W3F1-2004-0052, Supplement to Amendment Request NPF-38-249, Extended Power Uprate

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Supplement to Amendment Request NPF-38-249, Extended Power Uprate
ML042010150
Person / Time
Site: Waterford Entergy icon.png
Issue date: 07/14/2004
From: Venable J
Entergy Nuclear South, Entergy Operations
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
NPF-38-249, W3F1-2004-0052
Download: ML042010150 (179)
v • d • e


Text

Entergy Nuclear South Entergy Operations, Inc.

17265 River Road Entergy Killona, LA 70066 Tel 504 739 6660 Fax 504 739 6678 jvenabl~entergy.com Joseph E. Venable Vice President. Operations Waterford 3 W3F1 -2004-0052 July 14, 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" (W3Fl-2003-0074)
2. Entergy Letter dated March 4, 2004, "Supplement to Amendment Request NPF-38-249 Extended Power Uprate" (W3F1-2004-0017)
3. Entergy Letter dated May 7, 2004, "Supplement to Amendment Request NPF-38-249 Extended Power Uprate" (W3Fl-2004-0035)
4. Entergy Letter dated May 13, 2004, "Supplement to Amendment Request NPF-38-249 Extended Power Uprate" (W3Fl-2004-0042)
5. Entergy Letter dated May 21, 2004, "Supplement to Amendment Request NPF-38-249 Extended Power Uprate" (W3Fl-2004-0043)

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.

Since Reference 1 was submitted for NRC staff review, Entergy has responded to several requests for additional information (RAI) from the NRC staff. In three of these RAI responses (References 2, 3, and 5) Entergy committed to make changes to the requested Technical Specification changes or provide the staff with additional information, as necessary, based on the results of ongoing analysis. Additionally, in Reference 3, Entergy identified an issue where the extended power uprate (EPU) small break loss of coolant accident (SBLOCA) analysis did not support the Technical Specification (TS) changes as proposed in Reference 1. This supplement addresses these items. Specifically:

  • The reduction of the proposed primary-to-secondary leakage limit forTS 3.4.5.2c to 75 gallons per day (Reference 2). (The 75 gallon per day limit is being used in a separate alternative source term submittal.)

A Oo(

W3Fl-2004-0052 Page 2 of 5

  • Provides revised marked-up TS pages reflecting the withdrawal of the request to add the term 'indicated" and "an indicated" to various TS (Reference 3).
  • A revised Power Uprate Report (PUR) (i.e., Attachment 5 of Reference 1) Section 2.12.4, Small-Break LOCA (SBLOCA) Analysis" and updates to references listed in PUR Section 2.12.8, "References," is provided in Attachment 5 (Reference 3). (This submittal fulfills the 10 CFR 50.46(a)(3)(ii) requirement to notify the NRC staff of PCT changes of greater than 50'F from the last acceptable model. Therefore, no additional submittals will be made in accordance with 10 CFR 50.46(a)(3)(ii) to notify the NRC staff of this change to the SBLOCA analysis.)
  • The revision of the proposed new atmospheric dump valve TS necessary as a result of the SBLOCA reanalysis.

In Reference 3 and 5, Entergy, due to the emerging SBLOCA analysis issue, deferred its response to questions specific to the new digital ADV controllers that were proposed in Reference 1. As a result of the reanalysis of the post-EPU SBLOCA, Entergy has determined that digital ADV controllers will not be needed in support of the EPU. Therefore detailed responses to these questions will not be provided. All references / discussion / commitments regarding digital ADV controllers contained in Reference 1 are withdrawn. The currently installed analog ADV controllers will be utilized for EPU with setpoint indications available on the plant monitoring computer. Marked-up pages of Attachments 5 and 8 of Reference 1 reflecting the withdrawal of the use of digital ADV controllers are provided in Attachment 6.

Ongoing reviews have identified one EPU TS change and one non-EPU TS change that should have been included in Reference 1. Specifically:

  • A reference to Figure 3.1-1 in TS 3.1.2.2b should have been changed to Figure 3.1-2 to reflect the new figure being added in TS 3.1.2.8.
  • In Reference 1, it is proposed to relocate portions of Table 5.7-1, 'Component Cyclic or Transient Limits" to the Technical Requirements Manual (TRM) consistent with NUREG-1432, 'Standard Technical Specifications Combustion Engineering Plants."

To further be consistent with NUREG-1432 a "Component Cyclic or Transient Limit" program should have been added to TS as Section 6.5.5.

These changes are proposed in Attachment 1.

Additionally, reviews have identified that the lower limit for boric acid makeup tank solution temperature listed in Surveillance Requirements 4.1.2.1a, 4.1.2.2a, 4.1.2.7a, and 4.1.2.8b is non-conservative in regards to boron precipitation when instrument uncertainty is considered.

Changes to increase this lower limit and reduce redundant information are proposed in .

Reference 1 contained TS mark-ups along with information only TS Bases and Technical Requirements Manual (TRM) mark-ups. Ongoing reviews and analysis have identified minor typographical errors and the omission of a number of TS Bases and TRM changes that are necessary to support EPU. Additionally, a number of non-EPU changes have been approved and implemented impacting the marked-up pages. Revised mark-ups to reflect the corrections, resolve the omissions, and reflect the updated pages are provided in Attachments 2, 3, and 4. Not all of the marked-up pages in Reference 1 were impacted; however, a complete set of marked-up pages is provided in this letter. These marked-up pages supersede those previously provided in Attachments 2, 3, and 4 of Reference 1.

W3Fl-2004-0052 Page 3 of 5 In Reference 1, a commitment was made to complete the necessary Emergency Operating Procedure (EOP) changes within 60 days of the NRC staffs approval of the license amendment request (LAR) to eliminate the hydrogen recombiners. As stated in Reference 1, the elimination of the hydrogen recombiners is required in support of the EPU. The LAR to eliminate the hydrogen recombiners was approved by the staff as Amendment 192 on March 9, 2004 however; the necessary EOP changes to support EPU were not completed within 60 days. This issue was entered into the Waterford 3 corrective action program and discussed with the NRR Project Manager for Waterford 3. The commitment has been rescheduled to be completed prior to the beginning of the last operator training cycle before the EPU is implemented.

A commitment was made in Reference I to replace the moisture separator reheater (MSR) safety valves. Ongoing analysis indicates that MSR safety valve replacement is not necessary. Entergy will not be replacing the MSR safety valves for EPU and therefore withdraws this commitment.

Marked-up pages of Attachments 5 and 8 of Reference 1 reflecting changes discussed in this letter or previously reported to the NRC staff in References 3 and 4 are provided in .

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 no significant hazards consideration provided in Reference I requires minor corrections to address the revisions and additional changes proposed in this supplement. A revised no significant hazards consideration is provided at the end of Attachment 1 and supersedes the one previously provided in Reference 1. The additional changes proposed in this submittal do not change the environmental considerations previously submitted in Reference 1.

The proposed change does not include any new commitments.

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 July 14, 2004.

Sincerely, JEV/dbm

W3F1 -2004-0052 Page 4 of 5 Attachments:

1. Analysis of Proposed Technical Specification Change
2. Proposed Technical Specification Changes (mark-up)
3. Changes to Technical Specification Bases Pages (mark-up) - For Information Only
4. Changes to Technical Requirements Manual Pages (mark-up) - For Information Only
5. Revised Power Uprate Report Section 2.12.4, Small-Break LOCA Analysis
6. Miscellaneous Power Uprate Report Changes

W3Fl-2004-0052 Page 5 of 5 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 70066-0751 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 29h S. Main Street West Hartford, CT 06107-2445

Attachment 1 W3FI-2004-0052 Analysis of Proposed Technical Specification Change to W3FI -2004-0052 Page 1 of 21

1.0 DESCRIPTION

This letter supplements license amendment request NPF-38-249 to Operating License(s)

NPF-38 for Waterford Steam Electric Station, Unit 3 (Waterford 3) dated November 13, 2003 (Reference 1).

Since Reference 1 was submitted for NRC staff review requesting approval for an Extended Power Uprate (EPU), Entergy Operations, Inc. (Entergy) has responded to several requests for additional information (RAI) from the NRC staff. In three of these RAI responses (References 2, 3, and 4) Entergy committed to make changes to the proposed Technical Specification changes or provide the staff with additional information, as necessary, based on the results of ongoing analysis. Additionally, in Reference 3, Entergy identified an issue where the extended power uprate (EPU) small break loss of coolant accident (SBLOCA) analysis did not support the atmospheric dump valve (ADV) Technical Specification (TS) as proposed in Reference 1.

Other miscellaneous changes, both technical and editorial, have been identified since Reference 1 was submitted to the NRC staff for review. These miscellaneous changes are included in this supplement as described below. Changes to the TS Bases markups and Technical Requirements Manual markups are included for information only. No technical justification is provided for these information only changes.

Not all of the marked-up pages submitted with Reference 1 are impacted; however a complete set of marked-up pages is provided with this supplement. These marked-up pages supersede those previously provided in Attachments 2, 3, and 4 of Reference 1.

2.0 PROPOSED CHANGE

S New changes being proposed in this supplement for which technical justification is provided in Section 4.0, 'Technical Analysis," are identified in Section 2.1. Changes other than those listed in Section 2.1 that are identified in Section 2.2 are considered to be editorial /

administrative in nature or are the result of previous NRC interactions and correspondence (e.g., withdrawal of the use of "indicated" or "an indicated") and no further technical justification is provided.

2.1 Proposed TS Changes Different From Those Previously Reauested In Reference 1:

  • Technical Specification Index page VI, delete "PRESSURIZER ......... 3/4 4-33"
  • Technical Specification 3/4.1.2.1, Boration Systems Flow Paths - Shutdown:

Delete SR 4.1.2.la, "At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when the Reactor Auxiliary Building air temperature is less than 55 *F by verifying the Boric Acid Makeup Tank solution is greater than 55 'F (when the flow path from the boric acid makeup tank is used)."

Combine SR 4.1.2.1 b with first line of SR 4.1.2.1 so that it reads, "At least one of the above required flow paths shall be demonstrated OPERABLE at least once per 31 days by verifying that each valve (manual, power-operated, or to W3F1-2004-0052 Page 2 of 21 automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position."

  • Technical Specification 3/4.1.2.2, Boration Systems Flow Paths - Operating:

Change reference in 3.1.2.2b from "Figure 3.1-1" to "Figure 3.1-2."

Delete SR 4.1.2.2a, "At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when the Reactor Auxiliary Building temperature is below 550 F by verifying that the temperature of the boric acid makeup tank(s) is above 550 F."

Re-label SRs b, c, and d as a, b, and c respectively.

  • Technical Specification 3/4.1.2.7, Borated Water Sources - Shutdown:

Increase minimum temperature limit from 550 F to 600 F such that SR 4.1.2.7a states, "At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when the Reactor Auxiliary Building air temperature is less than 550 F by verifying the boric acid makeup tank solution is greater than 600 F (when it is the source of borated water).

  • Technical Specification 3/4.1.2.8, Borated Water Sources - Operating:

Add reference to TS 3.5.4 to 3.1.2.8b and delete refueling water storage pool (RWSP) information that is duplicated in TS 3.5.4. Following the change, TS 3.1.2.8b will read, "The refueling water storage pool in accordance with Specification 3.5.4."

  • Technical Specification 3/4.1.2.8, Borated Water Sources - Operating:

Delete SR 4.1.2.8a, "At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by verifying the RWSP temperature when the Reactor Auxiliary Building air temperature is less than 550 F or greater than 1000F."

Increase minimum temperature limit from 550 F to 600F such that SR 4.1.2.8b (new SR 4.1.2.8a) states, "At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by verifying the boric acid makeup tank solution temperature is above 600 F when the Reactor Auxiliary Building air temperature is less than 55 0F."

Re-label SRs b and c, as a and b respectively.

  • Technical Specification 3/4.4.5.2, Operational Leakage:

Revise 3.4.5.2c from:

"1 gpm total primary-to-secondary leakage through all steam generators and 720 gallons per day through any one steam generator,"

to:

"75 gallons per day primary-to-secondary leakage per steam generator,"

  • Technical Specification 3/4.5.4, Refueling Water Storage Pool Revise 3.5.4c from:

"A solution temperature of between 55 'F and 100 'F."

to:

"A solution temperature of greater than or equal to 55 'F and less than or equal to 100 'F."

to W3F1 -2004-0052 Page 3 of 21

  • Technical Specification 3/4.7.1.3, Condensate Storage Pool Revise newly proposed 3.7.1.3b from:

"A water temperature of between 55 'F and 100 'F."

to:

"A water temperature of greater than or equal to 55 'F and less than or equal to 100 'F."

  • Technical Specification 3/4.7.1.7, Atmospheric Dump Valves:

Add a new SR 4.7.1.7b that reads, "By verifying each ADV automatic actuation channel is in automatic with a setpoint of less than or equal to 1040 psia at least once per 92 days when the automatic actuation channels are required to be OPERABLE."

Re-label previously proposed SRs b, c, and d as c, d, and e respectively.

Change previously proposed setpoint in footnote from 1000 psia to 1040 psia.

Add, "The following programs shall be established, implemented, and maintained."

As the first line in Section 6.5.

Revise "6.5.1 through 6.5.6 will be used later." to "6.5.1 through 6.5.4 will be used later."

Add new Section 6.5.5, "Component Cyclic or Transient Limit" that reads, 'This program provides controls to track Technical Requirements Manual Section 5.7 cyclic and transient occurrences to ensure that components are maintained within the design limits."

Add '6.5.6 will be used later."

2.2 Tabulation of Differences in Marked-up TS Pages from Those Submitted in Reference 1 A brief discussion of the changes proposed to the marked-up TS pages different from those previously submitted in Reference 1 is provided in tabular form below.

TS Page # in TS Page # in Aft. 2 of the Aft, 2 of this Discussion Reference I submittal OL-4 OL-4 No changes made.

IV IV No changes made.

NA VI New page:

  • With issuance of Amendment 195 and approval of the proposed change to delete the spray nozzle usage factors and related surveillance requirements, TS 3/4.4.8.2 will be deleted.

to W3F1 -2004-0052 Page 4 of 21 TS Page # in TS Page # in Aft. 2 of the Aft, 2 of this Discussion Reference 1 submittal Vill Vil No changes made.

XV XV No changes made.

XIX XIX

  • Corrected proposed title for Figure 3.1-1 to match title on proposed Figure 3.1-1.
  • Corrected proposed page number for new Figure 3.1-2 from "3/4 14a" to "3/4 1-14a."
  • Marked on updated page to reflect the issuance of Amendment 196.

XXII XXII No changes made.

XXIII XXIII No changes made.

1-3 1-3 No changes made.

1-6 1-6 No changes made.

2-3 2-3 No changes made.

NA 3/4 1-6 New page:

NA 3/4 1-7 New page:

  • Correct figure reference in 3.1.2.2b from "3.1-1" to "3.1-2."

3/4 1-12 3/4 1-12

  • Removed 'indicated" three places and removed "an indicated" one place. (Reference 3)

3/4 1-14 (13) 3/4 1-14 (13)

  • Removed "indicated" five places. (Reference 3)
  • Added reference to TS 3.5.4, "Refueling Water Storage Pool" and deleted information duplicated in TS 3.5.4.
  • Revise "BAMT temperature" to "boric acid makeup tank solution temperature."

3/4 1-13 (14) 3/4 1-13 (14) No changes made.

Insert Fig Insert Fig Removed "Indicated" one place. (Reference 3) 3.1-1 3.1-1 Insert Fig Insert Fig Removed "Indicated" one place. (Reference 3) 3.1-2 3.1-2 3/4 2-11 3/4 2-11 Removed "indicated" two places. (Reference 3) 3/4 2-13 3/4 2-13 Removed "indicated" one place. (Reference 3) 3/4 3-19 3/4 3-19 No changes made.

3/4 3-20 3/4 3-20 No changes made.

to W3F11-2004-0052 Page 5of21 TS Page # in TS Page # in Af. 2 of the Aft, 2 of this Discussion Reference I submittal 3/4 4-18 3/4 4-18

  • Revised TS 3.4.5.2c to "75 gallons per day primary-to-secondary leakage per steam generator," superseding previous request for 0.75 gpm and 540 gpd limits.
  • Note: A separate (non-EPU related) license amendment request was submitted for NRC staff review on May 7, 2004 that impacts other aspects on this page.

3/4 4-33 3/4 4-33

  • Marked on updated page to reflect the issuance of Amendment 195 which relocated the pressurizer heatup and cooldown limits to the Technical Requirements Manual (TRM).
  • With the issuance of Amendment 195 and the approval of the proposed change, TS 3/4.4.8.2 will be deleted.

3/4 5-1 3/4 5-1 Removed "indicated" 15 places. (Reference 3) 3/4 5-9 3/4 5-9 Removed "indicated" three places. (Reference 3)

Revised '...between 55 'F and 100 'F." in 3.5.4c to "... greater

._ than or equal to 55 'F and less than or equal to 100 *F."

3/4 6-13 3/4 6-13 Removed "indicated" one place. (Reference 3) 3/4 7-1 3/4 7-1 No changes made.

Marked on updated page to reflect the issuance of Amendment 189 which replaced the reference to TS 4.0.5 with a reference to the Inservice Test Program.

3/4 7-2 3/4 7-2 No changes made.

3/4 7-3 3/4 7-3 Modified title of Table 3.7-2 to be consistent with proposed action statement and proposed index page change by deleting "During Operation with Both Steam Generators" from the end of the title.

Insert Table Insert Table Removed "indicated" one place. (Reference 3) 3.7-2 3.7-2 3/4 7-6 3/4 7-6 Removed Vindicated" two places. (Reference 3)

Revised proposed 3.7.1.3b from "...between 55 'F and 100

'F." to "... greater than or equal to 55 'F and less than or equal to 100 'F."

3/4 7-9 3/4 7-9 No changes made.

3/4 7-9a 3/4 7-9a No changes made.

3/4 7-9b 3/4 7-9b

  • Removed "indicated" two places. (Reference 3)
  • Proposed new surveillance requirement 4.7.1.7b to verify controller is in automatic and setpoint set at less than or equal to 1040 psia and made associated editorial changes.
  • Changed previously proposed automatic actuation setpoint from 1000 psia to 1040 psia in footnote.

5-1 5-1 No changes made.

to W3Fl-2004-0052 Page 6 of 21 TS Page # in TS Page # in Aft. 2 of the Aft, 2 of this Discussion Reference 1 submittal 5-5 NA The change to TS 5.3.1 proposed in Reference I is withdrawn.

5-6 5-6 No changes made.

5-7 5-7 No changes made.

5-8 5-8 No changes made.

5-9 5-9 No changes made.

NA 6-7 New page:

  • Adds new Section 6.5.5, Component Cyclic or Transient Limit" program.

6-20a 6-20a No changes made.

6-20a Insert 6-20a Insert Previously proposed document designator"CENPD-282-P-A" was changed to "WCAP-15996-P-A."

2.3 Tabulation of Differences in Marked-up TS Bases Pages from Those Submitted in Reference 1 (Information Only)

A brief discussion of the changes made to the marked-up TS Bases pages from those previously submitted in Reference 1 is provided in tabular form below. (For information only)

Bases Page Bases Page # in Aft. 3 of

  1. in Aft. 3 of this Discussion Reference 1 submittal NA B2-3 New page:
  • Remove reference to CEA ejection NA B2-4 New page:
  • Revise SG full load operating point from -900 psia to -810 psia NA B2-6 New page:
  • Add Bases Insert 2.2.1 NA Bases 2.2.1 New page:

Insert

  • Bases 2.2.1 Insert B3/4 0-7 B3/4 0-7 No changes made.

Bases 3/4.0- Bases 3/4.0- No changes made.

1 Insert 1 Insert B3/4 0-## B3/4 0-## Page "1" (Bases Table (Bases Table

  • Revised 4.1.2.7 to increase minimum BAMT solution 3/4.0-1) 3/4.0-1) temperature to 60'F and changed analytical limit to 49-F.
  • Deleted 3.1.2.8 RWSP information.

to W3FI-2004-0052 Page 7 of 21 Bases Page Bases Page # in At. 3 of

  1. in Aft. 3 of this Discussion Reference I submittal B3/4 0-## B3/4 0-## Page "2" (Bases Table (Bases Table
  • Deleted 3.1.2.8 and 4.1.2.8 RWSP information.

3/4.0-1) 3/4.0-1)

  • Revised 4.1.2.8 to increase minimum indicated BAMT solution temperature to 60'F and changed analytical limit to 49-F. (Information now on Page "1.")
  • Revised 3.4.5.2c to 75 gpd B3/4 0-## B3/4 0-## Page "3" (Bases Table (Bases Table
  • Revised units for TS 3.4.8.2 Pressurizer Heatup and 3/4.0-1) 3/4.0-1) Cooldown rates from "'F" to "F/hr" two places.

(Information now on Page "2.")

  • Add comment to 3.5.4 and 3.7.1.7 Indicated on PMC."
  • Revised TS 3.7.1.7 ADV setpoints from "1000 psia" and "985 psia" to "1040 psia" and "992 psig."

B3/4 1-2 B3/4 1-2 No changes made.

B3/4 1-3 83/4 1-3 Deleted redundant discussion regarding reactor water storage pool information.

Bases Bases Added discussion regarding the BAMT low temperature limit.

3/4.1.2 Insert 3/4.1.2 Insert 63/4 2-4 B3/4 2-4 Deleted last sentence of existing paragraph in Bases 3/4.2.8.

Bases Bases Revised "553 'F and "556 'F" to "533 'F" and "536 F" 3/4.2.6 Insert 3/4.2.6 Insert respectively. (Reference 3)

Bases Bases No changes made.

3/4.2.8 Insert 3/4.2.8 Insert 83/4 4-3 13/4 4-3 Revised pri-to-sec leakage value to 150 gallons per day (gpd) in regards to tube integrity and to 75 gpd regarding TS 3.4.5.2. (Reference 2) 83/4 4-4 83/4 4-4 Revised pri-to-sec leakage discussion (Reference 2)

NA Bases New page:

3/4.4.5.2

  • Revised pri-to-sec leakage discussion to match new Insert leakage limit of 75 gpd. (Reference 2)

B3/4 4-10 83/4 4-10 No changes made.

13/4 5-1 b NA No longer necessary.

Bases 3.5.2 NA No longer necessary.

Insert B3/4 5-3 83/4 5-3 No changes made.

83/4 6-2 83/4 6-2 No changes made.

Bases Bases No changes made.

3/4.6.1.5 3/4.6.1.5 Insert Insert 83/4 7-1 83/4 7-1 No changes made.

Bases Bases No changes made.

3/4.7.1.1 3/4.7.1.1 Insert #1 Insert #1 to W3F1 -2004-0052 Page 8 of 21 Bases Page Bases Page # in Aft. 3 of

  1. in Aft. 3 of this Discussion Reference I submittal Bases Bases No changes made.

3/4.7.1.1 3/4.7.1.1 Insert #2 Insert #2 B3/4 7-2e B3/4 7-2e No changes made.

63/4 7-2f B3/4 7-2f No changes made.

B3/4 7-3 B3/4 7-3

  • Revised primary-to-secondary leakage from "0.75 gpm" to "540 gallons per day."

B3/4 7-3b NA Proposed MFIV changes previously marked on this page are now shown on page 83/4 7-3e due to the implementation of Bases Change #31 which expanded the Bases for 3/4.7.1.5,

'Main Steam Line Isolation Valve (MSIV)."

NA 83/4 7-3c New page:

  • Proposed MSIV and MFIV changes previously marked on page 63/4 7-3 are now shown on this page due to the implementation of Bases Change #31 which expanded the Bases for 3/4.7.1.5, "Main Steam Line Isolation Valve (MSIV)."

NA 83/4 7-3e Proposed MFIV changes previously marked on page 83/4 7-3b are now shown on this page due to the implementation of Bases Change #31 which expanded the Bases for 3/4.7.1.5, "Main Steam Line Isolation Valve (MSIV)."

B3/4 7-## B3/4 7-## Page "1"

  • Revised last paragraph before "Limiting Condition for Operation" section to match the SBLOCA reanalysis.
  • Revised the setpoint from "1000 psia (985 psia indicated)"

to "1040 psia (992 psig indicated)" one place.

B3/4 7-## 63/4 7-## Page "2" In second paragraph under Actions, changed "Actions (a) and (b) would be entered only ..." to "Actions (a) and (b) would be applicable only ..."

to W3Fl-2004-0052 Page 9 of 21 Bases Page Bases Page # in Att. 3 of

  1. in Aft. 3 of this Discussion Reference I submittal B3/4 7-## 33/4 7-## Page "3"
  • Clarified manual capability in Action c.
  • Revised the setpoint from "1000 psia (985 psia indicated)"

to "1040 psia (992 psig indicated)" one place.

  • Added new description of surveillance requirement 4.7.1.7b to match new surveillance requirement proposed and re-labeled previously proposed SRs accordingly.

83/4 7-## 13/4 7-## Page "4"

  • Removed sentence from surveillance requirement b (new c) that allowed crediting valve stokes during a cooldown for inservice testing.
  • Revised the setpoint from "1000 psia (985 psia indicated)"

to "1040 psia (992 psig indicated)" one place.

B3/4 7-4 B3/4 7-4 No changes made.

Bases Bases No changes made.

3/4.7.4 Insert 3/4.7.4 Insert B3/4 7-4(1) 13/4 7-4(1) No changes made.

B3/4 8-1 83/4 8-1 No changes made.

Bases 3.8.1 Bases 3.8.1 No changes made.

Insert Insert 2.4 Tabulation of Differences in Marked-up TRM Pages from Those Submitted in Reference 1 (Information Only)

A brief discussion of the changes made to the marked-up TRM pages from those previously submitted in Reference 1 is provided in tabular form below. (For information only)

TRM Page#

TRM Page # in Att.4 of in Aft. 4 of this Discussion Reference I submittal 3/4 3-2 3/4 3-2 No changes made.

TRM Table TRM Table No changes made.

3.3-2 Insert 9 3.3-2 Insert 9 _

to W3Fl-2004-0052 Page 10of21 TRM Page#

TRM Page# in Att. 4 of in Aft. 4 of this Discussion Reference I submittal TRM Table TRM Table No changes made.

3.3-2 Insert 3.3-2 Insert 10 10 3/4 3-3 3/4 3-3 No changes made.

B 3/4 # B 3/4 # Page 1"

  • No changes made.

B 3/4 # B 3/4 # Page 2"

  • No changes made.

3/4 3-6 3/4 3-6 Mark on updated page to reflect recently implemented TRM Amendment 82 which revised item 4.c.

NA 3/4 3-40 New page:

  • Changes to Ultrasonic Flowmeter requirement NA B 3/4 2a New page:
  • Changes to Ultrasonic Flowmeter requirement NA B 3/4 2b New page:
  • Changes to Ultrasonic Flowmeter requirement New New No changes made.

Requirement Requirement 3/4.4.3 3/4.4.3 New Bases New Bases No changes made.

314.4.3 3/4.4.3 3/4 6-5a 3/4 6-5a No changes made.

