comply with requirements of Table 5.7-1.

SURVEILLANCE REQUIREMENTS 4.4.8.2.1 Theepressurzer-tempeatures shal"e-deternmned-o-be-within4he Ibmitsat-least-Gnee-perF-30-mRutes-duriRg-systemheatup-er-Goeldewndeleted.

4.4.8.2.2 The spray water temperature differential shall be determined to be within the limit at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during auxiliary spray operation.

4.4.8.2.3 Each spray cycle and the corresponding AT (water temperature differential) shall be recorded whenever main spray is initiated with a AT (water temperature differential) of > 130'F and whenever auxiliary spray is initiated with a AT (water temperature differential) of > 140'F.

WATERFORD - UNIT 3 3/4 4-33 Amendment No.

Attachment 3 W3FI -2004-0014 Changes to Technical Specification Bases Pages For Information Only to W3Fl -2004-0014 Page 1 of 1 REACTOR COOLANT SYSTEM BASES PRESSURE/TEMPERATURE LIMITS (Continued)

The maximum RTNDT for all Reactor Coolant System pressure-retaining materials, with the exception of the reactor pressure vessel, has been determined to be 90 0F. The Lowest Service Temperature limit line shown on Figures 3.4-2 and 3.4-3 is based upon this RTNDT since Article NB-2332 of Section III of the ASME Boiler and Pressure Vessel Code requires the Lowest Service Temperature to be RTNDT + 100F for piping, pumps, and valves. Below this temperature, the system pressure must be limited to a maximum of 20% of the system's hydrostatic test pressure of 3125 psia (as corrected for elevation and instrument error).

The limitations imposed on the pressurizer heatup-an4GoGdewR4-ates-and spray water temperature differential are provided to assure that the pressurizer is operated within the design criteria assumed for the fatigue analysis performed in accordance with the ASME Code requirements.

The OPERABILITY of the shutdown cooling system relief valve or an RCS vent opening of greater than 5.6 square inches ensures that the RCS will be protected from pressure transients which could exceed the limits of Appendix G to 10 CFR Part 50 when one or more of the RCS cold legs are less than or equal to 2721F. Each shutdown cooling system relief Valve has adequate relieving capability to protect the RCS from overpressurization when the transient is either (1) the start of an idle RCP with the secondary water temperature of the steam generator less than or equal to 100F above the RCS cold leg temperatures or (2) inadvertent safety injection actuation with injection into a water-solid RCS. The limiting transient includes simultaneous, inadvertent operation of three HPSI pumps, three charging pumps, and all pressurizer backup heaters in operation. Since SIAS starts only two HPSI pumps, a 20%

margin is realized.

The restrictions on starting a reactor coolant pump in MODE 4 and with the reactor coolant loops filled in MODE 5, with one or more RCS cold legs less than or equal to 2721F, are provided in Specification 3.4.1.3 and 3.4.1.4 to prevent RCS pressure transients caused by energy additions from the secondary system which could exceed the limits of Appendix G to 10 CFR Part 50. The RCS will be protected against overpressure transients and will not exceed the limits of Appendix G by restricting starting of the RCPs to when the secondary water temperature of each steam generator is less than 100IF above each of the RCS cold leg temperatures. Maintaining the steam generator less than 100F above each of the Reactor Coolant System cold leg temperatures (even with the RCS filled solid) or maintaining a large surge volume in the pressurizer ensures that this transient is less severe than the limiting transient considered above.

WATERFORD - UNIT 3 B 3/4 4-1 0 Amendment No. 2, 406,

Attachment 4 W3FI -2004-0014 List of Regulatory Commitments to W3Fl -2004-0014 Page 1 of 1 List of Regulatory Commitments The following table identifies those actions committed to by Entergy in this document. Any other statements in this submittal are provided for information purposes and are not considered to be regulatory commitments.

TYPE TYPeck one) SCHEDULED ONE COMPLETION COMMITMENT TIME CONTINUING DATE (If ACTION COMPLIANCE Required)

Technical Specification (TS) 3.4.8.2, X Upon Pressurizer, item a., the maximum heatup rate, Implementation and b., the maximum cooldown rate, will be deleted and moved to the Technical Requirements Manual (TRM).

TS 3.4.8.2 Action a. will be deleted and moved to the TRM.

TS Surveillance Requirement (SR) 4.4.8.2.1 will be deleted and moved to the TRM.