B 3/4 3 B 3/4 3 No changes made.

3/4 7-1c 3/4 7-1c Marked on updated page to reflect recently implemented TRM Amendment 80 which updated this requirement.

B 3/4 3e B 3/4 3e No changes made.

B 3/4 3f B 3/4 3f Mark on updated page to reflect recently implemented TRM Amendment 80 which updated this requirement.

3/4 6-12 3/4 6-12 No changes made.

2.5

  • Summary Changes are proposed to Technical Specifications (TS) 3/4.1.2.1, 3/4.1.2.2, 3/4.1.2.7, and 3/4.1.2.8 that:
  • Delete redundant information and requirements relative to the boric acid makeup tanks and refueling water storage tank,
  • Provide the correct reference to Figure 3.1-2, and
  • Increase the minimum required boric acid makeup tank solution temperature to 60'F.

A change is proposed to TS 3/4.4.5.2 to limit primary-to-secondary operational leakage in each steam generator to 75 gallons per day.

to W3Fl-2004-0052 Page 11 of 21 A revision is proposed to new TS 3/4.7.1.7 to increase the ADV setpoint to 1040 psia and to add new Surveillance Requirement (SR) 4.7.1.7b to verify that the ADV controller is in automatic with a setpoint of less than or equal to 1040 psia.

A change is proposed that adds the "Component Cyclic or Transient Limit" program to Section 6.5. "Programs."

A revision is proposed to change a Combustion Engineering document designator to the recently assigned Westinghouse designator.

3.0 BACKGROUND

In 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.

Since Reference 1 was submitted, Entergy has responded to several requests for additional information (RAI) from the NRC staff. In three of these RAI responses (References 2, 3, and

4) Entergy committed to make changes to the proposed Technical Specification changes or provide the staff with additional information, as necessary, based on the results of ongoing analysis. Additionally, in Reference 3, Entergy identified an issue where the extended power uprate (EPU) small break loss of coolant accident (SBLOCA) analysis did not support the Technical Specification (TS) changes as proposed in Reference 1. Specifically:
  • In Reference 2 Entergy committed to submit a change to reduce the primary-to-secondary leakage limit (TS 3.4.5.2c) to 150 gallons per day from that proposed in Reference 1. Due to unique equipment configurations and control room dose considerations, Entergy is proposing to reduce this limit to 75 gallons per day. (The 75 gallon per day limit is being used in separate alternative source term submittal.) This submittal satisfies the commitment made in Reference 2.
  • In Reference 3 Entergy withdrew its request to add the term 'indicated" or 'an indicated" to various TSs and committed to provided marked-up TS pages reflecting this withdrawal.

The TS pages reflecting this withdrawal are provided in Attachment 2. This submittal satisfies the commitment made in Reference 3.

  • In Reference 3 Entergy identified "... that the SBLOCA analysis performed in support of the EPU did not support that one high pressure safety injection (HPSI) pump and one

[atmospheric dump valve] ADV could mitigate the SBLOCA as Entergy had assumed in the November 13, 2003 submittal." Entergy has reanalyzed the EPU SBLOCA and the reanalysis now supports that one HPSI pump and one ADV can mitigate the post-EPU SBLOCA. The revision to Power Uprate Report Section 2.12.4, "Small-Break LOCA (SBLOCA) Analysis" reflecting the reanalysis of the SBLOCA is presented in Attachment 5.

  • In References 3 and 4 Entergy deferred its responses to questions specific to the new digital ADV controllers proposed in support of EPU. Since that time, based on the SBLOCA reanalysis, Entergy has determined that digital ADV controllers will not be needed in support of EPU. The current ADV controllers will be utilized. Therefore, to W3F1 -2004-0052 Page 12 of 21 responses to questions specific to digital ADV controllers are not required. This closes the commitment made in Reference 4.
  • Revisions to the new ADV TS proposed in Reference 1 are required based on the SBLOCA reanalysis. Specifically it is proposed that the analysis setpoint be increase from 1000 psia to 1040 psia and new surveillance requirement 4.7.1.7b be incorporated to verify that the ADV controllers are in automatic and set at the appropriate setpoint when required to be operable.

Ongoing Entergy reviews have identified one EPU TS change and one non-EPU TS change that should have been included in Reference 1. Specifically:

  • A reference to Figure 3.1-1 in TS 3.1.2.2b should have been changed to 3.1-2 to reflect the new figure being added in TS 3.1.2.8.
  • In Reference 1, it is proposed to relocate portions of Table 5.7-1, "Component Cyclic or Transient Limits" to the Technical Requirements Manual (TRM) consistent with NUREG-1432, "Standard Technical Specifications Combustion Engineering Plants."

To further be consistent with NUREG-1432 a "Component Cyclic or Transient Limit" program should have been added to TS as Section 6.5.5.

These changes are proposed in this supplement.

Additionally, reviews have identified that the lower limit for boric acid makeup tank solution temperature listed in Surveillance Requirements (SRs) 4.1.2.1a, 4.1.2.2a, 4.1.2.7a, and 4.1.2.8b is non-conservative in regards to boron precipitation when instrument uncertainty is considered. Changes to increase this lower limit and reduce redudant information are proposed.

4.0 TECHNICAL ANALYSIS

4.1 Index Page VI The proposed change to Index page VI to delete 'PRESSURIZER ............. 3/4 4-33" is an editorial change necessary due to recently issued Amendment 195. Amendment 195 relocated the pressurizer heatup and cooldown limits from TS 3/4.4.8.2 to the TRM. With the approval of the changes proposed to TS 3/4.4.8.2 in Reference 1, TS 3/4.4.8.2 will be completely deleted.

.4.2 Boric Acid Makeup Requirements (TS 3/4.1.2.1. 3/4.1.2.2. 3/4.1.2.7. and 3/4.1.2.8)

TS 3.1.2.2b references Figure 3.1-1 in reference to the combined contents of the boric acid makeup tanks (BAMTs). A new figure (Figure 3.1-2) was proposed in Reference I specifically addressing the combined contents of the BAMTs. The reference to Figure 3.1-1 in TS 3.1.2.2b therefore should be changed to reference new Figure 3.1-2 to match the changes proposed in Reference 1. This is an editorial change and therefore acceptable.

SR 4.1.2.1a requires that the BAMT solution be monitored once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to ensure it is above the minimum temperature limit whenever RAB air temperature is below a specified temperature. SR 4.1.2.1a is redundant to SR 4.1.2.7a which also requires that the BAMT to W3Fl-2004-0052 Page 13 of 21 solution temperature be monitored once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to ensure it is above the minimum temperature limit whenever RAB air temperature is below the specified temperature. TSs 3/4.1.2.1 and 3/4.1.2.7 are both applicable in Modes 5 and 6. Similarly, SR 4.1.2.2a, which requires the same BAMT solution temperature monitoring, is redundant to existing SR 4.1.2.8b (new SR 4.1.2.8a). TSs 3/4.1.2.2 and 3/4.1.2.8 are both applicable in Modes 1, 2, 3, and 4. Therefore, the change to delete SRs 4.1.2.1a and 4.1.2.2a is an editorial change and is acceptable because it does not result in the elimination or reduction of any BAMT solution temperature monitoring requirements.

The remaining SRs in 4.1.2.1 and 4.1.2.2 will be rearranged or renumbered to account for the deletion of SRs 4.1.21a and 4.1.2.2a. This is an editorial change and therefore acceptable.

TS 3.1.2.7a and current TS Figure 3.1-1, "Required Stored Boric Acid Volume as a Function of Concentration," as well as proposed Figures 3.1-1 and 3.1-2, allow a maximum BAMT boron concentration of 6,125 parts per million (ppm) or 3.5 weight percent (WT%). The minimum temperature of solubility for boric acid solution at this concentration is 50.2 'F. The instrument accuracy of the BAMT solution temperature instrumentation has been determined to be i 5.76 'F. Therefore, if the BAMT solution was at the maximum boron concentration allowed by TSs and the minimum allowed TS temperature of 55 'F, it is possible that the actual temperature of the BAMT solution could be as low as 49 'F (i.e., below the minimum temperature of solubility). Therefore, it is proposed to increase the minimum allowed BAMT solution temperature to 60 'F in SRs 4.1.2.7a and 4.1.2.8b (i.e., new 4.1.2.8a) to prevent the possible precipitation of boric acid in the BAMT and thus provide additional assurance of BAMT operability.

The two BAMTs are located in the Reactor Auxiliary Building (RAB); therefore, in the absence of BAMT heaters, the BAMT temperature would approximate RAB temperature. Each BAMT has two independent full capacity electrical strip heater banks and is insulated. The heaters normally cycle to maintain BAMT temperature between 100 'F and 110 'F. Control room annunciators exist to alert the operator if either of the two heaters per BAMT loses power or if BAMT solution temperature is outside of a 90-120 'F range.

Since the BAMTs are in the RAB, RAB air temperature is monitored as a prerequisite for monitoring BAMT solution temperature per SRs 4.1.2.7a and 4.1.2.8b (i.e., new 4.1.2.8a). If the RAB temperature is below 55 'F, then BAMT solution temperature would be verified to be above 60 'F. This is acceptable because the instrument accuracy of the RAB air temperature instruments is +2.4 'F, which means an indicated RAB temperature of 55 'F would result in an RAB temperature no lower than 52.6 'F which is greater than the 50.2 'F solubility temperature at the maximum allowed BAMT boron concentration. Therefore BAMT temperature monitoring would begin before the RAB air temperature reaches the solubility temperature of the BAMT.

TS 3/4.1.2.8 is applicable in Modes 1, 2, 3, and 4 and requires that the refueling water storage pool (RWSP) be operable. TS 3.1.2.8b specifies RWSP minimum water volume, minimum and maximum boron concentrations, and minimum and maximum solution temperature. The RWSP is specifically addressed by TS 3/4.5.4 which is also applicable in Modes 1, 2, 3, and 4. TS 3.5.4 specifies the identical RWSP parameters as are currently specified in TS 3.1.2.8b. Therefore, it is proposed that the duplicate RWSP parameters listed in TS 3.1.2.8b be deleted and TS 3.1.2.8b be revised to read, "The refueling water storage to W3Fl-2004-0052 Page 14 of 21 pool in accordance with Specification 3.5.4." Therefore, the change to revise TS 3.1.2.8b is an editorial change and is acceptable because it does not result in the elimination or reduction of any RWSP requirements.

It is proposed to revise "... BAMT temperature ..." in SR 4.1.2.8b (new 4.1.2.8a) to "... boric acid makeup tank solution temperature ..." BAMT is not previously defined in TS.

Additionally, the addition of the word "solution" clarifies that it is the solution temperature that is to be monitored and is consistent with the wording in SR 4.1.2.7a. This is considered to be an editorial clarification that does not change any requirements and is therefore acceptable.

4.3 Steam Generator Primary-to Secondary Operational Leakage Limit (TS 3/4.4.5.2)

In Reference 1, Entergy proposed to lower the primary-to-secondary leakage limit in TS 3.4.5.2c from:

"1 gpm total primary-to-secondary leakage through all steam generators and 720 gallons per day through any one steam generator,"

to:

"0.75 gpm total primary-to-secondary leakage through all steam generators and 540 gallons per day through any one steam generator,"

In Reference 2, Entergy agreed to lower the leakage through any one steam generator to 150 gallons per day to be consistent with NEI 97-06, "Steam Generator Program Guidelines." The 150 gallons per day limit is a defense in depth measure that serves to limit the potential for tube rupture and is acceptable to the NRC staff. Since Reference 2 was submitted, Entergy has determined that the steam generator primary-to-secondary operational leakage limit must be restricted to 75 gallons per day per steam generator to achieve acceptable control room dose results for the SBLOCA event due to the close proximity of one atmospheric dump valve to one control room air intake. (The 75 gallon per day limit is being used in separate alternative source term submittal.) Therefore Entergy proposes to revise TS 3.4.5.2c to read:

"75 gallons per day primary-to-secondary leakage per steam generator,"

This change is acceptable because it imposes a more restrictive steam generator primary-to-secondary operational leakage limit such that acceptable radiological control room doses are achieved. This operational leakage limit also bounds the 150 gallons per day per steam generator limit recommended in NEI 97-06 and thus will provide an acceptable defense in depth measure to limit the potential for tube rupture.

4.4 Refueling Water Storaae Pool (TS 3/4.5.4) and Condensate Storage Pool (TS 3/4.7.1.3)

TSs 3.5.4 and 3.7.1.3 contain requirements for acceptable refueling water storage pool (RWSP) and condensate storage pool (CSP) temperatures respectively. Currently existing TS 3.5.4c and TS 3.7.1.3b, as proposed in Reference 1, require that the temperature be maintained "between 55 'F and 100 'F." This phraseology is ambiguous in that it is unclear whether the requirement is inclusive or exclusive of 55 'F and 100 'F. An editorial clarification is proposed to use more precise wording (i.e., "greater than or equal to 55 'F and less than or equal to 100 'F") to eliminate the ambiguity and to be consistent with wording used in SRs to W3F11-2004-0052 Page 15of21 4.5.4b and proposed 4.7.1.3b. This is considered to be an editorial clarification and does not change the current requirement, therefore this change is acceptable.

4.5 Atmospheric Dump Valve (TS 3/4.7.1.7)

This is a new Specification that addresses atmospheric dump valve (ADV) operability for the following:

  • Plant cooldown to shutdown cooling entry conditions with offsite power unavailable.
  • Small break LOCA (SBLOCA) mitigation above 70% of rated (EPU) thermal power.
  • Containment isolation.

The ADVs were previously credited only for cooldown to shutdown cooling entry conditions (primary safety function) and containment isolation (secondary safety function). ADV operability for cooldown is presently addressed in TRM requirement 3/4.7.1.7 while ADV operability for containment isolation is presently addressed in Technical Specification 3/4.6.3.

There is currently no Technical Specification specifically addressing overall ADV operability.

For EPU, the ADVs are credited for SBLOCA mitigation at greater than 70% rated (EPU) thermal power. Therefore, a new Technical Specification ensuring ADV operability for this purpose is proposed. Since a new ADV Specification is being added, it was deemed appropriate to address ADV operability for cooldown and containment isolation in the new Specification as well. Consequently, Technical Specification 3/4.7.1.7 is being created to address the cooldown, SBLOCA mitigation, and containment isolation functions of the ADVs.

One ADV in conjunction with one high-pressure safety injection pump is credited for SBLOCA mitigation as discussed in revised Section 2.12.4 of the PUR (Attachment 5 of this submittal).

For SBLOCA mitigation the ADVs must be in automatic with a setpoint of less than or equal to 1040 psia when operating at greater than 70% rated EPU thermal power. Therefore, the limiting condition for operation has been modified by a footnote indicating when the automatic actuation function of the ADVs is not required to be operable (i.e., when less than or equal to 70% power for greater than 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />). This is consistent with the SBLOCA safety analysis assumptions.

Actions (a) and (b) specifically address the inoperability of the automatic actuation capability of one and both ADVs, respectively. A 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Allowed Outage Time (AOT) is provided for when the automatic actuation capability of one ADV is inoperable. This AOT is consistent with the AOT for the inoperability of one high-pressure safety injection pump and is therefore considered appropriate. If the ADV automatic actuation capability is not restored to operable within the AOT, then the action to reduce power to less than or equal to 70% power results in exiting the applicability of the specification in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> (six hours to get to 70% power plus six more hours for decay heat to reduce) which is also consistent with the time allowed to exit the mode of applicability for an inoperable high-pressure safety injection pump. When the automatic actuation capability of both ADVs is inoperable, Action (b) requires that automatic actuation capability of one ADV be restored to operable within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or power be reduced to less than 70% power within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. Once again, the specification will no longer be applicable 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after reaching 70% power for the automatic actuation capability. These action times are consistent with the action times that would be required if both high-pressure to W3Fl-2004-0052 Page 16 of 21 safety injection pumps were found to be inoperable. Therefore, the proposed action times are considered to be acceptable.

Action (c) addresses the manual capabilities of the ADVs required for the cooldown and containment isolation functions. A 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> AOT is provided to restore the ADV to operable and is consistent with the current licensing basis regarding the cooldown function contained in the TRM. For the containment isolation function, the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> AOT is a relaxation from the 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> AOT currently specified in Specification 3.6.3, Containment Isolation. This relaxation is considered acceptable based on the main steam system being a closed system inside containment and the capability to isolate the ADV with the block valve if needed. Due to the importance of the opening function of the ADVs, isolation of the ADV for extended periods of time is not appropriate therefore a shutdown is required if the ADV can not be restored to operable. The time allowed for exiting the mode of applicability is consistent with most other Section 3.7 Technical Specifications that are applicable in modes 1, 2, 3, and 4. Therefore the proposed times are considered to be acceptable.

Surveillance requirements and frequencies proposed for the ADV automatic actuation channels are consistent with those approved in Amendments 114 and 102 for South Texas Project Units 1 & 2 dated August 19, 1999 with the addition of a proposed requirement to verify that the controllers are in automatic and set at the appropriate setpoint at least once per 92 days. This added verification provides additional assurance that the controllers are set properly. The surveillance requirements and frequencies proposed for the ADV automatic actuation channels are considered to be appropriate. The surveillance requirement to cycle each ADV through a complete cycle is consistent with NUREG-1 432, Rev. 3 and the current licensing basis as stated in the TRM. This requirement provides assurance that the ADVs can be used for cooldown and can be closed when needed for containment isolation. These surveillance requirements and frequencies are therefore considered to be appropriate for the ADVs.

The existing TRM requirement for ADVs,3/4.7.1.7, is being deleted since it is now replaced by Technical Specification 3/4.7.1.7.

Results of the SBLOCA analysis that credit an ADV are presented in revised PUR Section 2.12.4 (Attachment 5 of this submittal) and demonstrate that acceptance criteria are met.

4.6 Component Cyclic or Transient Limit (TS 6.5.5)

This is a new TS to address an omission in Reference 1. In Reference 1 it was proposed to relocate a portion of Table 5.7-1, Component Cyclic or Transient Limits. With the relocation of this information, a Component Cyclic or Transient Limit program should have also been proposed for the Administrative section of the TS. New Section 6.5.5 is proposed and will state:

.This program provides controls to track Technical Requirements Manual Section 5.7 cyclic and transient occurrences to ensure that components are maintained within the design limits."

This program is currently implied in TS. With the relocation of portions of Table 5.7-1 to the TRM, the proposed program will clearly state the requirement to track cyclic and transient to W3F1-2004-0052 Page 17 of 21 occurrences and ensure that components are maintained within the design limits therefore this change is acceptable. The proposed program is also consistent with NUREG-1 432, Standard Technical Specifications Combustion Engineering Plants.

4.7 Document Designator Change (TS 6.9.1.11.1)

Since its acquisition of Combustion Engineering, Westinghouse has assigned Westinghouse designators to Combustion Engineering documents. In this case, 'CENPD-282-P-A" is now known as "WCAP-1 5996-P-A." There were no changes to the document other than the assignment of the Westinghouse designator. This is an administrative change and is therefore acceptable.

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 TS, and do not affect conformance with any General Design Criterion (GDC) differently than described in the Updated Final Safety Analysis Report (UFSAR.)

5.2 No Significant Hazards Consideration The proposed Extended Power Uprate (EPU) will increase the maximum steady state core power for Waterford 3 to 3716 MWt, an approximate 8.0% increase above the currently licensed power level of 3441 MWt. The following operating license changes are proposed to support the EPU request.

  • Raise rated thermal power from 3441 MWt to 3716 MWt.
  • Revise the definition for Dose Equivalent Iodine to reference ICRP 30.
  • Lower the plant protection system and engineered safety features actuation system trip setpoints and allowable values for steam generator pressure - low.
  • Increase minimum required boron concentration and minimum volume limits for the boric acid makeup tanks.
  • Raise the minimum pressurizer pressure limit.
  • Lower the maximum safety injection tank volume requirement.
  • Add a new Technical Specification specific to the atmospheric dump valves.
  • Add references for two NRC approved analysis methodologies.

Attachment I to W3Fl-2004-0052 Page 18 of 21 In addition to the changes proposed for the EPU, a number of non-EPU related changes have been proposed. These include:

  • Deleting dual limits (in different units) in various Technical Specifications leaving a single limit in the units used by Operations.
  • Increase minimum allowed boric acid makeup tank solution temperature.
  • Deleting unnecessary pressurizer spray nozzle fatigue monitoring requirements and adding a Cyclic and Transient Limit program.
  • Incorporating the existing licensing basis lower containment air temperature limit into Technical Specifications.
  • Incorporating the existing licensing basis minimum and maximum temperature limits for the condensate storage pool into Technical Specifications.
  • Increasing the indicated minimum condensate storage pool level limit to more accurately address process measurement issues (analytical limit did not change).
  • Revising the main steam and feedwater isolation valve stroke times to achieve consistency and to incorporate instrument response time (analytical values did not change).
  • Deleting unnecessary information to improve consistency with NUREG-1432, Rev.2 (Reference 4).

Entergy Operations, Inc. has evaluated whether the proposed amendment involves a significant hazards consideration 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 impacts of the proposed EPU on plant Structures, Systems, and Components (SSCs) were reviewed with respect to SSC design capability, and it was determined that following completion of plant changes to support the EPU, no system, structure, or component would exceed its design conditions or limits. Evaluations supporting those conclusions were performed consistent with proposed Technical Specification changes. Consequently equipment reliability and structural integrity will not be adversely affected. Control system studies demonstrated that plant response to operational transients under EPU conditions does not significantly increase reactor trip frequency, so there will be no significant increase in the frequency of SSC challenges caused by reactor trip.

New systems are not needed to implement the EPU, and new interactions among SSCs are not created. The EPU does not create new failure modes for existing SSCs.

Modified components do not introduce new failure modes relative to those of the components in their pre-modified condition. Consequently, new initiators of previously analyzed accidents are not created.

The fission product barriers - fuel cladding, reactor coolant pressure boundary, and the containment building -- remain unchanged. The spectrum of previously analyzed postulated accidents and transients was evaluated, and effects on the fuel, the reactor to W3Fl-2004-0052 Page 19 of 21 coolant pressure boundary, and the containment were determined. These analyses were performed consistent with the proposed Technical Specification changes. The results demonstrate that existing reactor coolant pressure boundary and containment limits are met and that effects on the fuel are such that dose consequences meet existing criteria at EPU conditions.

The non-EPU related proposed changes do not affect reactor operations or accident analyses and have no radiological consequences.

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.

New systems are not required to implement the EPU, and new interactions among SSCs are not created. The EPU does not create new failure modes for existing SSCs.

Modified components do not introduce failures different from those of the components in their pre-modified condition. Consequently, no new or different accident sequences arise from SSC interactions or failures.

Training will be provided to address EPU effects, and the plant's simulator will be updated consistent with EPU conditions. Operating procedure changes are minor and do not result in any significant changes in operating philosophy. For these reasons, the EPU does not introduce human performance issues that could create new accidents or different accident sequences.

The increase in power level does not create new fission product release paths. The fission product barriers - fuel cladding, reactor coolant pressure boundary, and the containment building - remain unchanged.

The non-EPU related proposed changes introduce no new mode of plant operation.

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.

Structural evaluations performed at EPU conditions demonstrated that calculated loads on affected SSCs remain within their design allowables for all design basis event categories. ASME Code fatigue limits continue to be met.

Fuel performance evaluations were performed using parameter values appropriate for a reload core operating at EPU conditions. Those evaluations demonstrate that fuel performance acceptance criteria continue to be met. Reload evaluation processes to W3Fl-2004-0052 Page 20 of 21 ensure that fuel in the actual Cycle 14 reload core, the first to be operated at the increased power level, will meet regulatory criteria.

LOCA and non-LOCA safety analyses were performed under EPU conditions.

Emergency core cooling system performance was shown to meet the criteria of 10CFR50.46. The non-LOCA events identified in Waterford 3 FSAR Chapter 15 were shown to meet existing acceptance criteria. The LOCA and non-LOCA analyses were performed consistent with the proposed Technical Specification changes.

The containment building response to mass and energy releases was evaluated under EPU conditions. The evaluations showed that temperature and pressure limits were met.

No plant changes associated with the EPU reduce the degree of component or system redundancy.

Except for the deletion of pressurizer spray nozzle fatigue monitoring, existing Technical Specification operability and surveillance requirements are not reduced by the non-EPU related proposed changes, thus no margins of safety are reduced. A more realistic assessment of pressurizer spray nozzle fatigue has shown that nozzle fatigue will not be as significant over the plant design life as had been previously concluded, thus no margins of safety are reduced.

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 environmental considerations evaluation, Extended Power Uprate Report (PUR), Section 5, Environmental Considerations, previously submitted in Attachment 5 of Reference 1 is unaffected by the changes proposed in this supplement. It concludes that EPU will not result in a significant change in nonradiological impacts on land use, water use, waste discharges, terrestrial and aquatic biota, transmission facilities, or social and economic factors, and will have no nonradiological environmental impacts other than those evaluated in the Final Environmental Statement. It further concludes that EPU will not introduce any new radiological release pathways, will not result in a significant increase in occupational or public radiation exposures, and will not result in significant additional fuel cycle environmental impacts.

Therefore, the proposed amendment does not involve a significant change in the types or significant increase in the amounts of any effluent that may be released offsite nor does it involve a significant increase in individual or cumulative occupational radiation exposure.

to W3Fl-2004-0052 Page 21 of 21 6.0 PRECEDENCE None

7.0 REFERENCES

1. Entergy Letter dated November 13, 2003, "License Amendment Request NPF-38-249 Extended Power Uprate" (W3F1 -2003-0074)
2. Entergy Letter dated March 4, 2004, "Supplement to Amendment Request NPF 249 Extended Power Uprate" (W3F1 -2004-0017)
3. Entergy Letter dated May 7, 2004, "Supplement to Amendment Request NPF-38-249 Extended Power Uprate" (W3F1 -2004-0035)
4. Entergy Letter dated May 21, 2004, "Supplement to Amendment Request NPF-38-249 Extended Power Uprate" (W3FI-2004-0043)

Attachment 2 W3FI-2004-0052 Proposed Technical Specification Changes (mark-up)

or indirectly any control over (i) the facility, (ii) power or energy produced by the facility, or (iii) the licensee of the facility. Further, any rights acquired under this authorization may be exercised only in compliance with and subject to the requirements and restrictions of this operating license, the Atomic Energy Act of 1954, as amended, and the NRC's regulations. For purposes of this condition, the limitations of 10 CFR 50.81, as now in effect and as they may be subsequently amended, are fully applicable to the equity Investors and any successors in Interest to the equity investors, as long as the license for the facility remains In effect.

(b) Entergy Louisiana, Inc. (or its designee) to notify the NRC in writing prior to any change in (i) the terms or conditions of any lease agreements executed as part of the above authorized financial transactions, (ii) any facility operating agreement invoving a licensee that is in effect now or will be In effect in the future, or (iii) the existing property Insurance coverages for the facility, that would materially alter the representations and conditions, set forth In the staffs Safety Evaluation enclosed to the NRC letter dated September 18,1989. In addition, Entergy Louisiana, Inc. or Its designee Is required to notify the NRC of any action by equity Investors or successors in Interest to Entergy Louisiana, Inc. that may have an effect on the operation of the facility.

C. This license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations and orders of the Commission now or hereafter In effect: and Issubject to the additional conditions specified or Incorporated below:

  • M U1.m PoL EOI is th to operate t facility at reactor core power levels not In excess of megawatts thermal (100% power) in accordance with the conditions specified herein.
2. Technical Soecifications and Environmental Protection Plan The Technical Specifications contained InAppendix A, as revised through Amendment No. 183, and the Environmental Protection Plan contained In Appendix B. are hereby incorporated in the license. EOI shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

AMENDMENT NO. 134, 169, 170,171, 13

NDEX LIMITING CONDITIONS FOR QpERATION AND SURVEILLANCE REQUIREMENTS SECTIQN PAGE 314.0 APPLICABILITY .................................................. 3/4 0-1 314.1 REACTIVITY CONTROL SYSTEMS 314.1.1 BORATION CONTROL SHUTDOWN MARGIN - ANY CEA WITHDRAWN ..... ....... 3/4 1-1 SHUTDOWN MARGIN - ALL CEAS FULLY INSERTED ........ 3/41-3 MODERATOR TEMPERATURE COEFFICIENT ............ .. 3/4 1-4 MINIMUM TEMPERATURE FOR CRITICALITY ..... ......... 3/41-5 3/4.1.2 BORATION SYSTEMS FLOW PATHS - SHUTDOWN ............ ................ 3/4 1-6 FLOW PATHS - OPERATING ............ ................ 3/41-7 CHARGING PUMPS - SHUTDOWN ...... ................. 3/4 1-8 CHARGING PUMPS - OPERATING ...... ................. 3/4 1-9 BORIC ACID MAKEUP PUMPS - SHUTDOWN ..... ........ 3141-10 BORIC ACID MAKEUP PUMPS - OPERATING ..... ........ 3/41-11 BORATED WATER SOURCES - SHUTDOWN .............. 314 1-12 BORATED WATER SOURCES - OPERATING ..... ......... 3/4 1- KQ BORON DILUTION .............. ...................... 3/41-15 3/4.1.3 MOVABLE CONTROL ASSEMBLIES CEA POSITION ...................................... '.3/4 1-18 POSITION INDICATOR CHANNELS - OPERATING ..........3/4 1-21 POSITION INDICATOR CHANNELS - SHUTDOWN ......... 3/4 1-22 CEA DROP TIME ..................................... 3/41-23 SHUTDOWN CEA INSERTION LIMIT ........ ............ 3/4 1-24 REGULATING AND GROUP P CEA INSERTION LIMITS ..... 3/41-25 WATERFORD - UNIT 3 IV AMENDMENT NO. t++,-39,462-2

LIIING CONDMON FOR OPEATON AND SURVEILLANCE REQUIRlEMENT 3/4.4 REACTOR COOLANT SYSTEM 3/4A.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION STARTUP AND POWER OPERATION . .31 44-HOT STANDBY . .3/4 4-2 HOT SHUrTDOWN .................................... SIA 4-3 3/.3.............

COLD SHUTDOWN - LOOPS FILLED . . 314 4-5 COLD SHUTDOWN - LOOPS NOT FILLED . .3/44-6 314.4.2 SAFETY VALVES SHUTDOWN ....... ..... 314 4-7 OPERATING. 314 48 314.4.3 PRESSURIZER PRESSURIZER ........................................ 3/44-9 AUXILIARY SPRAY ........................................ 3/4 4-9a 3/4.4.4 STEAM GENERATORS ........................................ 3/44-10 3/4.4.5 REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION SYSTEMS ....................... ,.......... 314 4-17 OPERAllONAL LEAKAGE .3/4 4-18 314.4.6 CHEMISTRY .......... 3/4 4-21 3/4.4.7 SPECIFIC ACTIVITY ........................................... 3/4 4-24 3/4.4.8 PRESSURE/EMPERATURE LIMITS RE&CTOR COOLANT SYSTEM .3/4 4-28 OVERPRESSURE PROTECTION SYSTEMS 3/4 4-34 3/4.4.9 DELETED ...... 3/4 4-36 314.4.10 REACTOR COOLANT SYSTEM VENTS ............................... 314 4-37 3/4.6 EMERGENCY CORE COOLING SY MS (ECCS) 3/4.5.1 SAFETY INJECTION TANKS ..... 314 5-1 3/4.5.2 ECCS SUBSYSTEMS - Modes 1 2, and 3 . .34 5.3 3/4.5.3 ECCS SUBSYSTEMS - Modes 3 and 4 . .314 5-B 3/4.5.4 REFUELING WATFR STORAGE POOL.3/4 5-9 WATERFORD - UNIT 3 VI AMENDMENT NO. 22.34. 1C8O 103.1

LMING CONDITINS FOROPERATION AND SURVPILLANCE REQUIREMEGE 3/4.7 PLANT SYSTEMS (Continued)

ACTIVIY ....................................... 314 7-7 MAIN STEAM UNE ISOLATION VALVES (MSIVs) ...... ... 314 7-9 MB EDA}sISOLATION VALE 3t4 7-9a 3/4.7.2 STEAM GENERATOR PRESSURErrEMPERATURE LIMITATION . ... 314 7-10 3/4.7.3 COMPONENT COOUNG WATER AND AUXILIARY COMPONENT COOLING WATER SYSTEMS 3/4 7-11 3/4.7.4 ULTIMATE HEAT SINK . ..... ........... 314 7-12 3/4.7.6 FLOOD PROTECTION ............. . 3/4 7-15 3/...............

314.7.6 CONTROL ROOM AIR CONDmONING SYSTEM ............. 3/4 7-16 CONTROL ROOM EMERGENCY AIR FILTRATION SYSTEM - OPERATING ................... 314 7-16 CONTROL ROOM EMERGENCY AIR FILTRATION SYSTEM - SHUTDOWN ... 3/4 7-18 CONTROL ROOM AIR TEMPERATURE - OPERATING 3/4 7-1 Ba CONTROL ROOM AIR TEMPERATURE - SHUTDOWN 3/4 7-1 Sb CONTROL ROOM ISOLATION AND-PRESSURIZATION 3/4 7-1 8c 3/4.7.7 CONTROLLED VENTILATION AREA SYSTEM................ 314 7-19 3/4.7.8 SNUBBERS . . . .... 3/4 7-21 3/4.7.9 SEALED SOURCE CONTAMINATION . ...... 314 7-27 3/4.7.12 ESSENTIAL SERVICES CHILLED WATER SYSTEM...._ 314 7-43 3/4.8 ELECTRICAL POER SYSTEMS 314.9.1 A.C. SOURCES OPERATING . . 314 8-1 SHUTDOWNU. 3/4 84 3/4.8.2 D.C. SOURCES OPERATING .. ................... 3/4 8-9 SHUTDOWN ...................... 314 8-12 WATERFORD UNIT 3 Vill AMENDMENT NO. 6G,tR4 6, -,

IND DESIGN FEATURES SECTION 51 SIE 5.1.1 EXCLUSION AREA.............................................................. 51-1 5.12 LOW POPULATION ZONE.................................................. 5-1 5.1.3 MAP DEFINING UNRESTRICTED AREAS FOR RADIOACTIVE GASEOUS AND LIQUID EFFLUENTS ............................ 5-1 5.3 REACTOR CORE 5.3.1 FUEL ASSEMBLIES............................................................. 5-5 5.3.2 CONTROL ELEMENT ASSEMBLIES.................................. 5-5 5.A NOTUSE 5 5 METFORI O(8IC.At TOWERS I OrcTION...................................... 5-5 5.6 FUEL STORAGE 5.6.1 CRITICALITY. 5-6

& -1 MnAlIA.

AI ........................................................................... 5-6 5.6.4 CAPACITY ............................................................................ 5-6 WATERFORD - UNIT 3 AMENDMENT NOB3&

LDN LIST Ol FIGiUmES FIBURE PhE 3.1-1 REQUIREDSTOREDBORICACIDVOLUMEASAFUNCTIONOF CONCENTRATIO_ 1....... 81 3

3, iS (VOLUM1E OFO0f 13A 34-1 DELETED .................. .......... 4-27

======'314 3.4-2 WATERFORD UNIT 3 HEATUP CURVE -32 EFPY REACTOR COOLANT SYSTEM PRESSURE - TEMPERATURE LIMITS.3/4 4-30 3.4-3 WATERFORD UNIT 3 COOLDOWN CURVE -32 EFPY REACTOR COOLANT SYSTEM PRESSURE -TEMPERATURE LIMITS .............. 344-31 3.6-1 DELETED .. I 3/4 6-12 4.7-1 SAMPLING PLAN FOR SNUBBER FUNCTIONAL TEST .......... 3/4 7-26 5.1-1 EXCLUSION AREA ... ............... 5-2 5.1-2 LOW POPULATION ZONE . .- 3 5.1-3 ST BOUNDARY FOR RA)OACT GASEOUS AND LIQUID EFFLUENTS .54 5.6-1 ALTERNATIVE CHECKERBOARD ARRANGEMENTS .- Sa 5.6-2 ACCEPTABLE BURNUP DOMAIN FOR UNRESTRICTED STORAGE OF SPENT FUEL INREGION 2.. .b 6.6-3 ACCEPTABLE BURNUP DOMWJN FOR SPENT FUEL IN CHIECKERBOARD ARRANGEMENT WITH FUEL OF 5% ENRICHMENT (OR LESS) AT 27 MWDflgU .................... 5-Sc 6.2-1 DELETED .6-3 6.2-2 DELETED ............. - .......... 6-4 xs-riEp 2o6Rrc Acro VOLUA4 AJ5 A FqA~c-rxo4' &F e-oArcA51T-R47.o'27 6cmewo-vo-iror i-w 14AT)..-- - II'c WATERFORD -UNIT3 XIX AMENDMENT NO.18,27.1 02,14. 18,-I96W

LIST OF TABLES (Continued)_

TABLE pog be AmyAG 3.7-1 STEAM LINE SAFETY V VES PER LOOP......................... 3/4 7-2 3.7-2 MAXIMUM ALLOWABLE NEAR POWER LEVEL- HIGH TRIP SETPOINT WITH INOPERABLE STEAM LINE SAFETY VALVES

IN 0 3/47-3 4.7-1 SECONDARY COOLANT SYSTEM SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM .................................... 3/4 7-8 3.7-3 ULTIMATE HEAT SINK MINIMUM FAN REQUIREMENTS PER TRAIN .......................................... 3/4 7-14 4.7-2 SNUBBER VISUAL INSPECTION INTERVAL ............................ 3/4 7-21a 4.8-2 BATTERY SURVEILLANCE REQUIREMENTS ......................... 3/4 8-11 WATERFORD - UNIT 3 AMENDMENT NO. 9, 23, 30, CO, 75,-"

INDEX LIST OF TABLES (Continued)

TABLE PAGE B3/4.4-1 REACTOR VESSEL FRACTURE TOUGHNESS ..................... B3/4 4-8 6.2-1 MINIMUM SHIFT CREW COMPOSITION ......................... 6-5 B33 ,O -I ALW4L'7r- l -l-4,D rMT* aF _ BD3Y ° WATERFORD - UNIT 3 XXIII

CORE ALTERATION 1.9 CORE ALTERATION shall be the movement or manipulation of any component within the reactor pressure vessel with the vessel head removed and fuel in the vessel. Suspension of CORE ALTERATION shall not preclude completion of movement of a component to a safe conservative position.

COLR - CORE OPERATING LIMITS REPORT 1.9a The CORE OPERATING LIMITS REPORT is the Waterford 3 specific document that provides core operating limits for the current operating reload cycle. These cyclo-specific core operating limits shall be determined for each reload cycle in accordance with Technical Specification 6.9.1.11. Plant operation within these operating limits is addressed in individual specifications.

DOSE FOUl VA! ENT f-ill 1.10 DOSE EQUIVALENT t-131 shall be that concentration of 1-131 (microccuries/garn) which alone would produce the same thyroid dose as the quantity and isotopic mixture of 1-131. 1-132, I-133. 1-134. and 1-135 actually pres T e dose cony rsin f ors used forth E-AVERAGE oiiNEmU oN ENEEM\

1.11 r shall betheeaverage (weightedIn proportion toAt c~oncentration of each radincici the reactor coolant at the time of sampling) of the sum of the average beta and gamm energies\

per disintegration (in MeV) for Isotopes. other than lodines, with half-lives greater than 15 minutes, making up at least 95% of the total noniodine activity in the coolant.

ENGINEERFD SAFETY FFATURES RFSPONSE TIME 1.12 The ENGINEERED SAFElY FEATURES RESPONSE TIME shall he that time Interval from when the monitored parameter exceeds its ESF actuation setpoint at the channel sensor until the ESF equipment is capable of performing Its safety function (i.e.. the valves travel to their required positions, pump discharge pressures reach their required values. etc.). Times shall Include diesel generator stasirig and s*quence loading delays where applicable. The response time may be measured by any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and methodology for verification have been previously reviewed and approved by the NRC.

FREQUIENCY NOTATION 1.13 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.1.

WATERFORD - UNIT 3 1-3 AMENDMENT NO. 402, 4 7 5/

tcgP-3e', S"/fOleooerl9 4e &rf I R 'es /19-217, 1bl/e s c- Jtce , 'r 1. -'e o4C anu 4 4i:vn .ra . f

DEFINITIONS RATED THERMAL POWER 1.24 RATEDHERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant REACTOR TRIP SYSTEM RESPONSE TIME 1.25 The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval rromn when the monitored parameter exceeds Its trip setpoint at the channel sensor until electrical power is Interrupted to the CEA drive mechanism. The response time may be measured by any series of sequential, overlapping, or total steps so that the entire response time Is measured. In lieu of measurement, response time may be verified for selected components provided that the components and methodology for verification have been previously reviewed and approved by the NRC.

REPORTABLE EVENT 1.26 A REPORTABLE EVENT shall be any of those conditions specified In Section 50.73 to 10 CFR Part 50.

SHIELD BUILDING INTEGRITY 1.27 SHIELD BUILDING INTEGRITY shall exist when:

a. Each door In each access opening is closed except when the access opening Is being used for normal transit entry and exit, then at least one door shalt be closed,
b. The shield building filtration system Is in compliance with the requirements of Specification 3.6.6.1, and
c. The sealing mechanism associated with each penetration (e.g., welds, bellows.

or O-rings) Is OPERABLE.

SHUTDOWN MARGIN 1.28 SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which the reactor Is subcritical or would be subcritical from Its present condItIon assuming all control element assemblies are fully inserted except for the single assembly of highest reactivity worth which Is assumed to be fully withdrawn.

WATERF-ORD - UNIT 3 1-6 AMENDMENT NO. 175,182,48a

TABLE' 2.-1 REACTOR PROTECTIVE INSTRUMENTATION T]RP SETPOINT LIMITS FUNCTIONAL UNIT TRIP SETPOINT ALLOWABIL VAUI F

1. Manual Reactor Trip Not Applicable Not Applicable
2. Linear Power Level - High Four Reactor Coolant Pum PS < 108% of RATED THERMAL POWER < 10876% of RATED THERMAL POWER Operating
3. Logarithmic Power Level - High (1) < 0.257% of RATED THERMAL POWER (6) < 0.280% of RATED THERMAL POWER (6)
4. Pressurizer Pressure - High < 2350 psia < 2359 psia
5. Pressurizer Pressure - Low >~1684 psia (2) Ž 1649.7 psla (2)
6. Containment Pressure - High < 17.1 pi S17.4 ps
7. Steam Generator Pressure - Low epsia (3) > sila(3)
8. Steam Generator Level - Low > 27.4% (4)
  • 26.48% (4)
9. Local Power Density - High < 21.0 kW/ft (5) S 21.0 kWlft (5)
10. DNBR- Low > 1.26 (5) > 1.26 (5)
11. Steam Generator Level - High < 87.7% (4) S 88.62% (4)
12. Reactor Protection System Logic Not Applicable Not Applicable
13. Reactor Trip Breakers Not Applicable Not Applicable
14. Core Protection Calculators Not Applicable Not Applicable
15. CEA Calculators Not Applicable Not Applicable
16. Reactor Coolant Flow - Low > 19.00 psid (7) > 18.47 psid (7)

WAEFORD - UNIT 3 2-3 mendmentN.4- 3

REACTIVITY CONTROL SYSTEMS 3t4.1.2 BORATION SYSTEMS FLOW PATHS - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.12.1 As a minimum, one of tne following boron Injection flow paths shall be OPERABLE and capable of being powered from an OPERABLE emergency power source:

a. A flow path from the boric acid makeup tank via either a boric acid makeup pump or a gravity feed connection and any charging pump to the Reactor Coolant System if the boric acid makeup tank in Specification 3.1.2.7a. is OPERABLE, or
b. The flow path from the refueling water storage pool via either a charging pump or a high pressure safety injection pump to the Reactor Coolant System if the refueling water storage pool in Specification 3.1.2.7b. is OPERABLE.

APPLICABILITY; MODES 5 and e.

ACTION:

With none of the above flow paths OPERABLE or capable of being powered from an OPERABLE emergency power source, suspend all operations Involving CORE ALTERATIONS or positive reactivity changes.*

SURVEILLANCE REQUIREMENTS 4.1.2.1 At least one of the above required flow paths shall be demonstrated OPERABLE_

/tmpratreisle o~hn 550Fb verifyinga^Boi Ac~ ke

/ Tang oltrgreater o t5°F (whhe flow pt r theS t east once per 31 days by verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, Is in its correct position.

' Plant temperature changes are allowed provided the temperature change is accounted for in the calculated SHUTDOWN MARGIN.

WATERFORD - UNIT 3 3/4 1-6 AMENDMENT NO. 4t, 0.-

REACTIVITY CONTROL SYSTEMS FLOW PATHS - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.2 At least two boron injection flow paths to the RCS via the charging pumps shall be OPERABLE. The following flow paths may be used:

a. With the contents of either boric acid makeup tank in accordance with Figure 3.1-1. the following flow paths shall be OPERABLE:
1. One flow path from an acceptable boric acid makeup tank via its boric acid makeup pump; and
2. One flow path from an acceptable boric acid makeup tank via its gravity feed valve; or
b. With the combined contents of both boric acid makeup tanks in accor-dance with Figure 3. both of the following flow paths shall be OPERABLE:
1. One flow path consisting of both boric acid makeup pumps, and
2. One flow path consisting of both gravity feed valves.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

With only one of the above required boron injection flow paths to the Reactor Coolant System OPERABLE, restore at least two boron injection flow paths to the Reactor Coolant System to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least NOT STANDBY and borated to a SHUTDOWN MARGIN equivalent to the requirements of Specification 3.1.1.1 or 3.1.1.2, whichever is applicable, within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; restore at least two flow paths to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQUIREMENTS 4.1.2.2 At least two of the above required flow paths shall be demonstrated OPERABLE:

7Ho tepJ~ftue 1ot'low5WPb verj)sg~ng cthtaatz;he tem' ftre~

a),'. At least once per 31 days by verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.

6 M. At least once per 18 months during shutdown by verifying that each automatic valve in the flow path actuates to its correct position on an SIAS test signal.

c %r. At least once per 18 months by verifying that the flow path required by Specification 3.1.2.2a.1 and 3.1.2.2a.2 delivers at least 40 gpm to the Reactor Coolant System.

WATERFORD - UNIT 3 3/4 1-7 AMENDMENT NO. Q+r

REACTIVITY CONTROL SYSTEMS BORATED WATER SOURCES - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.12.7 As a minimum, one of the following borated water sources shall be OPERABLE:

a. One ad makeu ank with a boron concentration betwee w andminimum borated water volume 36 ndicated
b. The refueling water storage pool (RWSP) with:
1. A minimum contained borated water volume of 12% indicated level, and
2. A minimum boron concentration of 2050 ppm.

APPLICABILITY: MODES 5 and 6.

ACTION:

With no borated water sources OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes.

  • SURVFELIL6LANCE REQUIREMENTS 4.1.2.7 The above required berated water source shall be demonstrated OPERABLE:
a. At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when the Reactor Auxiliary Building air temperature is less than 55*F by verifying the boric acid makeup tank solution 1e F ewhen it is the source of borated water).
b. At least once per7ays by.
1. Verifying the boron concentration of the water, and
2. Verifying the contained borated water volume of the tank.

' Plant temperature changes are allowed provided the temperature change is accounted for in the calculated SHUTDOWN MARGIN.

WAT.ERFORD - UNIT 3 3J4 1-12 AMENDMENT NO. 10, i29,485

REAUC I MvI YCONTROL SYSTEMS BoRATED WATER SOURCES - OPERATING i WITING CONDITION FOR OPER ON 3.1.2.8 Each of the following borated water sources stha be OPERABLE:

a. At east one of the following sources:
1) One boric acid makeup tank with th tank contents in accordance with Figure 3.1-1, or
2) Two boric add makeup tanks, with the ned contents of to tanks in accordance with Figure 3.1
b. The refuelingwater storep In L;,Ca (1 ~mumc nodofatewat vlme
2. / bonc7h of be 209ad2 nof 7n71

. ,psolffonf praturohe n 55; an I04:

A!PLICAB13M: MODES 1, 2, 3, and 4.

AMnON:

a. With the above reqired boric acid makeup tank(s) Inoperable, restore the tank(s) to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be In at least HOT STANDBY within the rext 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and boratod to a SHUTDOWN MARGIN equialent to the requirements of Specification 3.1.1.1 or 3.1.1.2, whichever Is applicable; restore the above required boric acid makeup tank(s) to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
b. With the refueling water storage pool Inoperable, restore the pool to OPERABLE stua within I hour or be in at least HOT STANDBY within the next 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REOUIREMENTS 4.1.2.8 Each borated water source shall be deonad OPERABLE:

X At least onoo per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by veiing te ur the ReactorAwllay B airtemperature is less than 55-F.

,c At least or per 7 daby.

1. Vein the bor concentration Inthe wor, and
2. Veying the cotined borad water volume of the water source.

WATERFORD- UNIT 3 3/4 1-6) AMENDMENT NO. 40.4 .129. 4Q

Replqce WIf;lA nIsert-F~igre FIGURE3.1-1 WATERFORD - UNIT 3 3/4 1-9k AMENDMENT NO. 10, 45. 1 44X

I 'e (JUtser/--e F; *3*1-1-)

REQUIRED STORED BORIC ACID VOLUME AS A FUNCTION OF CONCENTRATION (VOLUME OF ONE BAMT) 11500 . _*i l (96%) n' I i 10 - t.-. ;_REGION OF ACCEPTABLE _

110 8 ,OPERATION (92%) I 10500- . , I (87%).'\ lIi W ~sS3o 110000 (X

82%) .,;, , lws20l 0

oz 9 -i I

> (7_%)

j  : j / ' . .

(z69%)'

DORIC ACID CONCENTRATION, ppmn

,4e, --',u -6 - ;I REQUIRED STORED BORIC ACID VOLUME AS A FUNCTION OF CONCENTRATION (COMBINED VOLUME OF TWO BSAMT) 17.5wo i

ZI 0 =

2' c w

= ,

O w kn j aZ 5200 5400 5600 5800 BORIC ACID CONCENTRATION. .; .. ppm Figure 3.1-2 WATERFORD - UNIT 3 3/M 1-143

POWER DISTRIBUTION LIMITS 3/4.2.6 REACTOR COOLANT COLD LEG TEMPERATURE LHnITING CONDITION FOR OPERATION 3*26 ctorL ooiant cold leg temperature (T1) shall be maintained aIh betwee F and F.*

APPLICABILIT: MODE I above 30% of RATED THERMAL POWER.

ACTION:

With the reactor coolant cold leg temperature exceeding its liet, restore the temperature to within its limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to less than 30X of RATED THERKAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.2.6 The reactor coolant cold leg temperature shall be determined to be wlithin its 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 />.

  • Followng a reactor power cutback in which (1)Regulating Groups S and/or 6 are dropped or (2)Regulating Groups 5 and/or 6 are dropped and the remaining Regulating Groups (Groups 1, 2, 3r nd 4) are sequentially inserted, the upper IImit on T. may increase to F for up to 30 minutes.

WATERFORD - UNIT 3 3/4 2-11 AMENDMENT NO - Aea

POWER DISTRIBUTION LIMITS 3/4.2.8 PRESSURIZER PRESSURE

.LIMITING CONDITION FOR OPERATION I

2. LThe steady-state pressurizer pressure shall be maintained between sia and 2275 psia.

APPLICABILITY: MODE 1 ACTION:

With the steady-state pressurizer pressure outside its above limits, restore the pressure to within its limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to less than 5X uf RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.2.8 The steady-state pressurizer pressure shall be determined to be within its 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 />.

WATERFORD - UNIT 3 3/4 2-13

'KBLE3.34 ENGIEERD

$FET FE~URE A~U~lON-SYI ALLOWABLE EU OQNALUNI vnnet1DT

, . a,. __ .. _ .... -VALUM SAFETY INJECTION (SIAS)

a. Manual (Trip Buttons) Not Applicable Not AppNcable
b. Contahmenlt Pressure -Hgh s 17.1 psla s 17.4 psha
c. Pressurizer Pressure - Low 2 1684 psld" 2 1649.7 psla'"
d. Automatic Actuation Logic Not Apdicable Not APecb
2. CONTAINMENT SPRAY (CSAS)
a. Manual (rP Buttons) Not Ap1 cab4e Not cable
b. Contahment Pmessum - Hklh'High s 17.? psia '11.Opsla
c. Automatic Acuation Logic Not Apikable Not Apcable
3. CONTAINMENT ISOLATION (CIAS)
1. Manual CWAS (Trip Elutt) Not Apcable Not Applicable
b. Contuinment Pressure - High s 17.I pals s 17.4 psin
c. Pressurizer Pressure - low k 1684 psoi" 1849.7 pslgf"
d. Automatic Actuation Logic Not Applicable NotAmacabie
4. MAIN STEAM LINE ISOLATION
a. Manual (Trip Buttons) Not Applcable_,

2p~a

b. Steam Generator Pressure -Low
c. ContainmntPtessure -High s 17.1 psia 17.4 psis
d. Automatic Actuation Logic Not Apcable Not Appable

}=I314 3.19 AMENDMENT NO. +3&,

TABL - -4 (Continued)

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMIENTATION TRIP VALUES

- w^

ALLOWABLE

[UCTIONALUN TRIP VALUE ALUESL

5. SAFETY INJECTION SYSTEM SUMP RECIRCULATION (RAS)
a. Manual RAS (Trip Buttons) Not Applicable Not Applicable
b. Refueling Water Storage Pool - Low 10.0% (57,967 gallons) 9.08% (52,634 gallons)
c. Automatic Actuation Logic Not Applicable Not Applicable
6. LOSS OF POWER
a. 4.16 kV Emergency Bus Undervoltage  ? 3245 volts 2 3245 volts (Loss of Voltage)
b. 480 V Emergency Bus Undervoltage 2 372 volts 2 354 volts
c. 4.16 kV Emergency Bus Undervoltage (Degraded Voltage) 2 3875 volts 2 3860 volts
7. EMERGENCY rEEDwATER (EFAS)
a. Manual (Trip Buttons) Not Applicable Not Applicable
b. Steam Generator (112) Level - Low 2 27.I0) (4) 2_26.48X(3)(4)
c. Steam Generator AP - High (SG-1 > SG-2) 1 123 psid S 134 psid
d. Steam Generator AP - High (SG-2 > SG-l) S 123 < 134 p
e. Steam Generator (112) Pressure - Low k iŽac) f@psia('
f. Automatic Actuation Logic Not Applicable Not Applicable
g. Control Valve Logic (Wide Range SG Level - Low) k 363X (

(5) 2 35.3X°'")

REACTOR COOLANT SYSTEM i7 g z<twoo pe'r 4 a l4 p rW"r/-

at6- secodot-1 letk-ye per OPERATIONAL LEAKAGE s5 qe- Sf P eraz 4o 3.4.5.2 Reactor Coolant System leakage shall be limited to:

a. No PRESSURE BOUNDARY LEAKAGE,
b. 1 gpe UNIDENTIFIED LEAKAGE,
c. te
d. 10 gpa IDENTIFIED LEAKAGE from the Reactor Coolant System, and
e. 1 gpm leakage at a Reactor Coolant System pressure of 2250 i 20 psia from any Reactor Coolant System pressure isolation valve specified in Table 3.4-1.

APPLICABILITY: MODES 1, 2,. 3, and 4.

ACTION:

a. With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
b. With any Reactor Coolant System leakage greater than any one of the limits, excluding PRESSURE BOUNDARY LEAKAGE and leakage from Reactor Coolant System pressure isolation valves, reduce the leakage rate to within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
c. With any Reactor Coolant System pressure isolation valve leakage greater than the above limit, isolate the high pressure portion of the affected system from the low pressure portion within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by use of at least one closed manual or deactivated automatic valve, or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQUIREMENTS 4.4.5.2.1 Reactor Coolant System leakages shall be demonstrated to be within each of the above limits by:

a. Comrencing an RCS inventory balance within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to determine the leak rate when RCS leakage is alarmed and confirmed in a flow path with no flow rate indication.
b. Monitoring the containment atmosphere gaseous and particulate radioactivity monitor at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
c. Monitoring the containment sump inventory and discharge at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

WATERFORD - UNIT 3 3/4 4-3.8

4.4.8.2 'The spray water tampe re differential shall b etermined to be withi he limit at least once p 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> during auxilia spray operation.

.4.8.2.3 Each spray cy and the correspondi /T (water temperature differential) shall be rerded whenever main s Is Initiated with a dT (ft temperature differe ial) of> 1301F and whe ver auxiliary spray Is initl'ed with a aT (water mperature differential) 140F. /

WATERFORD - UNIT 3 314 4-33 AMENDMENT NO. jj-r-

3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3/4.5 1 SAFETY INJECTION TANKS LIMITING CQNDITION FOR QPERTION_

3.5.1 Each Reactor Coolant System safety injection tank shall be OPERABLE with:

a. The isolation valve open, 17
b. A contained borated water volume of between&l0°h/an
c. Between 2050 and 2900 ppm of boron, and
d. A nitrogen cover-pressure of between 600 and 670 psig.

APPLICABILI: MODES 1. 2, 3'. and 4*.

ACTION: MODES 1, 2, 3 and 4 with pressurizer pressure greater than or equal to 1750 psia.

a. With one of the required safety injection tanks inoperable due to boron concentration not within limits, restore the boron concentration to within limits within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be In at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and reduce pressurizer pressure to less than 1750 psia within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
b. With one of the required safety Injection tanks Inoperable due to inability to verify level or pressure, restore the tank to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and reduce pressurizer pressure to less than 1750 psla within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
c. With one of the required safety injection tanks inoperable for reasons other than ACTION a or b, restore the tank to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, or be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and reduce pressurizer pressure to less than 1750 psia within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

'With pressurizer pressure greater than or equal to 1750 psi When pressurizer pressure is less than 1750 psla. at least three safety Injection tanks m st be OPERABLE, each with a minimum pressure of 235 psig and a maximum pressur f 670 psig, and a contained borated water volume of between4f01%Tnd With all four safety Injection tanks OPERABLE, each tank shall have a minimum pressure of 235 psig and a maximum pressure of 670 psig, a boron concentration of betwen 2050 and 2900 m n, and a contained borated water volume of betweendQ9 eend In MODE 4 with pressunzer pressure less than 392 psia (700 psia for remote shutdown from LCP-43), the safety injection tanks may be isolated.

WATERFORD - UNIT 3 3t4 5-1 AMENDMENT NO. 121,12944 457

EMERGENCY CORE COOLING SYSTEMS 314.5.4 REFUELING WATER STORAGE POOL LIMITING CONDITION FOR OPERATION 3.5.4 The refueling water storage pool shall be OPERABLE with:

a. A minimum contained borated water volume ok% indicated
b. Between 2050 and 2900 ppm of boron, and Or e
c. A solution temperature of D55'FandF1Oa F.

APPLIC61A3-TY: MODES 1, 2, 3, and 4.

ACTION:

With the refueling water storage pool inoperable, restore the pool to OPERABLE status within I hour or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and In COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQUIREMENTS 4.5.4 The RWSP shall be demonstrated OPERABLE:

a. At least once per 7 days by:
1. Verifying the contained borated water volume in the pool, and
2. Verifying the boron concentration of the water.
b. At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by verifying the RWSP temperature when the RAB air temperature is less than 56F or greater than I 00F.

WATERFORD - UNIT 3 3/4 5-9 AMENDMENT NO. 4G49.42.4i-

CONTAINMENT SYSTEMS AIR TEMPERATURE LIMITING CONDITION FOR OPERATION 3.6.1.5 Primary containment average air temperature shal l 120 OF.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION: t With the containment average air temperature, the average air temperature to within the limfhwithin 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in\COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQUIREMENTS 4.6.1.5 The primary containment average air temperature shall be the arithmetical average of the temperatures at any three of the following locations and shall be determined at least once per 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />s:

Location

a. Containment Fan Cooler No. 1A Air Intake
b. Containment Fan Cooler No. 18 Air Intake
c. Containment Fan Cooler No. 1C Air Intake
d. Containment Fan Cooler No. 10 Air Intake
  1. T/ae m co~f+ ;nnsee d vercle CZtr Mperla are

/Inif t~ c,$4 only;Cz c c,+

-e Srreqier thkqnj 76 1AT*D otJ AM L P WF R WATERFORD - UNIT 3 3/4 6-13

314.7 PLANT SYSTEMS 314.7.1 TURBINE CYCLE SAFETY VALVES LIMITING CONDITION FOR OPERATION 3.7.1.1 All main steam fine code safety valves shall be OPERABLE with lift settings as specified In Table 3.7-1.

APPLICABILITY: MODES 1, 2, and 3.

ACTION:

a. =eco nte ao 4one or mere main steam linecode safetyvaves oint peration In POWER fistedin Table3.7o2 and withi12ours Le thesLinear Powe Levr etpoll g

trip int in pccordTanewith2 theweleast be It HOTe nDBY within The provisions nsevicofwth accodane Specification Tetin he Proram 3.0.4 Fol~Ig are not applicable.

tstig, iltseting shale wIthin SURVEILLANCE REQUIREMENTS-4.7.1.1 Verify each required main steam line code safety valve lift setpoint per Table 3.7-1 in Iniae pwrt accordance lDES tha n Testing with the Inservice or eqalProgram.

tocidpoie~t apliabe erenrRTEbTERA ihn4hXete the lift settings shall be within Following testing, v lhone or more main steam line code satety valve inoperable, withfin 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> reduce indicated power to less than or equal to the applicable percent RATED THERMAL POWER listed in Table 3.7-2 and within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> reduce the Linear Power Level - High trip setpoint in accordance with Table 3.7-2, otherwise, be In at least HOT STANDB3Y within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN withiln the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

WATERFORD - UNIT 3 3/4 7-1 AMENDMENT NO. H+i48,

.41- c¢:Aais ctre- he,'")

TABLE3..7A1 mfqde fo0 ' pA'e.

S1EAM LINE SAFETY VALVES-PER LOOP iv l 4 o P

&LA NUMBER LIFT SETTING (. 3%)

Lf No LkbflQ2

a. 2MS-R613A 2MS-R6198 1070 psig (MS-106A) (MS-1068)
b. 2MS-R614A 2MS-R620B 1085 psig (MS-1OOA) (MS-1088)
c. 2MS-R615A 2MS-R621 B 1100 psig (MS-110A) (MS-1108)
d. 2MS-R6i6A 2MS-R622B 11 15 psig (MS-112A) (MS-112B)
e. 2MS-R617A 2MS-R623B 1t25 psig (MS-113A) (Ms-I 138)
f. 2MS-R618A 2MS-R624B 11 35 psig (MS-i14A) (MS-1148) rThe Off setting pressure shall correspond to ambient conditions of the valve at nroinal operaing temperature and pressure.

WATERFORD -UNIT 3 314 7-2 AMENDMENT NO. 4--iti*

< TABLE 3.7-2 MAXMUMALLWALE INER P~iR LEVEL-HIGH DBIP SMIQW iR BL STEAM LIN SAFE VLE 99EnQ 3 1Amw EN 90

/MA~zMALLaAB L MAXIMUM NUMBER OF INOPERABLE SAFETY LEVE GH E VALVES ON ANY QPEERTING STEAM GENERATOR lllEl 2 .4 WATERFORD -1UNIT 3 3/4 7-3 AMENDMENT NO. 444, 44-

Insert Table 3.7-2 MAXIMUM ALLOWABLE POWER LINEAR POWER LEVEL - HIGH

(% RTP) ' TRIP SETPOINT (% RTP) 85.3 5 93.3 66.7 s74.7

PLANT SYSTEMS CONDENSATE STORAGE POOL LIMITING CONOITION FOR OPERATION 3.7.1.3 The condensate stora to (CSP) shall be OPERABLE w1, nimuu contained volume of at least indicated level.*

APPLICABILITY: MODES 1, 2, 3 and 4. b A re fer -eka er-ar'ire t ACTIQN: 6'S-' <°leJss 4-hq-f qkf ar In MODES 1. 2. and 3:

With the condensate storage pool inoperable, within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> restore the CSP to OPERABLE status or be In at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

In MODE 4:

With the condensate storage pool inoperable. within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> restore the.

CSP to OPERABLE status or be tn at least COLD SHUTDOWN within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.7J.,3L__Ihe condensate stora -poolshall be demonstrated OPERAB l on~ce per 12 h-our-slyy ve-rify ng the cntne-tr ou s th 1_--

t ¢jpeq4 okrfe per2 LI ho^ r b)S lt verlo-rx)vc Css less It or~ rcxe eS~ft O a fr. le l 4hat ,

  • In MODE 4, the CSP shall be OPERABLE with a minim=u contained volume of at least 11% indicated level.

WATERFORD - UNIT 3 3/4 7-6 Amendment No.+3r?-

MIN STEAM LINE ISOLATION VALVES WMSIW9 HLMTbLON _CONQIT1ON FOR QPERA11ON 3.7.1.5 Two MSIVs shall be OPERABLE.

APPLICABILfTYD MODE 1, and MODES 2.3. and 4. except when all MSIVs are closed and deactivated.

ACTIO MODE I

'Mh one MSiV Inoperable, restore the valve to OPERABLE status within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or be In STARTUP within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

MODES 2.3 and 4 With one MSIV Inoperable, close the valve within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and verify the valve Is dosed once per 7 days. Otherwise, be In HOT STANDBY wthin the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN Within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS Note: Required to be performed for entry into MODES 1 and 2 only.

4.7.1.5 Each MSIV shall be demonstrated OPERABLE:

a. By verifying full closure withi aecond3 when teeted pursuant to the Ilnscrvice Testing Program.
b. By verifying each MSIV actuator to the Isolation position on an actual or simulated actuation signal at least once per 18 months.

WATERFORD - UNIlT 3 314 7-9 AMENDMENT NO. ok89.+",

PLANSYSIM MAIN IFEEDWATER-5Q LATION-VALSU LIMMNG CONOmON FOR OPERATiON 3.7.1.6 Each Main Feedwater Isolation Valve (MFIV) shall be OPERABLE.

APPLICILITY: MODES 1,2, 3, and 4.

ACT'ON Note: Separate Condition entry Is allowed for each valve.

With one or more MFIV inoperable, close and deactivate, or Isolate the inoperable valve within 72 hourt and verify Inoperable valve closed and deactivated or Isolated once evary 7 days; otherwise, be Inat least HOT STANDBY within the nexd 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

The provisions of Specification 3.0.4 do not apply.

SURVEIL-ANCE REQUIREMENTS 4.7.1.6 Each main feedwater Isolation valve shall be demonstrated OPERABLE:

a. By verifying isolatlo seconds when tested pursuant to the Inservice Testing Program.
b. By verifying actuation to the Isolation position on an actual or simulated actuation signal at least once per 18 months.

WATERFORD - UNiT 3 314 7-9a AMENDMENT NO. 461Cd19;

T SYSTs 14.7.1.7 ATMOSPHE-RIC UPVLE sg LIMITING CONDITION FOR OPERATION 3.7.1.7 Each Atmospheric Dump Valve (ADV) shall be OPERABLE*.

APPLICABILITY: MODES 1, 2, 3, and 4 ACTION:

a. With the automatic actuation channel for one ADV inoperable, restore the inoperable ADV to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or reduce power to less than or equal to 70% RATED THERMAL POWER within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
b. With the automatic actuation channels for both ADVs inoperable, restore one ADV to OPERABLE status within 1hour or reduce power to less than or equal to 70% RATED THERMAL POWER within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
c. With one ADV Inoperable, for reasons other than above, restore the ADV to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

The provisions of Specificalion 3.0.4 are not applicable provided one ADV Is OPERABLE.

SURVEILLANCE REQUIREMENTS 4.7.1.7 The ADVs shall be demonstrated OPERABLE:

a. By performing a CHANNEL CHECK at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the automatic actuation channels are required to be OPERABLE.
b. By verifying each ADV automatic actuation channel is in automatic with a setpoint of less than or equal to 1040 psia at least once per 92 days when the automatic actuation channels are required to be OPERABLE.
c. By verifying one complete cycle of each ADV when tested pursuant to the Inservice Testing Program.
d. By performing a CHANNEL CALIBRATION of each ADV automatic actuation channel at least once per 18 months.
e. By verifying actuation of each ADV to the open position on an actual or simulated automatic actuation signal at least once per 18 months.
  • ADV automatic actuation channels (one per ADV, in automatic with a setpoint of less than or equal to 1040 psia) are not required to be OPERABLE when less than or equal to 70% RATED THERMAL POWER for greater than 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

WATERFORD - UNIT 3 3/4 7-9b

5-0 DESGIN FFATURFS 5,I SITE EXCLUSION AREA 5.1.1 The exclusion area shall be as shown in Figure 5.1-1.

LOW POPULATION ZONE 5.1.2 The low population zone shall be as shown in Figure 5.1-2.

MAP DEFINING UNRESTRICTED AREAS FOR RADIOACTIVE GASEOUS AND LIOUID EFFLUENTS 5.1.3 Information regarding radioactive gaseous and liquid effluents, which will allow identification of structures and release points as well as definition of UNRESTRICTED AREAS within the SITE BOUNDARY that are accessible to MEMBERS OF THE PUBLIC, shall be as shown in Figure 5.1-3.

The definition of UNRESTRICTED AREA used in implementing these Technical Specifications has been expanded over that in 10 CFR 20.1003. The UNRESTRICTED AREA boundary may coincide with the Exclusion (fenced) Area boundary, as defined in 10 CFR 100.3(a), but the UNRESTRICTED AREA does not include areas over water bodies. For calculations performed pursuant to 10 CFR 50.36a, the concept of UNRESTRICTED AREAS, established at or beyond the SITE BOUNDARY, is utilized in the Controls to keep levels of radioactive materials in liquid and gaseous effluents as low as is reasonably achievable.

5j2 _ - a.f e D WATERFORD - UNIT 3 A_ A_ . & I ... t

DESIGN FEATURES 5 6 FUEL STORAGE CRTICALY 5.6.1 The spent fuel storage racks are designed and shall be maintained with:

a. A normal krff of less that or equal to 0.95 when flooded with unborated water, which includes a conservative allowance for uncertainties.
b. A nominal 10.185 inch center-to-center distance between fuel assemblies placed in Region I (cask storage pit) spent fuel storage racks.
c. A nominal 8.692 Inch center-to-center distance between fuel assemblies Inthe Region 2 (spent fuel pool and refueling canal) racks, except for the four southern-most racks In the spent fuel pool which have an Increased N-S center-to-center nominal distance of 8.892 inches.
d. New or partially spent fuel assemblies may be allowed unrestricted storage in Region I racks.
e. New fuel assemblies may be stored in the Region 2 racks provided that they are stored in a 'checkerboard pattern' as illustrated in Figure 5.6-1.
f. Partially spent fuel assemblies with a discharge bumup in the 'acceptable range' of Figure 5.6-2 may be allowed unrestricted storage In the Region 2 racks.
g. Partially spent fuel assemblies with a discharge bumup in the 'unacceptable range of Figure 5.6-2 may be stored Inthe Region 2 racks provided that they are stored In a 'checkerboard pattern, as illustrated In Figure 5.6-1, with spent fuel in the

.acceptable range' of Figure 5.6-3.

5.6.2 The k,fr for new fuel stored in the new fuel storage racks shall be less than or equal to 0.95 when flooded with unborated water and shall not exceed 0.98 when aqueous foam moderation Is assumed.

DRAINAGE 5.6.3 The spent fuel pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation +40.0 MSL. When fuel is being stored Inthe cask storage pit and/or the refueling canal, these areas will also be maintained at +40.0 MSL CAPACIT 5.6.4 The spent fuel pool Is designed and shall be maintained with a storage capacity limited to no more than 1849 fuel assemblies in the main pool, 255 fuel assermblies in the cask storage pit and after permanent plant shutdown 294 fuel assemblies in the refueling canal. The heat load from spent fuel stored in the refueling canal racks shall not exceed 1.72x1DE6 BTU/IH-r. Fuel shall not be stored in the spent fuel racks Inthe cask storage pit or the refueling canal unless all of the racks are installed in each respective area per the design.

WATERFORD - UNIT 3 5-6 AMENDMENT NO. A106,1, 488-

M4 COMPONENT CYCLIC OR TRANSIE1tLMT

°ICS:' CYCLIC OR DESIGN CYCLE o COMPONENT TRANSIENT LIMIT OR TRANSIENT from < .700F c Reactor Coolant System 500 system heatup cycles and

~49at rates < 1O0°F/h. cyl 500 pressurizer heatup and Heatup cycle - Pressurizer temperature cooldown cycles at rates from < 700F to > 6530 F; cooldown cycle

< 2000F/h. > 653SF to < 7OuF 10 hydrostatic testing cycles. RCS pressurized to 3125 psia with RCS temperature > 601F above the most limiting components' NDTT value%

200 leak testing cycles. RCS pressurized to 2250 psia with RCS temperature greater than minimum for hydrostatic testing, but less than 4000F.

200 seismic stress cycles. Subjection to a seismic event equal to the operating basis earthquake (OBE).

480 cycles (any combination) Trip from ]00% of RATED THERMAL POWER; M of reactor trip, turbine trip, Turbine trip (total load rejection or complete loss of forced from 100% of RATED THERMAL POWER t reactor coolant flow followed by resulting reactor trip; simultaneous loss of all reactor coolant pumps at 100% of RATED THEF4MAL POVER

( 5 complete loss of secondary Loss of secondary pressure from

\ pressure cycles. either steam generator while in MODE 1,2, or 3.

\ DESIGN CYC OR TRANSIENT Main sp y (4 pumps operX ng).p t Main spra (less than 4 p s operating)\

wi th AT < OS0F\\

Auxiliavy sp with Ta < 14 0 F Main spray (Les than 4 pumps ating)

\with AT > 1300 F F r

'wxit spray wi AT- > 140°F /

Calculational Method.

1 The spray cycle i efined as the ope g and closing of a pray valve elt r 2.

y main spray or aux lary spray.

If! e difference betwee:pressurizer water emperature and the tempe attre exceeds 13D°F ray water r main spray or 1 0F for auxiliary s ay each

)

3.

spray cle and the correspo ing temperature d ference is logged.

The spray A.

zzle usage factor sh 11 be calculated a follows::

Fill in C umn HNI above.

(

B. Calculate N " (Divide N and NA C. Add Column "N/NA to find YN/NA.

Add IN/N for both in and auxiliary s ray.

\ sage fahlor. \ *\\\

This total 1 the cumulative (

4. A. I the cumulative usage actor is equal to o less than 0.65 no rther action is tuired.

B. If the >ulative usage facto exceeds 0.65, subse ent pressurizer sp y operation hall continue to be nitored and an eng eering evaluation o nozzle fatigue sha be perforied within 0 days. The evalu ion shall determine that the nozzle re ains acceptable for a itional service beknd the 90-day per. o or

\ subsequent spra operation shall be stricted so that th differende between he essurizer wate temperature and the ray water temperaturN shall be limited le < than or equal 130°F for main s r and 140°F for auxiftary s

zz ADMINISTRATIVE CONTROLS 6.3 UNIT STAFF QUALIFICATIONS 6.3.1 Each member of the unit staff shall meet or exceed the minimum qualifications of ANSI/ANS 3.1-1978 except that

a. The Radiation Protection Superintendent shall meet or exceed the minimum qualifications of Regulatory Guide 1.8. September 1975.
b. Personnel Inthe Health Physics, Chemistry and Radwaste Departments shall meet or exceed the minimum qualifications of ANSI N18.1-1971.
c. The licensed Operators and Senior Operators shall also meet or exceed the minimum qualifications of 10 CFR Part 55.
d. Personnel Inthe Nuclear Quality Assurance Department, and other staff personnel who perform Inspection. examination, and testing functions, shall meet or exceed the minimum qualifications of Regulatory Guide 1.58, Rev. 1, September 1980.

(Endorses ANSI N45.2.6-1978).

6.4 TRAINING 6.4.1 A retraining and replacement training program for the unit staff shall be maintained under the direction of the Training Manager-Nuclear and shall meet or exceed the requirements and recommendations of Section 5.2 of ANSI 3.1.1978 and 10 CFR Part 55.

6.5 PRO S Ae proyrcav tldowhuy sh i e 4& I beusf~led la eH1ez qros 6.5.1 through ll be used later. _ _

.5.7 REACTOR COOLANT PUMP FLYWHEEL INSPECTION PROGRAM This program shall provide for the inspection of each reactor coolant pump flywheel per the recommendation of Regulatory Position CA.b of Regulatory Guide 1.14, Revision 1.August 1975. ThA volumetric examination per Regulatory Position C.4.b.1 will be performed on approximately 10-year intervals.

6;, 1;i' be ast '4 6.5-, ;J/ e 5, /ofP.

WATERFORD - UNIT 3 6-7 AMENDMENT NO. 1O643.61,63.9.109,1e9e-

ADMINIsTRATiVE CONTROLS CORE OPERATING LIMITS REPORT COLR (Continued)

6) CESEC - Digttal.Slmulatlon for a Combustlon Engineering Nuclear Steam Supply System,' CENPD-1 07. (Methodology for Specification 3.1.1.1 and 3.1.12 for Shutdown Margins, 3.1.1.3 for MTC, 3.1.3.1 for Movable Control Assemblies - CEA Position, 3.1.3.6 for Regulating and group P CEA Insertion Limits, and 32.3 for Azimuthal Power Tilt).
7) "Qualification of Reactor Physics Methods for the Pressurized Water Reactors of the Entergy System,' ENEAD-01-P. (Methodology for Specifications 3.1.1.1 anr 3.1.1.2 for Shutdown Margins, 3.1.1.3 for MTC, 3.1.3.6 for Regulating and group P CEA Insertion Limits, 3.1.2.9 Boron Dilution (Calculation of CBC & IBW), and 3.9.1 Boron Concentration).
8) OFuel Rod Maximum Allowable Gas Pressure: CEN-372-P-A (Methodology for Specification 32.1, Unoar Heat Rate).

6.9.1.112 The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal limits, core thermal-hydraulic limits. ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met.

6.9 1.11.3 The CORE OPERATING LIMITS REPORT. Including any mid-cyde revisions or supplements thereto, shall be provided upon issuance, for each reload cycle, to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.

SPECIAL REPORTS 6.9.2 Special reports shall be submitted in accordance with 10 CFR 50.4 within the time period specified for each report.

\6.10 Not Used AMENDMENT NO. 402 146+58,42 WATERFORD - UNIT 3 6-20a 4887-49

TS 6.9.1.11.1 Insert

9) *Technical Description Manual for the CENTS Code," WCAP-1 5996-P-A. (Methodology for Specification 3.1.1.1 and 3.1.1.2 for Shutdown Margins, 3.1.1.3 for MTC, 3.1.3.1 for Movable Control Assemblies - CEA Position, 3.1.3.6 for Regulating and group P CEA Insertion Limits, and 3.2.3 for Azimuthal Power Tilt).
10) "Calculative Methods for the CE Nuclear Power Large Break LOCA Evaluation Model,"

CENPD-132, Supplement 4-P-A. (Methodology for Specification 3.1.1.3 for MTC, 3.2.1 for Linear Heat Rate, 3.2.3 for Azimuthal Power Tilt and 3.2.7 for ASI).

Attachment 3 W3FI-2004.0052 Changes to Technical Specification Bases Pages (mark-up)

For Information Only

Manual ReactonrTrin The Manual Reactor Trip is a redundant channel to the automatic protective instrumentation channels and provides manual reactor trip capability.

I Inear Power Level - High The Linear Power Lev !Ig11 .pri DI -sacto tection against rapid reactivity excursions ht :rl __ I-u This trip initiates a reactor trip at a linear power level of less than or ecual to 108%& TED THERMAL POWER.

Logarithmic Power Level - High The Logarithmic Power Level - High trip is provided to protect the Integrity of fuel cladding and the Reactor Coolant System pressure boundary in the event of an unplanned criticality from a shutdown condition. A reactor trip Is Initiated by the Logarithmic Power Level - High trip at a THERMAL POWERS level of less than or equal to 0.257% of RATED THERMAL POWER' unless this trip is manually bypassed by the operator. The operator may manually bypass this trip when the THERMAL POWER* level is above 10-4% of RATED THERMAL POWER'; this bypass is automatically removed when the THERMAL POWER' level decreases to 10-4% of RATED THERMAL POWER*.

Pressurizer Pressure - High The Pressurizer Pressure - High trip, in conjunction with the pressurizer safety valves and main steam safety valves, provides Reactor Coolant System protection against overpressurization in the event of loss of load Without reactor trip. This trip's setpoint is at less than or equal to 2350 psia which is below the nominal lift setting of 2500 psia for the pressurizer safety valves and Its operation avoids the undesirable operation of the pressurizer safety valves.

Presnuirizer Pressure - Low The Pressurizer Pressure - Low trip Is provided to trip the reactor and to assist the Engineered Safety Features System in the event of a Loss of Coolant Accident During normal operation, this trip's setpoint Is set at greater than or equal to 1684 psia. This trip's setpoint may be manually decreased, to a minimum value of 100 psla, as pressurizer pressure is reduced during plant shutdowns, provided the margin between the pressurizer pressure and this trip's setpoint is maintained at less than or equal to 400 psi; this setpoint increases automatically as pressurizer pressure Increases until the trip selpont is reached.

'As measured by the Logarithmic Power Channels.

WATERFORD - UNIT 3 192-3 Amendment No. +0:1S

SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS BASES Containment Pressure - High The Containment Pressure - High trip provides assurance that a reactor trip is initiated concurrently with a safety injection, a containment isolation, and a main steam isolation. The setpoint for this trip Is Identical to the ESFAS setpoint.

Steam Generator Pressure - Low The Steam Generator Pressure - Low t provides protection against an excessive rate of heat extraction from tie steam generators and subsequent cooldown of the reactor coolant. The j oint is sufficiently below the full load operating point of approximately psia so as not to interfere with normal operation, but still high enougbto provide the required protection in the event of excessively high steam flow. This trip's setpoint may be manually decreased as steam generator pressure is reduced during plant shutdowns, provided the margin between the steam generator pressure and this trip's setpoint is maintained at less than or equal to 200 psi; this setpoint increases automatically as steam generator pressure increases until the trip setpoint is reached.

Steam Generator Level - Low The Steam Generator Level - Low trip provides protection against events involving a mismatch between steam and feedwater flow. These may be due to a steam or feed line pipe break or other increased steam flow or decreased feed flow events. A large feedwater line break avent inside containment establishes the trip setpoint. The setpoint ensures that a trip will occur before the steam generator heat sink is lost. The trip setpoint also ensures that the Reactor Coolant System design pressure will not be exceeded prior to the time emergency feedwater can be supplied for decreased heat removal events such as a loss of condenser vacuum or loss of feedwater flow.

Local Power Density - High The Local Power Density - High trip is provided to prevent the linear heat rate (kW/ft) in the limiting fuel rod in the core from exceeding the fuel design limit in the event of any anticipated operational occurrence. The local power density is calculated in the reactor protective system utilizing the following information:

a. Nuclear flux power and axial power distribution from the encore flux monitoring system;
b. Radial peaking factors from the position measurement for the CEAs;
c. Delta T power from reactor coolant temperatures and coolant flow measurements.

WATERFORD - UNIT 3 3 2-4

DNBR - Low (Continued) in actual core DNBR after the trip will not result In a violation of the DNBR Safety Limit of 1.26.

CPC uncertainties related to DNBR cover CPC input measurement uncertainties, algorithm modelling uncertainties, and computer equipment processing uncertainties. Dynamic compensation is provided in the CPC calculations for the effects of coolant transport delays, core heat flux delays (relative to changes in core power), sensor time delays, and protection system equipment time delays.

The DNBR algorithm used in the CPC is valid only within the limits indicated below and operation outside of these limits will result in a CPC initiated trip.

a. RCS Cold Log Tamporaturo-Low I405F
b. RCS Cold Leg Temperature-High c 580*F
c. Axial Shape Index-Positive Not more positive than +0.5
d. Axial Shape Index-Negative Not more negative than -0.5
e. Pressurizer Pressure-Low 2 1860 psia
f. Pressurizer Pressure-High c 2375 psia
9. Integrated Radial Peaking Factor-Low 2 1.28
h. Integrated Radial Peaking Factor-High .57.00
i. Quality Margin-Low >0 ..

Steam Generator Level - Hiah The Steam Generator Level - High trip Is provided to protect the turbine from excessive moisture carry over. Since the turbine Is automatically tripped when the reactor is tripped, this trip provides a reliable means for providing protection to the turbine from excessive moisture carry over. This trip's setpoint does not correspond to a Safety Limit and no credit was taken in the safety analyses for operation of this trip. Its functional capability at the specified trip setting is required to enhance the overall reliability of the Reactor Protection System.

Reactor Coolant Flow - Low e {ORN 034)

The Reactor Coolant Flow - Low trip provides protection against a reactor coolant pump sheared shaft event and a steam line break event with a loss-of-offslte power. A trip Is initiated when the pressure differential across the primary side of either steam generator decreases below a nominal setpoint of 19.00 psid. The specified setpoint ensures that a reactor trip occurs to prevent violation of local power density or DNBR safety limits under the stated conditions.

4- (DrN 034)

WATERFORD - UNIT 3 B 2-6 AMENDMENT NO. 42, CHANGE NO. 4.-2e-

BASES 2.2.1 INSERT The CPCs contain several auxiliary trip functions which are credited in the safety analysis. These trips manifest themselves as DNBR trips however they are making the trip determination on parameters other than DNBR.

The CPC Variable Overpower Trip (VOPT) is provided to include a trip on power which is compensated for the decalibrating effects of changes in coolant temperature in the reactor vessel downcomer. Additionally, the trip setpoint is allowed to change with slow changes in plant power.

Thus at intermediate steady state powers, the plant is protected by a power trip which is a small distance above steady state power levels. The rate at which the automatic increases and decreases in the setpoint may change are limited and accounted for in the safety analysis.

The CPCs contain a trip which detects asymmetries in cold leg loop temperatures resulting from an asymmetric steam generator transient. The trip occurs if the cold leg asymmetry exceeds 11 "F.

The CPCs contain a trip monitoring margin to saturation conditions in the hot legs. A trip will be generated if margin to saturation is less than 13 "F.

The CPCs contain a direct trip on low RCP speed. This trip will occur if the RCP speed drops below 0.965.

BASES

  • -O( 03-24. Ch 23)

If a Surveillance is not completed within the allowed delay period, then the equipment Is considered inoperable or the variable is considered outside the specified limits and the allowed outage times of the required actions for the applicable LCO begin Immediately upon expiration of the delay period. If a Surveillance is failed within the delay period, then the equipment is inoperable, or the variable is outside the specified limits and the allowed outage times of the required actions for the applicable LCO begin immediately upon the failure of the Surveillance.

Satisfactory completion of the Surveillance within the delay period allowed by this Specification, or within the allowed outage time of the actions, restores compliance with Specdcatlon 4.0.1.

  • qDRN 03-24. Ch. 23)

Surveillance Requirements do not have to be performed on inoperable equipment because the ACTION requirements define the remedial measures that apply. However, the Surveillance Requirements have to be met to demonstrate that Inoperable equipment has been restored to OPERABLE status.

Specification 4.0.4 establishes the requirement that all applicable surveillance must be met before entry into an OPERATIONAL MODE or other condition of operation specified In the Applicability statement. The purpose of this specification Is to ensure that system and component OPERABILIlY requirements or parameter limits are met before entry into a MODE or condition for which these systems and components ensure safe operation of the facility. This provision applies to changes In OPERATIONAL MODES or other specified conditions associated with plant shutdown as well as startup.

Under the provisions of this specification, the applicable Surveillance Requirements must be performed within the specified surveillance Interval to ensure that the Uimiting Condition for Operation are met during Initial plant startup or following a plant outage.

When a shutdown Is required to comply with ACTION requirements, the provisions of Specification 4.0.4 do not apply because this would delay placing the facility in a lower MODE of operation.

- s~ot>-S *o?.cA.. o AMENDMENT NO. 99; WATERFORD - UNIT 3 B 3/4 0-7 CHANGE NO.-293;7--

Bases 314.0-1 Insert Table B 314.0-1 lists both the analytical and indicated values for various specifications.

Technical Specifications ensure that the plant is operated within the analyzed envelop for which analyses have demonstrated acceptable plant response to postulated events. In so doing, Technical Specifications provide operational restrictions and limitations on equipment. setpoints, and initial plant conditions assumed in the safety analysis. The Technical Specifications contain a mixture of analytical and indicated plant values based on various factors (e.g., safety setpoint, sensitivity of safety analysis to value, etc.) Choosing an indicated value or an analytical value for a limit depends on which value is judged most appropriate in each particular instance. For most initial conditions, the analytical value equals the indicated value. For more sensitive parameters, the indicated value varies from the analytical value to explicitly account for instrument inaccuracies and other process requirements thus providing increased assurance that expected automatic action occurs at or before the specified limit is reached. To highlight the association between indicated and analytical values, Table B314.0-1 presents both values for limits specified in the Technical Specifications.

mANALYTICAL-INDICATED VAiLUES

,Analytical Indicated I\

TS Parameter Value Plant Value Comments \

1.24 RATED THERMAL 3735 MWt 3716 MWt POWER _

Table #2 Linear Power 115% s 108%

2.2-1 Level High l Table #3 Logarithmic 4.4% s0.257%

2.2-1 Power Level High Table #4 Pressurizer 2422 psia s 2350 psia 2.2-1 Pressure High /

Table #5 Pressurizer 1560 psia 1684 psia 2.2-1 Pressure Low Table #6 Containment 19.7 psia s 17.1 psia 2.2-1 Pressure High .

Table #7 Steam Generator 576 pSia2 662 psia 2.2-1 Pressure Low Table #8 Steam Generator 5%NR Ž27.4%NR 2.2-1 Level Low Table #9 Local Power 21kWlft s21kW/ft 2.2-1 Density High Table #10 DNBR Low 1.26 21.26 2.2-1 Table #11 Steam Generator 90% (NR) sO7.7% (NR) 2.2-1 Level High Table #16 Reactor Coolant 60% flow 2 19.00 psid Analysis utilizes 2.2-1 Flow Low units in % flow while indications are in

____ ___psid 3.1.2.7 BAMT Boron  :!4551 ppm 24000 ppm Concentration 56187 ppm s6125 ppm 3.1.2.7 RWSP Volume NA 212% level Modes 5 &6 3.1.2.7 RWSP Boron 2029 ppm 22050 ppm Concentration 4.1.2.7 RAB Air Temperature 50'F 255'F 4.1.2.7 BAMT Solution 49'F 2607 Temperature 4.1.2.8 RAB Air Temperature Ž50'F Ž55'F 4.1.2.8 BAMT Solution 49-F 2 60'F Temperature WVATERFORD - UNIT 3 B 3/4 0-##

TABLE B 3/0-1 -s -

ANALYTICAL-INDICATED VALUES Analytical Indicated I TS Parameter Value Plant Value Comments 3.2.6 RCS Cold Leg 2533-F Ž536-F Temperature s552 F s549 F 3.2.8 Pressurizer Pressure 22090 psia Ž2125 psia s_2310 psia s2275 psia Table #1.b, 3.b & 4.c 19.7 psia s17.1 psia 3.3-4 Containment Pressure High Table #1.c & 3.c 1560 psia 21684 psia @NOP 3.3-4 Pressurizer Pressure Low Table #2.b Containment 19.7 psia S17.7 psia 3.3-4 Pressure High High Table #4.b & 7.e 576 psia 2662 psia @NOP 3.3-4 Steam Generator Pressure Low Table #5.b RWSP Level for 28,843 gal. Ž10% level Analysis utilizes 3.3-4 RAS. units Ingallons while indications are in %.

Table #6.a 4.16 kV 1E Bus 3245 volts Ž3245 volts 3.3-4 Undervoltage (Loss of Voltage)

Table #6.b 480 V 1E Bus 354 volts 2372 volts 3.3-4 Undervoltage Table #6.c 4.16 kV 1E Bus 3860 volts Ž3875 volts 3.3-4 Undervoltage (Degraded Voltage)

Table #7.b Steam 5% (NR) >27.4% (NR) 3.3-4 Gencrator Lcvel Low Table #7.c & 7.d 230 psid s123 psid 3.3-4 Steam Generator Delta Pressure High Table #7.g Control Valve 21.3% level Ž36.3% level 3.3-4 Logic (WR SG Level Low) 3.4.5.2c Primary - Secondary 75 gpd s75 gpd per SG Leakage I I Is2C 3.4.8.2

....AEIR Pressurizer Heatup &

Cooldown U Rate 200'F/hr

/ 0_#.

)O'F/hr I

_ ./ ANALYTCAL-INDICATEDIALE Analytical IndicatedI TS Parameter Value Plant Value Comments 1 3.5.1 SIT Borated Water 240%/926 ft3 244% level Analysis utilizes Volume s77.8%/1686 ft3 573.8% level Indications are in %.

3.5.1 SIT Boron 22029 ppm 22050 ppm Concentration 52929 ppm 52900 ppm 3.5.1 SIT Pressure 2570 psig 2600 psig 5700 psig 5670 psig \

3.5.4 RWSP Volume 383,000 gal 283% level Analysis utilizes units in gallons while Indications are In %.

Indicated on PMC 3.5.4 RWSP Boron 22029 ppm 22050 ppm Concentration s2929 ppm 52900 ppm/

3.5.4 RWSP Solution 250-F/255'F 255-F ECCS Analysis I Teff1purature 51lOO'F Si00OF Inadvertent Spray Cntmt 4.5.4 RAB Air Temperature 250-F 255'F s100F s100 F 3.6.1.4 Containment >14.275 psia >14.275 psia Pressure <27" w.g. <27" w.g.

3.6.1.5 Containment 290'F 290-F Average Air 5120-F 5120-F Temperature Table Maximum Allowable 87.3% 585.3% w/1 MSSV Inop 3.7-2 Power with MSSVs 69.9% 566.7% wl2 MSSVs Inop Inlopelable Table Linear Power Level - NA s93.3% Not aedited for loss 3.7-2 High Trip Setpoint NA s74.7% vu event.

3.7.1.3 CSP Volume 170,000 gal. Ž92% level Analysis utilizes units in gallons while

.indications

__ __ are in %.

3.7.1.3 CSP Water Ž50-F 255'F Temperature slOOF 5100°F 4.7.1.5 MSIV Closure Time 8.0 sec 4.0 sec Static test 4.7.1.6 MFIV Closure Time 6.0 sec 6.96 sec Static test with only one of two accumulators 3.7.1.7 ADV Setpoint 1040 psia 5992 psig Indicated on PMC WATERFORD) - UnNIT 3 B 3/4 0-4#

REACTIVITY CONTROL SYSTEMS BASES 3t4.1.2 BORATION SYSTEMS The boron Injection system ensures that negative reactivity control Is available during each mode of facility operation. The components required to perform this function include (1) s rl -

borated water sources. (2) charging pumps. (lIseparateflow paths. (4) boric acid makeup to I eetr an emergency power supply from 1-OPERABLE diesel ge erators.rii With the RCS average temperature above 200'F, a minimum of two separate and it redundant boron Injection systems are provided to ensure single functional capability in the a event an assumed failure renders one of the systems Inoperable. Allowable out-of-service periods ensure that minor component repair or corrective action may be completed without undue risk to overall facility safety from Injectio tem failures during the repair period. ' I Th hnratio is sufficient to provide a SHUTDOWN MARGIN O CL iofaOdelt k/ke rAVt 91 The maximum expected boraton capabi ty requirement occurs at EOL from power .

equilibrium xenon conditions assuming the most reactive CEA stuck out of the core and requires % ee-

' lboric acid solution from the boric acid makeu tanks In the allowable concentrations and volumes of Specification 3.1.2.8 plus allons of 2050 ppm borated water z (A specieo e consisten with Specification In order to meet thCCS r uiremen

-4(DRN M-375s Ch.19) .i  % ~c+c With the RCS temperature below 200°F one Injection system Is accep le without single failure consideration on the basis of the stable reactivity condition of the reactor and the additional restrictions prohibiting CORE ALTERATIONS and positive reactivity changes in the event the single injection system becomes Inoperable. Temperature changes In the RCS impose reactivity changes by means of the moderator temperature coefficient. Plant temperature changes are allowed provided the temperature change is accounted for in the calculated SDM. This will require a new SDM calculation be performed If the current SDM calculation does not bound the temperature change. Small changes In RCp temperature are unavoidable and so long as the required SDM Is maintained during these changes, any positive reactivity additions will be limited to acceptable levels. Introduction of temperature changes must be evaluated to ensure they do not result in a loss of required SDM.

4-(DRN 03-375. Ch 19)

The boron capability required below 200°F is based upon providing a 2% delta k/k SHUTDOWN MAIN after xenon decay and cooldown from 200'F to 140'F. This condition requires eithe , allons of 2050 ppm borated water from the refueling water storage pool or boric acid solution fr the boric acid makeup tanks in accordance with the requirements of Specification 3.1.2.7.

AMENDMENT NO. 10, 129 WATERFORD - UNIT

..... _ _ . _ ...... _ 3_ B 3/4 1-2 CHANGE NO. 49-

REACTIVITY CONTROL SYSTEMS BASES BORATION SYSTEMS (Continued)

The contained water volume limits include allowance for water not available because of discharge line location instrumen lerances, and other physical characteristics.

The OPERABILITY of one boron injection system during REFUELING ensures that this system is available for reactivity control while In MODE 6.

The lower limit on the contained water volume, the specified boron concentration, and the physical size (approximately 600,000 gallons) of the RWSP also ensure a pH value of between 7.0 and 11.0 for the solution recirculated within containment after a LOCA. This pH band minimizes the evolution of iodine and minimizes the effect of chloride and caustic stress corrosion on mechanical systems and components.

Tus ist spefontatnme ispressure topreendt es dliaccibr condto\

ltknsrgn valid fnid avoids tb osb1ti o cont imnt o Sprssr \

iathXt~x111ton the RWS8ReP ertr rqu1 to prgSn feesz a 0orbtnpetftation in teRS._~

3/4.1.2.9 BORON DILUTIO This specification is provided to revent a boron dilution event, and to prevent a loss of SHUTDOWN MARGIN should an inadvertent boron dilution event occur. Due to boron concentration requirements for the RWSP and boric acid makeup tanks, the only possible boron dilution that would remain undetected by the operator occurs from the primary makeup water through the CYCS system.

Isolating this potential dilution path or the OPERABILITY of the startup channel high neutron flux alarms, which alert the operator with sufficient time available to take corrective action, ensures that no loss of SHUTDOWN MARGIN and unanticipated criticality occur.

The ACTION requirements specified in the event startup channel high neutron flux alarms are inoperable provide an alternate means to detect boron dilution by monitoring the RCS boron concentration to detect any changes. The frequencies specified In the COLR provide the operator sufficient time to recognize a decrease in boron concentration and take appropriate corrective action without loss of SHUTDOWN MARGIN. More frequent checks are required with more charging pumps in operation due to the higher potential boron dilution rate.

The surveillance requirements specified provide assurance that the startup channel high neutron flux alarms remain OPERABLE and that required valve and electrical lineups remain in effect.

3/4.1.3 MOVABLE CONTROL ASSEMBLIES The specifications of this section ensure that (1)acceptable power distribution limits are maintained, (2)the minimum SHUTDOWN MARGIN is WATERFORD - UNIT 3 B 3/4 1-3 AMENDMENT No. OT"2-

Bases 314.1.2 Insert The unusable water volume in one Boric Acid Makeup Tank is half the unusable water volume when using two Boric Acid Makeup Tanks. Consequently, Figures 3.1-1 and 3.1-2 are provided for using one or two Boric Acid Makeup Tanks to satisfy the requirements of TS 3.1.2.2 and 3.1.2.8.

The 60 'F minimum Boric Acid Makeup Tank solution indicated temperature limit insures that the boron will not precipitate even at the maximum allowed boron concentration when instrument accuracies arc considered. The precipitation temperature at the maximum allowed Boric Acid Makeup Tank boron concentration is 50.2 F. The 60 F minimum indicated temperature limit also insures that the minimum Boric Acid Makeup Tank solution temperature assumed in the safety analysis (49 'F) is bounded. The 55 'F Reactor Auxiliary Building temperature prerequisite for monitoring Boric Acid Makeup Tank solution temperature is acceptable due to the increased accuracy of the Reactor Auxiliary Building temperature indications available on the plant monitoring computer.

POWER DISTRIBUTION LIMITS BASES DNBR MARGIN (Continued)

A DNBR penalty factor has been included in the COLSS and CPC DNBR calculations to accommodate the effects of rod bow. The amount of rod bow in each assembly is dependent upon the average bumup experienced by that assembly. Fuel assemblies that incur higher average burnup will experience a greater magnitude of rod bow. Conversely, lower bumup assemblies will experience less rod bow. In design calculations, the penalty for each batch required to compesatle rur rud bow is determinred from a batchs maximum average assembly bumup applied to the batch's maximum integrated planar-radial power peak. A single net penalty for COLSS and CPC is then determined from the penalties associated with each batch, accounting for the offsetting margins due to the lower radial power peaks Inthe higher bumup batches.

3/4.2.5 RCS FLOW RATE This specification is provided to ensure that the actual RCS total flow rate is maintained at or above the minimum value used in the LOCA safety analyses, and that the DNBR is maintained within the safety limit for AntIcIpated Operational Occurrences (AOO).

314.2.6 REACTOR COOLANT COLD LEG TEMPERATURE This specification is provided to ensure that the actual value of reactor coolant cold leg temperature is maintained within the raeof values used In the safety analyses, with adjustment for instrument accuracy of and that the peak linear heat generation rate and the moderator temperature coefficienXe~cts are validated. 4 314.2.7 AXIAL SHAPE INDEX _ -

e (DRN 02458)

This specification is provided to ensure that the actual value of AXIAL SHAPE INDEX is maintained within the range of voluos used In the safety analyses, to ensure that the peak fuel centerline temperature and DNBR remain within the safety limits for Anticipated Operational Occurrences (AOO).

4- CDRN 02.458) 3142.8 PRESSURIZER PRESSURE n cr4 This specification Is provided to ensure that the actual value of pressuri r pressure Is maintained within the range of values used in the safety analyses. The in o CPCs and COLSS are the most limiting. The values are adjusted for an instrumentof +/- si.

(Bysej 3/vA.Z,S U per CHANGE NO.4K-WATERFORD - UNIT 3 B 3/4 2-4 AMENDMENT NO. +2,

Bases 314.2.6 Insert The safety analysis assumes that cold leg temperature is maintained between 533XF and 552-F or indicated temperatures of 536-F and 549F.

Bases 314.2.8 Insert The safety analysis assumes fiat pressurizer pressure Is maintained between 2090 psia and 2310 psia or indicated pressurizer pressures of 2125 psia and 2275 psia.

REACTOR COOLANT SYSTEM STEAM GENERATORS (Continuedl based on a modification of Regulatory Guide 1.83, Revision 1. Inservice inspection of steam generator tubing is essential Inorder to maintain surveillance of the conditions of the tubes in the event that there is evidence of mechanical damage or progressive degradation due to design, manufacturing errors, or inservice conditions that lead to corrosion.

Inservice inspection of steam generator tubing also provides a means of characterizing the nature and cause of any tube degradation so that corrective measures can be taken. ,,6 T The plant is expected to be pera at the secondary coolant will be maintained withl those chemistry limits found to result in negligible corrosion of the ste generator tubes. If the secondary coolant.

chemistry is not maintained wit in these limits, localized corrosion may likely result in stress corrosi n cracking. The extent of cracking during plant operation would be liit by the limitation of steam generator tube eakage between the primary coo s ten and the secondary coolant system (primary-to-socondary leakage a per steam generator). Cracks having a primary-to-secondary leakage less than this limit during operation will have an adequate margin of safety to withstand the loads imposed during normal operation and by postulated accidents 0 erating plants have demonstrated that primary-to-secondary leakage of per steam generator can readily be detect rad ation monitors o team generator blowdown. Leakage In excess of w11l require p ant shutdown and an unscheduled inspection Vduring which the leakag tubes wl-be located ad lugged or -

repal lastage-type defects are unlikely with proper chemistry treatment of the se ndary coolant. However, even if a defect should develop in service, it I1 be found during scheduled inservice steam generator tube examinations.

lugging or sleeving will be required for all tubes with iperfections exceeding the plugging or repair limit as defined in Surveillance Requirement 4.4.4.4. Defective tubes may be repaired by sleeving in accordance with CENS Report CEN-605-P 'Waterford 31 team Generator Tube Repair Using Leak Tight Sleeves, Revision 00-P, dated December 1992. Steam generator tube inspections of operating plants have demonstrated the capability to reliably detect degradation that has penetrated 20% of the original tube wall thickness. Sleeved tubes will be included in the periodic tube inspections for the inservice Inspection program.

Whenever the results of any steam generator tubing inservice inspection fall into Category C-3, these results will be promptly reported to the Comission pursuant to Specification 6.9.1 prior the resumption of plant operation. Such cases will be considered by the Comiission on a case-by-case basis and may result in a requirement for analysis, laboratory examinations, tests, additional eddy-current inspection, and revision of the Technical Specifications, if necessary.

ATER D - UNIT 3 3/erjlpe7 Aln.n No.;

WRATERFORD - UNIT 3 B 3l4 4;1'f Amendment No. 2-17

REACTOR COOLANT SYSTEM BASES 3/4.4.5 REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.5.1 LEAKAGE DETECTION SYSTEMS The RCS leakage detection systems required by this specification are provided to monitor and detect leakage from the Reactor Coolant Pressure Boundary. These detection systems are consistent with the recommendations of Regulatory Guide 1.45, "Reactor Coolant Pressure Boundary Leakage Detection Systems," May 1973.

If one of the required systems becomes inoperable, 30 days are permitted for restoration since two diverse and redundant RCS leakage detection systems remain OPERABLE. If, however, the inoperable system is the containment gaseous or particulate monitoring system, grab samples are also performed as a backup to the single remaining atmospheric monitoring system.

3/4.4.5.2 OPERATIONAL LEAKAGE Industry experience has shown that while a limited amount of leakage is expected from the RCS, the unidentified portion of this leakage can be reduced to a threshold value of less than 1 gpm. This threshold value is sufficiently low to ensure early detection of additional leakage.

The 10 gpm IDENTIFIED LEAKAGE limitation provides allowances for.a limited amount of leakage from known sources whose presence will not interfere with the detection of UNIDENTIFIED LEAKAGE by the leakage detection systems.

The Surveillance Requirements for RCS pressure isolation valves provide added assurance of valve integrity thereby reducing the probability of gross valve failure and consequent intersystem LOCA. Leakage from the RCS pressure isolation valves is IDENTIFIED LEAKAGE and will be considered as a portion of the allowable limit.

hrttal ste sgenera o~ub leakg/5t 0of 1 y~ foI ~ r~Pa

/eeitrs ensure that the osage cont f ut on from he tbe leaka:e wil be\

eted i to a rctin ail of Part, I r'imits in the vnt of eiter a sts' tgnrator tf~ rpuev~ steam lin treak. Tejgolimit scnitt' I with the sumptions sad in the alysis of se accident . The 0.5 f leaa*8im~ pe Stam eneratg'C2 gal/tg ensures tb>6t steam gerao tuv~tgrt ixinained rftevent jra main steped line rupWe or n Ice conditions6 -- '-q..Z

, J-2~---e--rt -

PRESSURE BOUNDARY LEAKAGE of any magnitude is unacceptable since it may be indicative of an impending gross failure of the pressure boundary.

Therefore, the presence of any PRESSURE BOUNDARY LEAKAGE requires the unit to be promptly placed in COLD SHUTDOWN.

3/4.4.6 CHEMISTRY The limitations on Reactor Coolant System chemistry ensure that corrosion of the Reactor Coolant System is minimized and reduces the potential for Reactor Coolant System leakage or failure due to stress corrosion. Maintaining WATERFORD - UNIT 3 a 3/4 4-4

BASES 314.4.5.2 INSERT The 75 gallon per day (gpd) per steam generator tube leakage limit ensures that the radiological consequences, including that from tube leakage, will be limited to the Part 100 limits for offsite dose and within the limits of General Design Criterion 19 for control room dose. For those analyzed events that do not result in faulted steam generators, greater than or equal to 75 gpd primary-to-secondary leakage per steam generator is assumed in the analysis. For those analyzed events that result in a faulted steam generator (e.g., MSLB),

540 gpd primary-to-secondary leakage is assumed through the faulted steam generator while greater than or equal to 75 gpd primary-to-secondary leakage is assumed through the intact steam generator.

Steam generator tube cracks having primary-to-secondary leakage less than 150 gpd per steam generator during operation will have an acceptable margin of safety to withstand loads imposed during normal operation and postulated accidents (Reference NEI 97-06).

Due to the proximity of the east atmospheric dump valve to the east control room intake, the primary-to-secondary leakage limit required to achieve acceptable radiological consequences, for accidents that rely on reactor coolant system cooldown using the steam generators, is limiting. Therefore, 75 gpd per steam generator is imposed as the primary-to-secondary operational leakage limit.

REACTOR COOLANT SYSTE BASES PRESSURETE1MPERATURE LIMITS (Continued)

The maximum RT for all Reactor Coolant System pressure-retaining materials. with the exception of the reactor pressure vessel. has been determined to be 90'F. The Lowest Service Temperature limit line shown on Figures 3.4-2 and 3.4-3 is based upon this RT since Article NB-2332 of Section III of the ASHE' Boiler and Pressure 4ssel Code requires the Lowest Service Temperature to be RTNT + 100lF for piping, pumps, and valves. Below this temperature, the system pressure must be limited to a maximum of 20X of the system's hydrostatic test pressure of 3125 psia (as corrected for elevation and instrument error).

ITheiiitatlonsAgoiosed on thep ressurlzer heatup and cooldown rates w are provided to assure that the withn the design criteria assumed for the fatigue analysis performed in accordance with the ASHE 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 2720F. 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 100lF 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 2O0 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 2720F, 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 6 to 10 CFR Part 50. The RCS will be protected against overpressure transients and will not exceed the limits of Appendix 6 by restricting starting of the RCPs to when the secondary water temperature of each steam generator is less than lOOF above each of the RCS cold leg temperatures. Maintaining the steam generator less than IOO F 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 4-10 Amendment Ho. ;2,4-

Cv 4-caeo 4 4br.4'e 4 H CVA rtm' bt EMERENCY CORE COOLING SnSTEMS BASES 314.5.4 REFUELING WATER STORAGE POOL (RWSP)

The OPERABILITY of the refueling water storage pool (RWSP) as part of the ECCS also ensures that a sufficient supply of borated water Is available for injection by the ECCS in the event of a LOCA. The limits on RWSP minimum volume and boron concentration ensure that (1) sufficient water Is available within containment to permit recirculation cooling flow to the core, and (2) the reactor will remain aubftical in the cold condition following mixing of the RWSP and the RCS water volumes with all CEAs inserted except for the most reactive control assembly. These assumptions are consiste ter volume limittincludes atrlowanco for ot usable because of pool discharge line locahyier physical characteristic The lower limit on contained water volume, the specific boron concentration and the physical size (approximately 600.000 gallons) of the RWSP also ensure a pH value of between 7.0 and 11.0 for the solution recirculated within containment after a LOCA. This pH band minimizes the evolution of Iodine and minimizes the effect of chloride and caustic stress corrosion on mechanical systems and components.

The maximum limit on the RWSP temperature ensures that the assumptions used in the containment pressure analysis under design base accident conditions remain valid and avoids the possibility of containment overpressure. The minimum limit on the RWSP temperature is required to prevent freezing andlor boron precipitation in the RWSP.

faioet qi5, ~a0 OAc' Jo arts"eel m-evi4 vxHerr4,Q',Y-Y. NO ,t2eeS l h, t.S&Mr h, C X flue le Vole -1y CO Cn , YA_ A -

WATERFORD - UNIT 3 B 3/4 5-3 AMENDMENT NO. 42q, 43(Y

CONTAINMENT $SYSEEMS BASES 3/4.6.1.4 INTERNAL PRESSURE The limitations on containment internal pressure ensure that (1)the containment structure is prevented from exceeding its design negative pressure differential with respect to the annulus atmosphere of 0.65 psid, (2)the containment peak pressure does not exceed the design pressure of 44 psig during either LOCA or steam line break conditions, and (3)the minimum pressure of the ECCS performance analysis (BTP CSB 61) is satisfied.

The limit of +27 inches water (approximately 1.0 psig) for initial positive containment pressure is consistent with the limiting containment pressure and temperature response analyses inputs and assumptions.

The limit of 14275 psia for initial negative containment pressure ensures that the minimum containment pressure isconsistent with the ECCS performance analysis ensuring core reflood under LOCA conditions, thus ensuring peak cladding temperature and cladding oxidation remain within Emits. The 14275 psia limit also ensures the containment pressure will not exceed the containment design negative pressure differential with respect to the annulus atmosphere in the event of an Inadvertent actuation of the containment spray system.

314.6.1.5 AIR TEMPERATURE es The limit of 1200 Fon high average containment temperature is consistent with the limiting containment pressure and temperature response analyses inputs and assumptions.

The fimits currently adopted by Waterford 3 are 269.30 F during LOCA conditions and 413.50 F durying MSLB3 conditions. r The4 12OF maximum'alu ,pecifled Inthe TS the valu ed Inthe accident analysis 3/4.6.1.6 CONTAINMENT VESSEL STRUCTURAL INTEGRITY This limitation ensures that the structural Integrity of the containment steel vessel will be maintained comparable to the original design standards for the life of the facility. Structural integrity Is required to ensure that the containment vessel will withstand the maximum pressure resulting from the design basis LOCA and main steam line break accident. A visual Inspection in conjunction with Type A leakage test is sufficient to demonstrate this capability.

314.6.1.7 CONTAINMENTVENTILATION SYSTEM The use of the containment purge valves is restricted to 90 hours0.00104 days <br />0.025 hours <br />1.488095e-4 weeks <br />3.4245e-5 months <br /> per year in accordance with Standard Review Plan 6.2.4 for plants with the Safety Evaluation Report for the Construction License issued prior to July 1. 1975. The purge valves have been modified to limit the opening to approximately 52 to ensure the valves will close during a LOCA or MSLB;and therefore. the SITE BOUNDARY doses are maintained within the guidelines of 10 CFR Part 100.

The purge valves, as modified, comply with all provisions of BTP CSB 64 except for the recommended size of the purge line for systems to be used during plant operation.

WATERFORD - UNIT 3 B 3/4 6-2 AMENDMENT NO.-27 CHANGE NO. 2B,4;-

Bases 314.6.1.5 Insert The limitation on containment minimum average air temperature ensures that the ECCS is capable of maintaining a peak clad temperature (PCT) less than or equal to 2200°F under LOCA conditions. A lower containment average air temperature results in a lower post accident containment pressure, a lower reflood rate, and therefore a higher PCT. The containment minimum average air temperature limit is only applicable above 70% rated thermal power. At power levels of 70% or below and a containment minimum average air temperature of less than 900F, ECCS is capable of maintaining the peak clad temperature (PCT) less than or equal to 2200'F under LOCA conditions

3/4.7 PLANT SYSTEMS BASES 3/4.7.1 TURBINE CYCLE 314.7.1.1 SAFETY VALVES The OPERABILITY of the main steam line code afety valves ensures that the se nda system pressure will be limited to within 11Oi4g4pf~of its design pressure of uring the most severe anticipated system operationaT ransient. The maximum c relieving capacity is associated with a turbine trip from 100% RATED THERMAL POWER

' coincident with an assumed loss of condenser heat sink (i.e., no steam bypass to the

\/ condenser).

The specified valve lift settings and relieving capacities are in accordance with the requirements of Section III of the ASME Boiler and Pressure Vessel Code, 1974 Edition. The MSSVs rated capacity passes the full steam flow at 102% RATED THERMAL POWER (100% -

2% for instrument error) with valves open. A minimum of 2 OPERABLE safety valves per steam generator ensures that sufficient relieving capacity I6available for removing decay heat.

All the MSSVs have an orifice size of 28.27 Wn.

fiwitinfh~atiowtheAgh~lN~qurern padTAQlifqibe usefa o~n basis~ofG the re~dfction),secoway l u er P~oz Le ^QhS 9Tuctieons are derived on th folwng le n ofa lssof ondnse vcuu evA4initiated at the reduced power1gylls listed in Table 3.7-2 that shows peak steam generator pr~ue arQrQnand eo 10~-

WATERFORD - UNIT 3 B 3/47-1 AMENDMENT NOR.-t-

Bases 314.7.1.1 Insert #1 An alternative to restoring the inoperable MSSV(s) to OPERABLE status is to reduce THERMAL POWER so that the available MSSV relieving capacity meets Code requirements for the power level. Operation may continue provided the allowable THERMAL POWER is less than or equal to the maximum allowable power as listed in Table 3.7-2 and the Linear Power Level - High trip setpoint is less than or equal to that listed in Table 3.7-2.

The 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> completion time is a reasonable time period to reduce power level and is based on the low probability of an ovent occurring during this period that would require activation of the MSSVs. An additional 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> is allowed to reduce the setpoints. This completion time is based on the time required to perform the power reduction, operating experience in resetting all channels of a protective function, and on the low probability of the occurrence of a transient that could result in steam generator overpressure during this period.

Bases 314.7.1.1 Insert #2 110% of system design pressure. The maximum allowable power limitations listed in Table 3.7-2 are reduced from the analytical values used in the analysis by at least 2% to account for power measurement uncertainties.

The reactor trip setpoint reductions are determined by adding 8% to the maximum THERMAL POWER limit derived from the analysis ot condenser vacuum event. The 8%

margin is consistent with margin between the normal Linear Power Level - High trip setpoint and 100% RATED THERMAL POWER. The 8% difference provides sufficient margin to avoid an inadvertent trip. The Linear Power Level - High trip Is not credited Inthe analysis of the loss of condenser vacuum event but is reduced to reinforce the requirement to remain at the reduced power levels for extended periods of time.

PLANT SYSTEMS BASES 3/4.7.1.2 EMERGENCY FEEDWATER SYSTEM (Continued)

Surveillance Reauirements (Continued)

This SR is modified to indicate that the SR should be deferred until suitable test conditions have been established. This deferral is required because there is an insufficient steam pressure to perform post maintenance activities which may need to be completed prior to performing the required turbine-driven pump SR. This deferral allows the unit to transition from MODE 4 to MODE 3 prior to the performance of the SR and provides a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period once a steam generator pressure of 750 psig is reached to complete the required post maintenance activities and SR. If this SR is not completed within the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period or fails, then the appropriate ACTION must be entered. The twenty-five percent grace period allowed by TS 4.0.2 can not be applied to the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period.

d. The SR for flow testing ensures that the EFW system is aligned properly by verifying the flow paths from the condensate storage pool (CSP) to each steam generator before entering MODE 2 operation after being in MODE 4. 5. 6. or defueled. for 30 days or longer, or whenever feedwater line cleaning through the emergency feedwater line has been performed. Various combinations of pumps and valves may be used such that all flow paths (and flow legs) are tested at least once during the Surveillance.

OPERABILITY of EFW flow paths must be verified before sufficient core heat is generated that would require the operation of the EFW System during a subsequent shutdown. The frequency is reasonable, based on engineering judgment, and other administrative controls to ensure that flow paths remain OPERABLE. To further ensure EFW system alignment, the OPERABILITY of the flow paths is verified following extended outages to determine that no misalignment of valves has occurred. This SR ensures that the flow paths from the CSP to the steam generators are properly aligned.

314.7.1.3 CONDENSATE STORAGE POOL The OPERABILITY of the condensate storage pool (CSP) with the minimum water volume of 173,500 gallons (170,000 gallons for EFW system usage and 3,500 gallons for CCW makeup system usage), plus makeup from one Wet Cooling Tower (WCT) basin, ensures that sufficient water is available to cool the Reactor Coolant System to shutdown cooling entry conditions following any design basis accident. This makeup water includes the capability to maintain HOT STANDBY for at least an additional 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> prior to initiating shutdown cooi The combined capacity (CSP and one WCT) provides sufficient cooling for until shutdown cooling is initiated in the event the ultimate heat sink sustains tornado damage concurrent with the tornado event.

WATERFORD - UNIT 3 B 3/4 7-2e CHANGE NO.+-7

PLANT SYSTEMS BASESP 3/4.7.1.3 CONDENSATE STORAGE POOL (Continued)

If natural circulation is required, the combined capacity (CSP and one \NCT) is sufficient to maintain the plant at HOT STANDBY for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, followed by a cooldown to shutdown cooling entry conditions assuming the availability of only onsite or only offsite power, and the worst single failure (loa;of a,diesel generator or atmospheric dump valve). This requires a roximatelyallons of EFVV and complies ith BTP RSB 5-1.

Zo Co,Coozew92

-hThe CSP contained water volume limit m indicated in MODES 1, 2, and 3) includes an allowance for water not usable because of vortexing and instrumentation uncertainties. This provides an assurance that a minimum of 170,000 gallons is available for the EFW system and that 3,500 gallons is available for the CCW makeup system. The CSP contained water volume limit (11% indicated in MODE 4) also includes an allowance for water not usable because of vortexing and instrumentation uncertainties. This provides an assurance that minimum of 3,500 gallons is available in the CSP for the CCW makeup system.

7-l gJ;mI'm vl-~t CsP +eosfoer4,1-t tre eisl S (1o~a# the r5%i7pior1: 4'Lse' J 't ore5iyP busees l n P teenperc 9s-ir re eni5c4v-eS tp9-lie f + Be cased ini .fhe A ~tf s re+"ro +o poclet evevtn7s rerta;>

WATERFORD - UNIT 3 B 3/4 7-2f CHANGE NO.4-

PLANT SYSTEMS BASES 3/4.7.1.4 ACTIVITY 340 s41 The limitations on secondary system specific activity ensure that the esultant offsite radiation dose will be limited to a small fraction of 10 CFR Part 100 rim' tsJ e event of a steam line rupture. This dose also Includes the effects of a coincident Primary to secondary tube leak in the steam generator of the affected steam line ;iaconcurrent loss-of-offsite electrical power. These values are consistent with the assumptions used In the safety analyses.

-9WnRN-73.O 31j 314.7.1.5 MAIN STEAM LINE ISOLATION VALVE (MSIVI The MS lVs isolate steam flow from the secondary side of the steam generators following a high energy line break. MSIV closure terminates flow from the unaffected (intact) steam generator.

One MS(V is located in each main steam line outside of, but close to, containment. The MSiVs are downstream from the main steam safety valves (MSSVs), atmospheric dump valves, and emergency feedwater pump turbine steam supplies to prevent their being Isolated from the steam generators by MSIV closure. Closing the MSIVs Isolates each steam generator from the other, and isolates the turbine, Steam Bypass System, and other auxiliary steam supplies from the steam generators.

The MSIVs close on a main steam Isolation signal (MSIS) generated by either low steam generator pressure or high containment pressure. The MSIVs fail as ison loss of power to the actuator however; the operators for the MSIV are furnished with redundant hydraulic fluid dump valves powered by diverse power, to ensure that no single electrical failure will prevent valve closure. The MSIVs may also be actuated manually.

A description of the MSIVa is found in Final Safety Analysis Repad (FSAR), Section 10.3.

The design basis of the MSiVs is established by the containment analysis for the large steam line break (SLB) Inside containment as discussed In FSAR. Section 6.2. It is also influenced by the accident analysis of the SLB events presented In FSAR, Section 15.1.3. The design precludes the blowdown of more than one steam generator, assuming a single active component failure (e.g., the failure of one MSIV to close on demand).

The OPERABILITY of the MSIVs ensures that no more than one steam generator will blow down in the event of a steam line rupture. This restriction is required to (1) minimize the positive reactivity effects of the Reactor Coolant System cooldown associated with the blowdown, and (2) limit the pressure rise within containment in the event the steam line rupture occurs within containment.

The MSIVs satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).

H4ow 0 Ot 31) t-1737.

AMENDMENT NO. 6-46 ,

WATERFORD - UNIT 3 B 3/4 7-3 CHANGE NO.-3r91

PLANT SYSTEMS BASE S \

O0R0 031737. CM 31f 314.7.1.5- MAIN S-TEAM LINE ISOLATiO0 V w MI (ntnuedi)

SR 4.7.1.5a verifiers that the ci§re U of each iiIV Is within Its limit when tested pursuant to the Inservice Testing P ram. tatic te sing 4.0 seconds demonstrates the ability of the MSIVs to close in les than or equal to th econds required closure time under This test dmay berfconduct InMD 3, wormal the ut edduring refueling a outage n ut may T perfoomed uponbereturninnthe unit to operation following a refueling outage. The MSIVs should not be tested at power since even a part stroknths.rTheincreases the risk of a valve closure with the unit generating poler. As the MSIVs are not tested at powern they cro mapompt from the ASME Code, Section li (Ins rvice Inspei son Art main4feedwreremeoto icle thring This test may be conducted In MODE 3, with the unit at operating temperature and pressure.

SR 4.7.1t.ib verifies that each MSIV can close on an actual or simulated actuation signal.

This Surveillance may be performed upon retumivng the plant to operadtion following a refueling outage. The Frequency of MStIV testing is every 18 months. The 18 month frequency for testing is based on the refueling cycle. Operating experience has shown that these components usually pass the Surveillance. Therefore, this Frequency is acceptable from a reliability standpoint.

314,7.1.6 MvAIN FEEDWATER ISOt ATIQN VALVES The Main Feedwater Isolation Valves (MFIVs) isolate mnain feedwater (MFW) flow to the secondary side of the steam generato following a high energy line break (nELBta Closure of the MFiVs terminates flow to both stream generytfaters cotaing the consequences for feedwater line breaks (FWsBs). Closure of the MFIVs effectively terminates addwteon of main Fhe feedwater to an affected steam generator, liming the mass and energy release for Main Steam Une Breaks (MSlBs) or eWLas inside containment. and reducing the coodown effects for MSiBs.

The MFIVs isolate the non-safety related feedwater supply from the safety related portion of the system., In the event of a secondary aide pipe rupture inside containment, the valves limit the quantity of high energy fluid that enters containment through the break, and provide a pressure boundary for the controlled addition of Emergency Feedwater (EFW) to the intact steam generator.

One MFIV is located on each MFW line, outside, bdut close to, containment. The MFIVS are located upstream of the EFW inction mtso that FW ma be sulied to a steam generator f flowinMIdourn pirlVolm eya o tq a gpror

/ifi d~~~~~

foCw~lan s and erryel ndrf pir E w t ng tty~team gpatr ey#getelSBoL AMENDMENT NO. 6-45 WATERFORD - UNIT 3 B 314 7-3c CHANGE NO. 15;31

bson d d MFsshIc pwe fAt sincc ev rtilsrk PLANT SYSTEMS t f a ve closnew thelo a n lcretes BASES==

314.7.1.6 MAIN FEEDWATER ISOLATION VALVES (con't)\

The TS isannotated with a 3.0.4 exemptionh allowing entryinto the applicable MODES to be made with an inoperable MFIV losed or isolated as required by the ACTiONS. Thenr ACTIONS allow separate condition entry for each valve by using aeh one or more MFiV...n .

Thi prevents s immedia s erforme0.3 if bothMFIn s are decs n d inopera T ble.

The Surveillanc eqIrmn erity isolation in ls lo eqatscnds i bared on the time assumred i aentand containment nlsar demonstrates fhe ability ofi MFIV to sclose I lessn than ow to e secn Tsting design basis accident ditl T (Re ) nheulirt b eed ater se P n a parHial stroke exercise Increi es arereitof a valve closure with the plant generating power anti would create added cyduic resses.nihe Surveillance to veriny each MFiV can dose on an actual or simulated actuation signal is nCharley perfomed when the plant Is returning to operation follow In a reAeling outage. Adifcation of valve closure on an actuation signal Is not required until entry Into Mode 3 conR Staf with TS 3.3.2. The 18 month frequency Is based on the refueling cycle.

Verification ofasure ti is at perfemed per the Inservice Testing Program. This frequencyis acceptable from a in lit T inture r s thcodiffe with the Inservioe Testing Program.

cosure,tChu allodin rapid valc osure Therefr t T fea st beabluet ea I Credited Non-Safel _ mrI_

Reactor Trip Overridn (RTO) and the Auxiliary Fopdwater (AFO) Pump High DiTocharf Pressure Trip (HDPT) are credited for rapid dosure of the Main Feedwater Isolation Valves (MFiVs) during main steam and feedwater line breaks. Crediting of these non-safety features was submitted to the NRC as a USQ and approved. (Reference letter dated September 5, 2000 from the NRC to Charles M.Dugger, 'Waterford 3 Steam Electric Station, Unit 3 - Issuance of Amendment RE: AddiOIon of Main Feedwater l tslahon Valves to Technical Specifications and Request for NRC Staff Review of an Unreviewed Sefety Question.')

The feature of RTO that Is credited for MFiV closure is the rapid SGFP speed reduction upon reactor trip ini3anon. This feature reduces the differential pressure across the valve disc at closure. thus allowing rapid valve closure. Therefore. the RTO feature must be able to dceaae G SGFP speed to minimum on a reactor tnip during SGFP operation for OPERABILITY of the MFIVs.

The AFW Pump HDPT reduces the differential pressure across thre valve disc at closure during AFW Pump operation. Therefore, this feature must be functional during AFW Pump operation for OPERABILITY of the MFIVs. When the AFW pump is not running, this trip Is not required.

In MODES 1,2, 3, and 4, the MFIVs are required to be OPERABLE. Because the MFIVs are required to be OPERABLE in MODES 1,2, 3, and 4, RTO must be able to decrease SGFP wDM 02-166. cO IS)

."t 0,IJ.

N CA 311 WATERFORD - UNIT 3 B 3/4 7-3e AMENDMENT NO. 6,167,

  1. q0M a3 tI. Cft2I CHANGE NO. X5, 38, &1

BASESS 3/4.7.1.7 ATMOSPHERIC DUMP VALVES (ADVs)

Two ADVs are provided, one per steam generator. The ADVs are provided with upstream block valves to permit their being tested at power, and to provide an alternate means of isolation. The ADVs are equipped with pneumatic controllers to permit control of the cooldown rate. The ADVs are provided with a pressurized nitrogen gas supply that, on a loss of pressure in the normal instrument air supply, automatically supplies nitrogen to operate the ADVs. The ADVs can also be operated manually once the nitrogen gas supply is depleted.

The ADVs provide a safety grade method for cooling the unit to Shutdown Cooling (SDC) System entry conditions, should the preferred heat sink via the Steam Bypass System to the condenser not be available, as discussed in the FSAR, Section 10.3.

This is done in conjunction with the Emergency Feedwater System providing cooling water from the condensate storage pool (CSP) to meet Branch Technical Position (BTP)

RSB 5-1.

The automatic operation of the ADVs to open is assumed in the Small Break LOCA (SBLOCA) analysis at power levels above 70% RATED THERMAL POWER. ADV's are credited for SBLOCA analysis to lower steam generator secondary side pressures, compared to crediting only MSSVs, and thus provide increased cooling of the RCS.

This results in a lower calculated peak cladding temperature (PCT) for SBLOCA ECCS analysis.

Analysis has shown that automatic operation of the ADV is not required when the unit is at or below 70% RATED THERMAL POWER for greater than six hours because, based on decay heat load, one high-pressure safety injection train is capable of mitigating the SBLOCA event. At greater than 70% RATED THERMAL POWER, one high-pressure safety injection train and one ADV, in automatic, are capable of mitigating the SBLOCA event. Therefore, the ADVs, in automatic, are required at greater than 70% RATED THERMAL POWER and for six hours after reducing power to less than or equal to 70%

RATED THERMAL POWER.

Lirniting Condition for Operation The LCO requires that each ADV be OPERABLE.

The ADV manual controls must be OPERABLE in MODES 1, 2, 3, and 4 to allow operator action needed for decay heat removal and safe shutdown in accordance with BTP RSB 5-1.

The LCO is modified by a footnote requiring that ADV automatic actuation controls be OPERABLE (i.e., ADVs Inautomatic and capable of automatic actuation at less than or equal to 1040 psia ( 992 psig indicated)) when operating at greater than 70% RATED THERMAL POWER and for six hours after reducing power to less than or equal to 70%

RATED THERMAL POWER for mitigation of the SBLOCA.

WATERFORD - UNIT 3 83/4 7- ##

PLN SYTM Z.aS Ts ~e~tat BASES ~

314.7.1.7 ATMOSPHERIC DUMP VALE/ES (Continued)

The ADVs are containment isolation valves and must be capable of manual isolation of the ADV lines in MODE 1, 2, 3, and 4 in order to be considered OPERABLE. Because the OPERABILITY of the ADVs is controlled by this Technical Specification, Technical Specification 3.6.3, Containment Isolation Valves,' does not apply to the ADVs.

ACTIONs The ACTIONs are modified by a note indicating that the provisions of Specification 3.0.4 are not applicable provided one ADV is OPERABLE. This allows for MODE changes with one ADV inoperable provided the appropriate ACTION is entered upon entry into the applicability MODEs.

ACTIONs (a) and (b) would be applicable only when the automatic actuation channels are required to be OPERABLE per the LCO footnote.

a. This ACTION addresses the condition when one ADV is incapable of automatic actuation. This condition includes:
  • A malfunctioning automatic actuation channel, or
  • When the automatic actuation controls for one ADV have been placed in manual.

A 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> allowed outage time is provided to restore the ADV to an OPERABLE status. The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> allowed outage time takes into account the capability afforded by the remaining OPERABLE ADV and is consistent with the allowed outage time of an inoperable high-pressure safety injection train.

If the ADV can not be restored to an OPERABLE status within the allowed outage time, the unit must be placed in a status in which the LCO does not apply. To achieve this status, power must be reduced to less than or equal to 70% RATED THERMAL POWER within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. The LCO will no longer apply once the unit has been at less than or equal to 70% RATED THERMAL POWER for greater than six hours.

b. This ACTION addresses the condition when both ADVs are incapable of automatic actuation. This condition includcs:
  • Malfunctioning of both ADVs automatic actuation channel,

. When the automatic actuation controls for both ADVs have been placed in manual, or

  • A combination of the above such that both ADVs are incapable of automatic operation.

In this condition, the SBLOCA can not be mitigated by one high-pressure safety injection train alone. Therefore, one of the ADVs must be restored to OPERABLE status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or power must be reduced to less than or equal to 70% RATED THERMAL POWER within the next six hours. The LCO will no longer apply once the unit has been at less than or equal to 70% RATED THERMAL POWER for greater than six hours.

WATERFORD - UNIT 3 B314 7- ##/

BASES =

3/4.7.1.7 ATMOSPHERIC DUMP VALVES (Continued)\

c. This ACTION address the condition when one ADV is inoperable for reasons other than those addressed in ACTIONs (a) and (b) above. This condition includes:
  • The inability to operate the ADV manually via the handwheel or the controller in the control room, or

A 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> allowed outage time is provided to restore the ADV to an OPERABLE status. The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> allowed outage time takes Into account the capability afforded by the remaining OPERABLE ADV, a nonsafety grade backup in the Steam Bypass System and MSSVs, the closed system inside containment, and the backup isolation capability of the block valve.

If the ADV can not be restored to an OPERABLE status within the allowed outage time, the unit must be placed in a status in which the LCO does not apply. To achieve this status, the unit must be placed in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

The following conditions are not addressed by the ACTION statements:

  • The automatic actuation channel for one ADV is inoperable and the other ADV is inoperable for other reasons.
  • Both ADVs are inoperable for reasons other than the automatic actuation channels.

For these conditions, Specification 3.0.3 is entered.

Surveillance Requirements

a. To mitigate the SBLOCA event, the ADVs must automatically open at a pressure of less than or equal to 1040 psia (992 psig indicated.) This CHANNEL CHECK provides assurance that the behavior of the steam line pressure input to the automatic actuation channel is reasonable for the existing plant conditions. This steam line pressure input is available on the plant monitoring computer or from appropriate maintenance and test equipment. This Surveillance Requirement (SR) need not be performed when the ADV automatic actuation channels are not required to be OPERABLE per the LCO footnote.
b. To mitigate the SBLOCA event, the ADVs must automatically open at a pressure of less than or equal to 1040 psia (992 psig indicated.) This Surveillance Requirement (SR) ensures that the ADV controllers are in automatic and set at an appropriate setpoint that is bounded by the SBLOCA safety analysis. The setpoint must be verified using the plant monitoring computer or appropriate maintenance and test equipment. This SR need not be performed when the ADV automatic actuation channels are not required to be OPERABLE per the LCO footnote.

WATERFORD - UNIT 3 B3/4 7- #

BASES\'

3/4.7.1.7 ATMOSPHERIC DUMP VALVES (Continued)

c. To perform a controlled cooldown of the reactor coolant system, the ADVs must be able to be opened and throttled through their full range. Additionally, the ADV must be capable of being closed to fulfill its secondary function of containment isolation. This SR ensures the ADVs are tested through a full control cycle.

The test interval is in accordance with the Inservice Testing Program.

d. The SR to calibrate the ADV automatic actuation channels ensures that the system will generate an actuation signal at 1040 psia (992 psig indicated) as assumed for the SBLOCA. The calibration should include the plant monitoring computer points used to set the setpoint.
e. The SR for actuation testing ensures that the ADV will automatically open on a high steam pressure signal as assumed for the SBLOCA. Credit may be taken for an actual or simulated actuation signal.

WATERFORD - UNIT 3 B314 7- ##

PLANT SYSTEMS

BASES, 3/4.7.4 ULTIMATE HEAT SINK The limitations on the ultimate heat sink level, temperature, and number of fans ensure that sufficient cooling capacity is available to either (1) provide normal cooldown of the facility, or (2) to mitigate the effects of accident conditions within acceptable limits.

The UIIS consists of two dry cooling towers (DCTs), two wet cooling towers (WCTs), and water stored in WCT basins. Each of two 100 percent capacity loops employs a dry and wet cooling tower.

Each DCT consists of five separate cells. Cooling air for each cell is provided by 3 fans, for a total of 15 per DCT. The cooling coils on three cells of each DCT (i.e. 60%) are protected AA from tornado missiles by grating located above the coils and capable of withstanding tornado X missile impact. With a Tornado Watch In effect and the number of fans OPERABLE within the missile protected area of a DCT less than that required by Table 3.7-3, ACTION c requires the tA restoration of inoperable fans within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or plant shutdown as specified. This ACTION is based on FSAR analysis (subsection 9.2.5.3.3) that assumes the worst case single failure as, 1 emergency diesel generator coincident with a loss of offsite power. This failure occurs h subsequent to a tornado strike and 60% cooling capacity of a DCT is assumed available.

Each WCT has a basin which is capable of storing sufficient water to bring the plant to cafe shutdown under all design basis accident conditions. Item a of LCO 314.7.4 requires a minimum water level in each WCT basin of 97% (-9.86 ft MSL). ev W watvor rna drift lospcalcu fons iccouded LO 64dunn oa ln 1glln 0 ith thvlon-es~ntial~do n ulln~l e i p t aogn a i ~ e lvli a na i L each asin as a minimum capacity of 174 000 gallons. The WCT basin is also credited as a source of Emergency Feedwater (EFW). e b L A er bounds the amount of EFW required from the WCT basin for all design is acciden Each WCT consists of two cells, each cell is serviced by 4 Idn ed draft fans, a total of 8 per WCT.

There is a concrete partition between the cells th prevents air recir tion between the fans of each cell. Covers are required on fans declared ut-of-service to preve ir circulation between fans within a cell. -he W T' -;- 1 W 4 Table 3.7-3 specifies increased or decreased fan OPERABILITY requirements based on outside air temperature and humidity. The table provides the cooling tower fan OPERABILITY requirements that may vary with outside ambient conditions. Fan OPERABILITY requirements are specified for each controlling parameter (i.e., dry bulb temperatures for DCT fans and wet bulb temperatures for WCT fans). The calculated temperature values (EC-M95-009) associated AMENDMENT NO.9', 123 139 WATERFORD - UNIT 3 B 3/4 7-4 CHANGE NO,4-

Bazos 3/4.7.4 Insert This minimum WCT basin capacity contains enough volume to account for water evaporation and drift losses expected during a LOCA. Additional volume is needed from the second WCT basin to handle the non-essential load of fuel pool cooling during the LOCA.

(The WCTs can be manually interconnected through a Seismic Category I line.)

PLANT SYSTEMS BASES (Continued) 3/4.7.4 ULTIMATE HEAT SINK (Continued) with DCT and WCT fan requirements have been rounded in the conservative direction and lowered at least one full degree to account for minor inaccuracies. Failure to meet the OPERABILITY requirements of Table 3.7-3 requires entry into the applicable action. Because temperature and humidity are subject to change during the day, ACTION d requires periodic.

temperature readings to verify compliance with Table 3.7-3 when any cooling tower fan Is inoperable.

The limitations on minimum water level and maximum temperature are based on providing a 30-day cooling water supply to z qipment without exceeding their design basis temperature and is consistent with the reco mendations of Regulatory Guide 1.27, "Ultimate Heat Sink for Nuclear Plants," March 1974.

AMENDMENT NO. 449, WATFRFC)Rn

.. ... - UINIT I B

- 314

-- 4

- (1)

I -

CHANGE NO.+-

3/4.8 ELECTRICAL POWER SYSTEMS BASES 3/4.8.1. 3/4.8.2. and 3/4.8.3 A.C. SOURCES. D.C SOURCES. AND ONSITE POWER DISTRIBUTION SYSTEMS The OPERABILITY of the A.C. and D.C. power sources and associated distribution systems during operation ensures that sufficient power will be available to supply the safety-related equipment required for (1) the safe shutdown of the facility and (2) the mitigation and control of accident conditions within the facility. The minimum specified independent and redundant A.C. and D.C. power sources and distribution systems satisfy the requirements of General Design Criterion 17 of Appendix A to 10 CFR Part 50.

The Limiting Condition for Operation (LCO) ensuresiEeach diesel generator 7or s ali~uel oil of a sufficient volume to operate each diesel generator for a period of 7 da s. The minimum required volume isbased on the time-dependent loads of the diesel ienrator following a loss of offsite power and a design bases accident and includes th, capacity to power the engineered safety features in conformance with Regulatory e 137 Ci b October 1979. The minimum onsite stored fuel oil is sufficient to operate the diesel tnerator for a period longer than the time to replenish the onsite supply from the outside sources discussed in FSAR 9.5.4.2.

An additional provision is included Inthe LCO which allow the diesel generators to remain C operable when their 7 day fuel oil supply Is not available. This provision is acceptable on the v basis that replacement fuel oil is onsite within the first 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after falling below the 7 day supply.

The ACTION requirements specified for the levels of degradation of the power sources provide restriction upon continued facility operation commensurate with the level of degradation. The OPERABILITY of the power sources are consistent with the initial condition assumptions of the safety analyses and are based upon maintaining at least one redundant set of oncito A.C. and D.C. powor courcoc and ascociatod distribution cystems OPERABLE during accident conditions coincident with an assumed loss-of-offsite power and single failure of the other onsite A.C. source. When one diesel generator is inoperable to perform either preplanned maintenance (both preventive and corrective) or unplanned corrective maintenance work, the allowed-outage-time (AOT) can be extended from 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to 10 days, if a temporary emergency diesel generator (TEDG) isverified available and aligned for backup operation to the permanent plant EDG removed from service. The TEDG will be available prior to removing the permanent plant EDG from service for the extended preplanned maintenance work or prior to exceeding the 72-hour AOT for the extended unplanned corrective maintenance work. A Configuration Risk Management Program (CRMP) is implemented to assess risk of this activity when applying this ACTION. The TEDG availability is verified by: (1) starting the TEDO and verifying proper operation: (2) verifying 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> onsite fuel supply; and (3) ensuring the TEDG is aligned to supply power through a 4.16 kV non-safety bus to the 4.16kV safety bus. A status check for TEDG availability will also be performed at least once every 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> following the initial TEDG availability verification. The status check shall consists of: (1) verifying the TEDG equipment is mechanically and electrically ready for manual operation; (2) verifying 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> onsite fuel supply; and (3) ensuring the TEDG is aligned to supply power through a 4.16 kV non-safety bus to the 4.16 kV safety bus. If the TEDG becomes unavailable during the 10 day AOT and cannot be restored to available status, the EDG AOT reverts back to 72-hours. The WATERFORD - UNIT 3 B 3/4 8-1 AMENDMENT NO. 2,-i66-

Bases 3.8.1 Insert This 7-day minimum required volume of fuel oil is bounded by the combined LCO required fuel oil volumes for the fuel oil storage tank and fuel oil feed (day) tank for each diesel generator.

Attachment 4 W3FI-2004-0052 Changes to Technical Requirements Manual Pages (mark-up)

For Intormatlon Only

TABLE 3.3-2 REACTOR PROTZCTIVE INSTRUMNTATION RESPONSE TIMES FUNCTIONAL UNIT RESPONSE TIHZ

1. Manual Reactor Trip Not Applicable
2. Linear Power Level - High 5 0.40 second*
3. Logarithmic Power Level - Nigh S 0.40 second*
4. Prensurizer Pressure - High 5 0.90 second
5. Pressurizer Pressure - Low S 0.90 second
6. Containment Pressure - High S 1.70 seconds
7. Steam Generator Pressure - Low S 0.90 second
8. Steam Generator Level - Low 5 0.90 second
9. Local Power Density - High Rge Iqce a'i#/ TR L40i5 3.3-,Z X4I'SZer Cq

/ b. A Poii ti5

10. DN8R - Low I

1;4oLgZ 3, 3-2-IP5r rT /c 3/4 3-2 AMENHEr S

TRM TABLE 3.3-2 INSERT 9

a. Neutron Flux Power from Excore Neutron Detectors < 0.191 second*
b. CEA Positions s 0.186 second"
c. CEA Positions: CEAC Penalty Factor #1 Channe IA 5 0.271 second Channel B 0s.236 second ChanneiC
  • 0.236 second ChannelD s 0.236 second
d. CEA Positions: CEAC Penalty Factor #2 Channol A
  • 0.236 second Channel B s 0.236 second Channel C
  • 0.236 second Channel D s 0.271 second TRM TABLE 3.3-2 INSERT 10
a. Neutron Flux Power from Excore Neutron Detectors
  • 0.191 second*
b. CEA Positions
  • 0.186 second"
c. Cold Leg Temperature 5 0.285 second#
d. Hot Leg Temperature s 0.285 second#
e. Primary Coolant Pump Shaft Speed s 0.185 seconds
f. Reactor Coolant Pressure from Pressurizer < 0.186 second##
9. CEA Positions: CEAC Penalty Factor #1 Channel A
  • 0.271 second Channel B s 0.236 second ChannelC
  • 0.236 second Channel D 5 0.236 second
h. CEA Positions: CEAC Penalty Factor #2 Channel A
  • 0.236 second Channel B 5 0.236 second Channel C s 0.236 second Channel D s 0.271 second

TABLE 3.3-2 (Continued)

REACTOR PROTECTIVE INSTRUMENTATION RESPONSE TIMES FUNCTIONAL UNIT RESPONSE TrME

11. Steam Generator Level - High Not Applicable
12. Reactor Protection System Logic Not Applicable
13. Reactor Trip Breakers Not Applicable
14. Core Protection Calculators Not Applicable C4 4
15. CZA Calculators Not Applicable
16. Reactor Coolant Flow - Low 0.10 second 5-_, b
  • Neutron detectors are exempt from response time testing. Vesponse time of the neutron flux signal I lb portion of the channel shall be measured from detector output or input of first electronic component in channel.  ;, n J
    • Response time shall be measured from the time the CPC/CZAC receives an input signal until the electrical '.'0 power in interrupted to the CEA drive mechanism.

-S_

fResponse time shall be measured from the output of the sensor. RTD response time for all the RTDs shall be measured at least once per 18 months. The measured Pe of the slowest RTD shall be less than or equal to 8 seconds (Pt assumed in the safety analysis).

t#Response time shall be measured from the output of the pressure transmitter. The transmitter response .1 time shall be less than or equal to 0.70 second.

3/4 3-3

3/4.3 INTRMENTATIO A,/eeuAF P Al -e -

\BASES i 3/4.3.1 and 3/4.3.2 REACTOR PROTECTIVE AND ENGINEERING SAFETY FEATURES ACTUATION SYSTEMS INSTRUMENTATION Items 9 and 10 in Table 3.3-2 provide the surveillance test acceptance criteria for Core Protection Calculator (CPC) response times consisting of:

  • CPC system hardware, plus
  • time to interrupt power to the CEA holding coils.

These are the maximum times allowed in surveillance procedures. With the sum of these measured response times less than the times shown In Table 3.3-2, the CPC system hardware, the reactor trip matrix and time to interrupt power to the CEA drive mechanism are demonstrated to be functioning to specification and will support the response times assumed in safety analyses.

Having established the functioning of the CPC hardware, the creditable trip times are based upon fixed execution time and soquencee of the various CPC system coftwarc modules.

The CPC delay times for these functional units that are acceptable to assume in accident and transient analysis corresponding to Table 3.3-2 instrumentation response times are listed below. The times include reactor trip matrix time, and time to interrupt power to the CEA holding coils, but omit detector response times.

Functional Unit Delay Time (second(s))

9. Local Power Density - High
a. Neutron Flux Power From Excore Detectors 0.2748
b. CEA Positions 1.3039
c. CEA Positions: CEAC Penalty Factor #1 0.4631
d. CEA Positions: CEAC Penalty Factor #2 0.4631
10. DNBR-Low
a. Neutron Flux Power From Excore Detectors 0.2748
b. CEA Positions 1.3422
c. Cold Leg Temperature 0.3693'
d. I-lot Leg Temperature N/A 2
e. Primary Coolant Pump Shaft Speed 0.2315
f. Reactor Coolant Pressure from Pressurizer 0.2693
g. CEA Positions: CEAC Penalty Factor #1 0.6014
h. CEA Positions: CEAC Penalty Factor #2 0.5014 1 Also applicable to CPC Auxiliary Trip on Differential Cold Leg Temperature 2 No Direct Trip on Hot Leg Temperature.

B 3/4#

314.3 INSTRUMENATO A/eu) 'tR M1 13arst BASES 314.3.1 and 3/4.3.2 REACTOR PROTECTIVE AND ENGINEERING SAFETY FEATURES ACTUATION SYSTEMS INSTRUMENTATION (Continued)

The time delays acceptable to assume in accident and transient analysis for module based CPC trip delay times for trips based on CPC logic include:

Hot Leg Saturation trip: 2.744 seconds Low Quallty Margin trip: 0.370 second Differential Cold Leg Temperature trip: 0.370 second Axial Shape Index trip: 0.275 second Integrated Radial Peaking trip: 1.521 seconds Variable Over Power trip (VOPT): 0.370 second '

Low DNBR Trip for Excess Load with Loss of AIC, 0.332 second 2 A 0.370 second response time is generally the minimum time assumption, where both temperature and neutron flux inputs to VOPT are considered.

2 A 0.332 second response time applies for this event, based on detection of a decrease of reactor coolant flow from conditions assumed to be at the power operating limit.

The time delay for the module based trips for DNBR-Low and LPD-High can vary dependent upon the dynamics of the various parameters which are input to the algorithms.

B 3/4#

TABLE 3.3-5 (Continued)

ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND IFUNCTION RESPONSE TIME IN SECONDS

2. Pressurker Press
a. Safety In;oction (ECCS)

(1) High ;Pressure Safety Injection (2) Low Pressure Safety Injection S

b. Cordainm ant Isolation 5 23.5112.0-C. Contalnn,nent Cooling S31.0 119.5-
3. Contaiment Pressure-iah
a. Safety Injection (ECCS)

(1) HKh Pressur Safaty Injrdion (2) Low Pressure Safety Injection

b. Containment Isolation
c. Main Stnam Isolation S 8.0 1B.0" d Mahi Feedwater Isolation ### S 6.0-/6.0-
e. Containment Cooling !S31 .0-119.5*
4. Containment Pressure-Hiot-ligh
a. Containment Spray Pump S 15.2 14.6-
b. Containment Spray Valves S 10.4/110.4-
c. CCW to RCP Valves (CC-641, CC-710 & CC-713) 2 5.0. 50.0
  • I 5.0. S 50.0-¢ S. Containment Area Radiation-HWgh #

Containment Purge Valves Isolation S62/16.2"

6. St
a. Main Steam Isolation S 8.0*I8.0"
b. Main Feedwater Isolation ### S 6.0*I6.0-
7. Refuelina Water Storsae Pool-Luw Safety Injection Sump S 50.0w Recirculation Valves Open
8. 4.16 kV Emeroency Bus Undervoftaoe (Loss of Volbae)

Loss of Power (0 volts) S 2~

9. 480V Emergencv Bus Undervoltace (Loss of Voltanea Loss of Power (0 volts) N.A.
10. 4.16 kV Emoroency Bus Underyoltage (Deoraded Mo oaoe)

Loss of Power S 14"'

3/4 3.6 AMENDMENT NO. 25,-38r4,

- (MDRN2477. Am. W 3/4.3 INSTRUMENTATION 3/4.3.5 ULTRASONIC FLOWMETER LIMITING CONDMTON EOR OPERATION 3.3.5 Two ultrasonic flowmeters (UFMs) shall be OPERABLE.

_ PM 934247. AM. 75)

APPLICABILITY: MODE 1, above 50% power

--(DOW034247, Am. 75)

ACTION:

With one or both UFMs inoperable, perform the following:

- (DRN 03-0247. Am. 75)

a. With COLSS In service:

. (DRN 02-1U3. Am. 74)02-139, Am. 74)

(O RDAN

- (ORN 02-1888. Am. 74: DRN 03.247. Am. 75

1. Reduce THERMAL POWER to less than or equal to 4pv1W) l within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> of last calibration factor update.
2. Reduce THERMAL POWER to less than or equal to9MvW .5%)

within 31 days of last calibration factor update.

b. With COLSS out of service, return the UFM to service prior to the next required 3 0 CPCS; calibration or reduce THERMAL POWER to less than or equal to t (98.5%).

4- (DRAN2-.131, Am. 74 R 0247. Am. 75)

The provisions of Specifications 3.0.3 and 3.0A are not applicable.

SURVEILLANCE REQUIREMENTS 4.3.5 The ultrasonic flowmeters shall be demonstrated OPERABLE:

a. By the performance of a CHANNEL FUNCTIONAL TEST at least once per 18 months.
b. By the performance of a CHANNEL CHECK at least once every 31 days to demonstrate continuous calibration of COLSS MSBSCAL and FWBSCAL.

4 (ORN 02477, Aim 4)

(DRN 034247. Aim 75)

Following a reactor power cutback, entry Is not required provided UFM is returned to service within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

4 (URN 8142A7. A.. 71) 314 3-40 AMENDMENT NO. 56.44.-,5-

- (MUM02477.A 5A 314.3 INSTRUMENTATION bAS. .

314.3 5 ULTRASONIC FLOWMETERS

-. (CDR8 2.18.,Am. 74. VR8103447, Am. 75)

The ultrasonic flowmeters (UFMs) measure feedwater flaw and bulk feedwater tomporaturo. Tho UFM foodwator flow and foedwater bulk temperature inputs will be used by the Core Operating Limits Supervisory System (COLSS) to calculate station secondary calorimetric power. The UFM feedwater flow and feedwater bulk temperature inputs are also used as inputs into calibration constant algorithMms that compensate or calibrate the altemate feedwater and main steam venturi-based flows and feedwater temperature instrumentation inputs used by COLSS on a loss of UFMs.

The loss ofa UFM wIN cause ntrol room a rm to ann nclate and COLSS to automatically default to the compe ed alternate venturi-based i strumentation inputs.

COLSS normally defaults to Mai m BSCAL (MSBSCAL) whe reactor core power Is greater than or equal to 05% MWt RATED THERMAL PO R (RTP) or Foodwator BSCAL (FWBSCAL) when reactor core power Is less than 95% of MWt RTP. MSBSCAL and FWBSCAL are calibrated by the UFM calibration factors. The requirement for the UFM to be operable above 50% power ensures that feedwater temperature Isgreater than the temperature (250 F) at which the UFM Is reliable and the most accurate power measurement instrumentation is used over a large power range.

- 024.18, Am 74. RN 03447. Am. 75)

(DRM4

-(DMN 02.1883, Am. 74: DRN 03-247. Am. 75) e (DRN 024-US, Am. 74; DRN 03247. Am. 75)

CORN02.1839, Am 74; DRN 03-247, Ar. 75)

Within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> following the laoss of Q LUFMs on tor action must be taken to reduce THERMAL POWER to less than or equal tot>M~t ). The decrease In power within the 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> completion time takes into account the reduction of confidence in the UFM based calibration factors resulting from COLSS alternate instrumentation loop drift caused by time and ambient temperature uncertainty effects. On restoration, THERMAL POWER should be maintained at the previous TRM action level until UFM calibration factors are developed.

-(iRN 02413S3, Am. DRN 0447. Am 75)

-I (DMN e47.

Amn.75 Within 31 days followi the loss of the UFMs, operator action must be taken to reduce power to less than or equal t Wt (98.5%). The decrease In power within the 31 day completion time takes Into accou the loss of confidence in the UFM based calibration factors.

4- PDRN0247. Am. 54: DRN 03-247. Am. 75)

B 3/4 2a AMENDMENT NO. 66.7445;

-* CORN 02477. Am 3/4.3 INSTRUMENTATION MASES 3/4.3.5 ULTRASONIC FLOWMETERS (contd)

DRk 02-18. Am. 74< MM 03447. Am. 75)

The 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> and 31 day LCO ACTION STATEMENTS are required to maintain consistency with the COLSS Secondary Calorimetric Measurement Uncertainty analyses. The entry into LCO ACTION STATEMENTS begins with the loss of the UFM, when greater than 50% power. The appropriate ACTION STATEMENT time limit entry will be based on the last time MSBSCAL and FWBSCAL wear updated by the UFM calibration factors prior to UFM failure. These ACTIONS ensure CPC margins to trip remain conservative and preserve the Appendix K ECCS limits.

4-- DRN 02-15W9,Am. 74. OM 33U4. Am74n)

(DRN 034247, Am. 777 If COLSS Is out of service Core Protection lator System (CPCS) will continue to maintain plant operations n the core power ope *ng imlts. Operating limits will be maintained through comPI' ce with Technical Specif ica (TS) sections 3.2.1. 32.3, 3.2.4, 3.2.7, and 4.3.1.1, Table .3-1 (2. 9, 10, 14) applica e ACTI N STATEMENTS and Surveillance Requirepnts (SR). If the UFM(s) ERABL during the period COLSS is out of service, then pla operation may continue at t R using the power indications from the CPCS a96 UFM based manual secondary calorimetri measurement. If the UFM(s) becomes IN P LE during the period COLSS is out of se *ce, then plant operation may continue a MWt RTP using the power indications from th CS. However. In order to remain in compliance with the bases for operation at a RTP of MWt. the UFM(s) must be returned to service prior to the next uired daily CPCS calibration or THERMAL POWER must be reduced to less than or equal to t (98.5%). This power reduction is performed prior to the next CPC calibration in order to in within the alternate venturi-based Instrumentation power measurement uncertainty analysis a aintain consistency with the COLSS Secondary Calorimetric Measurement Uncertainty analye e 247. As-7e PA 0D" The TRM is annotated with a 3.0.4 exemption, allowing entry into the applicable Mode to be made with UFMs INOPERABLE, as required by the Actions.

The SR 4.3.5.a. to perform a CHANNEL FUNCTIONAL TEST at least once per 18 months is based on the vendor recommendations. The UFM equipment contains on-line self-diagnostic capabilities to continuously verify operation within Its design bounds. The 18 month frequency Is based on the refueling cycle. This frequency is acceptable from a reliability standpoint.

The SR 4.3.5.b 31 day CHANNEL CHECK verifies the COLSS alternate calorimetric heat balances MSBSCAL and FWBSCAL are within a value bounded by engineering analyses. This comparison of alternate heat balances to USBSCAL assures the calibration factors derived by USBSCAL are valid and the basis for the original power measurement uncertainty assumptions for operation are maintained at the various thermal power levels when UFM Is Inoperable.

- (D0747. Am 58)

B 314 2b AMENDMENT NO. 66,74,- e

,31.4 RECTOR COOLAN ><f4}Sat rYSEM 314.4.3 PRESSURIZER HEATERS LIMITING CONDITION FOR OPERATION 3.4.3.1 The pressurizer heaters shall be OPERABLE with at least 650 kW of nominal heater capacity available in addition to the heater capacity specified in Technical Specification 3.4.3.lb.

APPLICABILITY: MODE I ACTION:

With less than the above heater capacity available, restore the required capacity within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or initiate a Condition Report which requires the performance of an evaluation justifying continued plant operation. The evaluation should be completed within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and should be approved by the General Manager Plant Operations (GMPO) or his designee (e.g., the duty plant manager).

SURVEILLANCE REQUIREMENTS 4.4.3.1 The specified heater capacity shall be verified at least once each refueling outage.

3/. AREACTOR COOLANT SYSEX

/BASES 314.4.3 PRESSURIZER The heater capacity cited in requirement 3.4.3.1 is sufficient, in conjunction with the heater capacity required by Technical Specification 3.4.3.1 b, to bound the capacity credited in the analysis of the CEA Withdrawal within Deadband event. The additional heater capacity cited in requirement 3.4.3.1 can be heaters powered from any combination of Class 1E or non-class 1E buses.

3/4 6-5a Amcndmcnt 42-64 3/4.6 CONTAINMENT SYSTEMS BASES 3/4.6.1 PRIMARY CONTAINMENT Primary CONTAINMENT INTEGRITY ensures that the release of radioactive materials from the containment atmosphere will be restricted to those leakage paths and associated leak rates assumed in the safety analyses. This restriction, in conjunction with the leakage rate limitation, will limit the SITE BOUNDARY radiation doses to within the limits of 10 CFR Part 100 during accident conditions.

3/4.6.12 CONTAINMENT LEAKAGE The limitations on containment leakage rates ensure that the total containment leakage volume will not exceed the value assumed Inthe safety analyses at the peak accident pressure, P., As an added conservatism, the measured overall integrated leakage rate is further limited to less than or equal to 0.75 La or less than or equal to 0.75 Lt as applicable during performance of the penodic tests to account for possible degradation of the containment leakage barriers between leakage tests.

The surveillance requirements for measuring leakage rates are consistent with the requirements of Appendix J of 10 CFR Part 50. A one time extension of the test Interval Isallowed for the third Type A test of the first 10-year service period, as required by Surveillance Requirement 4.6.1 .2.a and by Section IlI.D.(a) of Appendix J to 10 CFR Part 50, provided the performance of the Type A test occurs prior to unit restart following Refuel 7.

~TEMPERATURE (The tion on containm minimum avo air tomporature ens es that the ECC s capable maintaining a Pea d Temperatur CT) less than or eq o 22000 F under CA co ons. A lower cont 'ent average arI perature results In wer post accide ntainment pressurelower reflood ia and thereforA a high CT. Lowering t eak Unear Heat Rate setpoi COLSS by 0.2 Ift. for every 1OF be the minimum cotinment temperature Ii of 90SF will ens the resulting PCT re ins bounded by thoFSAR accide analyses.a value of 0.2 k . was calculated in a sitivity analysis peftmed by ABB dthe resu e documented I ABB letter to J.BHan dated June 29 3 (Letter num rL-93 l (rR02-15T1) /////

The 0F minimum value sp ied in the TRM is the lue used in the dent analys~ nd tdoes n~ontain any allowance temperature measur nt Instrument u rtainty. Ins ent 314.6.3 CONTAINMENT ISOLATION VALVES The OPERABILITY of the containment Isolation valves ensures that the containment atmosphere will be Isolated from the outside environment In the event of a release of radioactive material to the containment atmosphere or pressurization of the containment and Is consistent with the requirements of GDC 54 through GDC 57 of Appendix A to 10 CFR Part 50. Containment isolation within the time limits specified for those Isolation valves designed to close automatically ensures that the release of radioactive material to the environment will be consistent with the assumptions used in the analyses for a LOCA.

B 3/4 3 AmendmentU 4

s I/l DgFOROPFRATIQbV/ /

3.7.1.7 TA mosphericDum aves (ADVs) s beOPERABLE.

APPLK§ABLB: ,b(ODS 1, 2, 3, ay(

a. With one V nopeable. e TS 3.6.3 and te required ADV to PERABLE satus In 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />sor atastHOT ST BY wthin te next uraandln CO SHUTDOWN with the followia30
b. Wit two Atrospl Dump Valves ble, ente TS 3.6.3 rnstore one ADV to OPERABLE ituswhin hour n at east HOT ST DBYwthin the next hours and In LDSHUTDOWN the foflowing 30 The provisions Spedfictlon 3.0.4 a n pp6lcablo providonAtro phedc Du Vdivu b 4.7.1.7 The Atro5 rio Dump Valves shall pe meronstrated OP LE by veryng e a1

'mplete of each valve n a dnce with tInse( Ung Progr4/

+'W 02D-17KC-b /3 / f 314 7,1 c AMENDMENT NO. 46, 38, 47 GGrBO

7 The ADVs have two ose for dontainment isolation and 2) in order to provide for cooldo latiqWconditions. The ADV close s) function is addressed in To Wpdal (TRM) Table 3.6-2. Contairfni Isolation Valves, as coEainr Idis applicable to TS 3.. AON statements. This on ad, rfunction of the valves. /

The n safety function of the A !s provide a safety grade me>6d for cooling the unit to Shutdo Cooling (SDC) System en conditions, should the preferd heat sink via the Steam pass System to the condeo r not be available, as discus d in the FSAR, Section 10.3 his safety grade cooldown nction is performed in conju ion with the Emergency Fdwater System providing ng water from the condensa torage pool (CSP).

Two ADV9, one p steam generator, are provid Each ADV flow path consists3an ADV and an associated ock valve. The ADVs are pr ided within upstream block vat s to permit their being tes d at power, to provide an alt te means of isolation should a ADV not close on demand, d to allow maintenance of th valves with the steam headers ressurized.

The ADVs are e pped with pneumatic controars to permit control of the stea release rate and associate cooldown rate. The controll s provide both automatic and Tenual ADV operating m es. The ADVs are normall5perated using the plant non-s ty instrument air suipply. .tpsiste afety~-rAltAd nitr=~ g ,-timutators are provided as ackups to each of the ADVs I he event of a loss of instru nt air.

The ADVs are one of Use systems required to meet Bra Technical Position (BTP SB 1, Deiyiii Requihumeits ofie Residual Heat Removal S m, and 10 CFR SO. Appen R for taking the plant from no al operating conditions to co shutdown. A cooldown an ysis was performed assuming los of offsIte power and the failur of one ADV to open. The Analysis determined that shutd en cooling entry conditions w ld be reached in approximat'ey ten hours.

The Safety Class 3,Oeismic Category I accumula rs provide a ten hour minim backup supply of motive gas forge ADV actuators to assure valves remain operable fr the control room until shutdown,6 ling entry conditions are s sfied. This is based on the me needed to reach shutdown co ing (S1DC) conditions whon th hot log tomperoturoo o r duccd to 400'F. This temperatur is the design temperature the shutdown cooling syst (SDCS) components.

The ADVs are used durinn mal plant startups and for ldowns when either a vac m in the condenser or the team Bypass Control Syste (SBCS) is not available. The s are capable of being rated remotely from either Control Room or the Remot utdown Panel (LCP 43) nd locally from the local pn matic panel or with manual handwhAAIo.

In the )resented in the SAR, the AD/s are assum operator to co4 SDC Syste ntry conditions for accidents loss or ofrfte i operator Ion, the main steam safety v A used to ma"yf r press and temperature at the MSS esl action is $sur es fo wing the initiation of an event he Ii er r unavailable for RCS heat re oval. wit vents falling into this categoryre a main isolation valves and the steajd generators, .

orator tube rupture (SGTR)/

B 314 3e AMENDMENT NO. 4X,9

P"82.

314.T C DUMP (Contl Is required iOPERAB each:r ne r to ensure ADV Is OPE to conduct a nt cookk)wn an e ent, wh;ch one steam ge and ft con r booorno u AD red 0 LE ifit can be s thruugha te ram. Failure to LCO can In s coor, e conditions an event In the conden ) ADVs are uired to be OPE In MODES 19 20 39 and 4. In th the provide the rade cooklown to shutdown cooring en one ADV Is I 72 hou ed to return ERABLE sta outage t nto e capability a by the remaining RABLE and i on-safe S and IVISS, If ADVs are Inopera TS 3.0.3 Is red and I hou ProWded ADV W a P LE status. Ifth cannotbe restored to OPE LE statu e associated Time, the pla mu be placed In a MODE In the LCO d 3 iy. To achieve status. the pla be placed in at least MOD In6 hours7anj COLD SH in the follawl hours. al Compl es are reasonable based on operas experience, to require conditio full power con s Inan ordprly er and vAtho Ilenging P12N syste 4-The TRM Is an th a 3.0 alto" the applica ES to made wfth an ADV.

-#VMWUWAM8%

To pe led coold the RCS, the must be able to and throttled th eir full range. TheAurveillance requi nts ensure the AD re tested through a full le Inaccordance ft the Inservice ng Program.

(DM B DMENT

TABLE 3.6-2 CONTAINMFNT ISOILATION VALVES Ae---n Annm Nru -IGALS) zU 0 x Mc:^ -D n0aZC )

n=CD Ci0C e

K

)

n c tn m

g g t C)

VALVE NUMBER FUNCTICO N _

LRT MVAAA109 ILRT Connection 63 X -

LRT MVAAAI 10 ILRT Connection 63 X -

LRT MVAAA201 ILRT Test Connection 65 X - -

LRT MVAAA20 t ILRT Test Connection 65 X -

LRT MVAAA202 ILRT Test Connection 65 X -

LRT MVAAA203 ILRT Test Connection 65 X - - _

LRT MVAAA2031 ILRT Test Connection 65 X_.

LRT MVAAA204 ILRT Test Connection 65 X_

LRT MVAAA400 LLRT Test Connectbn 53 X MS l\VM1K Atgwpheric Rp1VIaive / )

(

MS MVAAA119A Main Steam Drain 1 X - - -

MS MVAAM119B Main Steam Drain 2 X_

MS MVAAA120A Main Stetm Drain 1 X -

MS MVAAA120B Main Steem Drain 2 X -

MS MVAAA1244A MSIV Bypass 1 X -

MS MVAM1244B MSIV Bypass 2 X 413RN 03.1738. Anm 81) 4-2RNO03.1738.AM 81) _

MS MVAAM401A Steam to Emergency Steam Generator 1 X (4 X Feed Pump Turbine I /open)

MS MVAAA401 B Steam to Emergencl Steam Generator 2 X "; X Feed Punp Turbine _- I iopen) 3/4 6-12 0 6 AMENDMENT NO. -48 t

Attachment 5 W3FI-20040052 Revised Power Uprate Report Section 2.12.4, Small-Break LOCA Analysis to W3Fl-2004-0052 Page 1 of 32 2.12.4 Small Break LOCA Analysis 2.12.4.1 Methodology The Small Break Loss-of-Coolant Accident (SBLOCA) Emergency Core Cooling System (ECCS) performance analysis used the Supplement 2 version (referred to as the S2M or Supplement 2 Model) of the Westinghouse SBLOCA ECCS Evaluation Model for Combustion Engineering (CE) Pressurized Water Reactors (Reference 2.12-23). This is the same methodology used in the current licensing basis Waterford 3 SBLOCA ECCS performance analysis (Reference 2.12-30). The Safety Evaluation Reports (SERs) documenting Nuclear Regulatory Commission (NRC) acceptance of the S2M are contained in References 2.12-13, 2.12-24, and 2.12-25.

In the S2M evaluation model, the CEFLASH-4AS computer program (Reference 2.12-26) is used to perform the hydraulic analysis of the Reactor Coolant System (RCS) until the time the Safety Injection Tanks (SITs) begin to inject. After injection from the SITs begins, the COMPERC-11 computer program (Reference 2.12-7) is used to perform the hydraulic analysis.

The hot rod cladding temperature and maximum cladding oxidation are calculated by the STRIKIN-Il computer program (Reference 2.12-11) during the initial period of forced convection heat transfer and by the PARCH computer program (Reference 2.12-1 0) during the subsequent period of pool boiling heat transfer. Core-wide cladding oxidation is conservatively represented as the rod-average cladding oxidation of the hot rod. The initial steady state fuel rod conditions used in S2M analyses are determined using the FATES3B computer program (Reference 2.12-12). The SERs for the SBLOCA ECCS performance analysis computer programs are documented in References 2.12-13, 2.12-16 through 2.12-18, and 2.12-24. The SERs for FATES3B are documented in Reference 2.12-20 through 2.12-22.

COMPERC-1l was not run in the Waterford 3 extended power uprate SBLOCA analysis because the limiting break size did not credit injection from the SITs. As is typical of S2M analyses, the limiting break size was determined to be the largest small break for which the Peak Cladding Temperature (PCT) occurs at approximately the same time that injection from the SITs starts.

In this case, the PCT for the limiting break size was calculated to occur approximately 13 seconds after SIT injection would have started had it been credited. STRIKIN-I1 was also not run in the analysis because the PCT, which occurs during the pool boiling portion of the transient calculated by PARCH, is not significantly impacted by the results of the forced convection period calculated by STRIKIN-Il.

The SBLOCA analysis was performed for the fuel rod conditions that result in the maximum initial stored energy in the core. Both U0 2 and erbia burnable fuel rods were analyzed. In addition, studies were performed using PARCH to determine the fuel rod internal pressures that cause cladding rupture to occur at the times that result in the maximum PCT and the maximum cladding oxidation for the limiting break.

The analysis credits operation of one steam generator Atmospheric Dump Valve (ADV) in the automatic mode of operation with an opening pressure of 1040 psia. The automatic operation of an ADV with an opening pressure of 1040 psia results in lower RCS pressure compared to only crediting the Main Steam Safety Valves (MSSVs) for secondary side pressure control. The Waterford 3 ADVs are safety grade equipment. They are required to be in automatic at power levels greater than 70% of the uprated power level. The ADVs are not required for SBLOCA ECCS performance for power levels below 70% of the uprated power level since, within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> to W3Fl-2004-0052 Page 2 of 32 of operation at 70% of the uprated power level, core decay heat is less than the decay heat associated with the pre-uprated power level and an ADV was not credited in the pre-uprate SBLOCA analysis.

The analysis was performed using the failure of a direct current (DC) bus as the most limiting single failure. Failure of a DC bus results in the failure of one Emergency Diesel Generator (EDG) and one ADV. The failure of an EDG in turn causes the loss of a High Pressure Safety Injection (HPSI) pump and a Low Pressure Safety Injection (LPSI) pump. This results in a minimum of safety injection water being available to cool the core. Based on the failure of an EDG and the design of the Waterford 3 ECCS, 75% of the flow from one HPSI pump is credited in the SBLOCA analysis. The LPSI pumps are not explicitly credited in the SBLOCA analysis since the RCS pressure never decreases below the LPSI pump shutoff head during the portion of the transient that is analyzed. Charging pump flow is not credited in the Waterford 3 extended power uprate SBLOCA analysis.

A spectrum of three break sizes, namely, 0.05, 0.055, and 0.06 ft2 breaks in the Reactor Coolant Pump (RCP) discharge leg, was analyzed. The RCP discharge leg is the limiting break location because it maximizes the amount of spillage from the ECCS. The limiting SBLOCA, the 0.055 ft2/PD (Pump Discharge) break, is the largest break size for which the hot rod cladding heatup transient is terminated solely by injection from a HPSI pump. The analysis of the 0.05 ft2/PD break demonstrates that break sizes smaller than the limiting break experience less and later core uncovery and, therefore, are less limiting. The analysis of the 0.06 ft2/PD break demonstrates that breaks larger than the limiting break size are sufficiently large to allow injection from the SITs, in conjunction with the injection from a HPSI pump, to recover the core and terminate the heatup of the cladding before the cladding temperature approaches the PCT of the limiting SBLOCA.

Two modifications were made in the application of the S2M in the Waterford 3 extended power uprate SBLOCA analysis. The modifications were a consequence of crediting an ADV in the analysis. The following is a brief description of the two modifications.

First, the CEFLASH-4AS model for representing steam generator secondary side steam relief valves was modified. Previously, the model was limited to representing both steam generators with steam relief valves that had the same opening pressures and relief areas. The model was modified to allow different opening pressures and relief areas for the two steam generators.

This was required to represent one ADV and the MSSVs on one steam generator and only the MSSVs on the other steam generator.

Secondly, the CEFLASH-4AS nodalization of the cold legs of the intact loop was modified.

Previously, the two cold legs of the intact loop were lumped together into a single set of nodes and flow paths to minimize the number of nodes and flow paths and therefore to minimize computer time. The nodalization was modified to explicitly represent the two cold legs using the same nodalization as used for the broken loop (see Figure B14 of Reference 2.12-23). This change was made to better model the asymmetry in RCS flows when the two steam generator secondary side pressures are different due to crediting an ADV on one of the steam generators.

2.12.4.2 Plant Design Data Important core, RCS, and ECCS design data used in the SBLOCA analysis are listed in Tables 2.12-7 and 2.12-8.

to W3Fl-2004-0052 Page 3 of 32 The fuel rod conditions listed in Table 2.12-7 are the initial conditions for the hot rod at the bumup that produces the maximum initial stored energy. As noted in Section 2.12.4.1, the hot rod heatup analyses that determined the PCT and maximum cladding oxidation for the limiting break used values for the rod internal pressure that resulted in the maximum PCT and maximum cladding oxidation.

The HPSI system minimum delivered flow rates listed in Table 2.12-8 implement a change in the application of uncertainties associated with HPSI system flow meter accuracy and HPSI header valve position repeatability. In particular, the HPSI system flow rates used in the pre-extended power uprate SBLOCA analysis were based on applying a constant flow rate penalty over the entire range of RCS pressures to address these two uncertainties. The HPSI system flow rates listed in Table 2.12-8 are based on the fact that these two uncertainties impact the HPSI system resistance and, consequently, impact the HPSI system delivery capability as a function of the system flow rate.

2.12.4.3 Results Table 2.12-9 lists the peak cladding temperature and oxidation percentages for the spectrum of SBLOCAs. Times of interest are listed in Table 2.12-10. The variables listed in Table 2.12-11 are plotted as a function of time for each break in Figures 2.12-36 through 2.12-59. The results for the 0.055 ft2/PD break, the limiting SBLOCA, demonstrate conformance to the ECCS acceptance criteria as summarized below.

Current Parameter Criterion EPU Result Licensing Basis Peak Cladding Temperature <22000 F 20180 F 19590 F Maximum Cladding Oxidation <17% 13.1% 9.0%

Maximum Core-Wide Oxidation <1% <0.99% <0.58%

Coolable Geometry Yes Yes Yes The results are applicable to Waterford 3 for a Peak Linear Heat Generation Rate (PLHGR) of 13.2 kW/ft and a rated core power of 3716 MWt (3735 MWt including a 0.5% power measurement uncertainty).

to W3Fl-2004-0052 Page 4 of 32 Table 2.12-7 SBLOCA ECCS Performance Analysis Core and Plant Design Data Quantity J Value I Units Reactor power level (including uncertainty) 3735 MWt PLHGR 13.2 kW/ft Axial shape index -0.25 Gap conductance at PLHGR(1 ) 1536 BTUl/hr-ft 2 -OF Fuel centerline temperature at PLHGRW 3278 OF Fuel average temperature at PLHGR('1 2091 OF Hot rod gas pressure(:) 1093 psia Moderator temperature coefficient at initial density O.0x1 04 Ap/°F RCS flow rate 148.0x1 06 Ibm/hr Core flow rate 144.15x10 6 Ibm/hr RCS pressure 2250 psia Cold leg temperature 552.0 OF Hot leg temperature 615.5 OF Plugged tubes per steam generator 1000 count MSSV first bank opening pressure 1117.2 psia Low pressurizer pressure reactor trip setpoint 1560 psia Low pressurizer pressure SIAS setpoint 1560 psia HPSI flow rate Table 2.12-8 gpm SIT pressure 584.7 psia ADV opening pressure 1040 psia J.

(1) These quantities correspond to the rod average burnup of the hot rod (1000 MWD/MTU) that yields the maximum initial stored energy.

to W3F1 -2004-0052 Page 5 of 32 Table 2.12-8 HPSI Pump Minimum Delivered Flow to RCS (Assuming Failure of an EDG)

RCS Pressure, psia Flow Rate, gpm 0 787 128 745 266 698 387 655 520 605 642 556 831 473 932 423 1009 382 1078 342 1213 249 1274 196 1316 152 1355 100 1396 0 Notes:

1. The flow is assumed to be split equally to each of the four discharge legs.
2. The flow to the broken discharge leg is assumed to spill out the break.

Table 2.12-9 SBLOCA ECCS Performance Analysis Results Maximum Cladding Maximum Core-PCT Oxidation Wide Cladding Break Size (OF) (%) Oxidation (%)

0.05 ft2IPD 1959 10.04 <0.99 0.055 ft2/PD 2018 13.10 <0.99 0.06 ft2IPD 1955 2.97 <0.99 to W3F1 -2004-0052 Page 6 of 32 Table 2.12-10 SBLOCA ECCS Performance Analysis Times of Interest HPSI Flow LPSI Flow SIT Flow Delivered to Delivered to Delivered to PCT RCS (seconds RCS (seconds RCS (seconds (seconds after Break Size after break) after break) after break) break) 0.05 ft2/PD 157 (a) 1946(b) 1880 0.055 ft2/PD 145 (a) 1693 1706 0.06 ft2I/PD 134 (a) 1501 1505 (a) Calculation completed before LPSI flow delivery to RCS begins.

(b) SIT begins at this time but is not credited.

Table 2.12-11 SBLOCA ECCS Performance Analysis Variables Plotted as a Function of Time for Each Break Variable Core Power Inner Vessel Pressure Break Flow Rate Inner Vessel Inlet Flow Rate Inner Vessel Two-Phase Mixture Level Heat Transfer Coefficient at Hot Spot Coolant Temperature at Hot Spot Cladding Temperature at Hot Spot to W3Fl-2004-0052 Page 7 of 32 Figure 2.12-36 Waterford-3 Small Break LOCA ECCS Performance Analysis 0.05 ft2 /PD Break Core Power 1.50 ..... ......

1.25 W

re 0

1.00 0r a.

I-0 0.75 0

w N

0 0.50 z

0.25 0.00

_ ~.......,,,, .,.,...,.

0 100 200 300 400 500 TIME, SEC to W3F1-2004-0052 Page 8 of 32 Figure 2.12-37 Waterford-3 Small Break LOCA ECCS Performance Analysis 0.05 ft2 /PD Break Inner Vessel Pressure 2400 2000 1600 La v 1200 C,,

w 800 400 0

0 600 1200 1800 2400 3000 TIME, SEC to W3F1 -2004-0052 Page 9 of 32 Figure 2.12-38 Waterford-3 Small Break LOCA ECCS Performance Analysis 0.05 ft2/PD Break Break Flow Rate 1000 001200 X -

w CO 6100 02 ... . .... ... , . .

0 400 200 0 600 1200 1800 2400 3000 TIME, SEC to W3F1I-2004-0052 Page 10 of 32 Figure 2.12-39 Waterford-3 Small Break LOCA ECCS Performance Analysis 0.05 ft2/PD Break Inner Vessel Inlet Flow Rate 50000 .. ......... .

40000 30000 _

C.)

Co W 20000-

-J.

0 LL 10000 0

0 600 1200 1800 2400 3000 TIME, SEC to W3Fl-2004-0052 Page 11 of 32 Figure 2.1240 Waterford-3 Small Break LOCA ECCS Performance Analysis 0.05 ft2 /PD Break Inner Vessel Two-Phase Mixture Level 48 . . . . . . . ... . .

40 32 i

Co 24 TOP OF CORE ILI-I.

16 8

'.O, O, C .. .OR .E. ., . ... , .'

0 0 600 1200 1800 2400 3000 TIME, SEC to W3Fl-2004-0052 Page 12 of 32 Figure 2.1241 Waterford-3 Small Break LOCA ECCS Performance Analysis 0.05 ft2/PD Break Heat Transfer Coefficient at Hot Spot 5

10 4

10 3

U- 10 0a I

2 10 C)

I-10 0

10 0 600 1200 1800 2400 3000 TIME, SEC to W3Fl-2004-0052 Page 13 of 32 Figure 2.1242 Waterford-3 Small Break LOCA ECCS Performance Analysis 0.05 ft2 /PD Break Coolant Temperature at Hot Spot 2000 1700 1400 0

us w

D EL 800 500 200 0 600 1200 1800 2400 3000 TIME, SEC to W3Fl-2004-0052 Page 14 of 32 Figure 2.12-43 Waterford-3 Small Break LOCA ECCS Performance Analysis 0.05 ft2/PD Break Cladding Temperature at Hot Spot 2200 1900 1600 Lu I-1300 w

r IL 1000 700 400

,1,.. . . ,....

0 600 1200 1800 2400 3000 TIME, SEC to W3F1 -2004-0052 Page 15 of 32 Figure 2.1244 Waterford-3 Small Break LOCA ECCS Performance Analysis 0.055 ft2IPD Break Core Power 1.50 ...

1.25 1.00 0

I-j 0 0.75 w

N 0 0.50 z

0.25 0.00 0 100 200 300 400 500 TIME, SEC to W3Fl-2004-0052 Page 16 of 32 Figure 2.12-45 Waterford-3 Small Break LOCA ECCS Performance Analysis 0.055 ft2 /PD Break Inner Vessel Pressure 2400 2000 1600 co 0-rL 1200 CI) w a.

800 400 0

0 600 1200 1800 2400 3000 TIME, SEC to W3FI-2004-0052 Page 17 of 32 Figure 2.1246 Waterford-3 Small Break LOCA ECCS Performance Analysis 0.055 ft2/PD Break Break Flow Rate 1200 1000 800

-j 600 LL 400 200 0

0 600 1200 1800 2400 3000 TIME, SEC to W3Fl-2004-0052 Page 18 of 32 Figure 2.12-47 Waterford-3 Small Break LOCA ECCS Performance Analysis 0.055 ft2/PD Break Inner Vessel Inlet Flow Rate 50000 40000 30000 C.)

W uj oD

-J 0 20000 co

-J LL 10000 0

-10000 0 600 1200 1800 2400 3000 TIME, SEC to W3Fl-2004-0052 Page 19 of 32 Figure 2.1248 Waterford-3 Small Break LOCA ECCS Performance Analysis 0.055 ft2 /PD Break Inner Vessel Two-Phase Mixture Level 48 ~~. ........ .... ;j ... ,, j.........

,i; .. ;..i, 40 32

-A mu 24 LU TOP OF CORE 0l) 16 8

~~~. .... ..... ... ,.... ...... ...... . .......... . .

0 0 600 1200 1800 2400 3000 TIME, SEC to W3F1 -2004-0052 Page 20 of 32 Figure 2.12-49 Waterford-3 Small Break LOCA ECCS Performance Analysis 0.055 ft2 /PD Break Heat Transfer Coefficient at Hot Spot 5

10 4

10 3

IL 0o 10 I

I-X3 I- 2 10 M

1 10 0 ..., ....

10 C 600 1200 1800 2400 3000 TIME, SEC to W3F1-2004-0052 Page 21 of 32 Figure 2.12-50 Waterford-3 Small Break LOCA ECCS Performance Analysis 0.055 ft2/PD Break Coolant Temperature at Hot Spot 2000 1700 1400 LL 0

uS W-1100 a:

0-w 800 500 200 0 600 1200 1800 2400 3000 TIME, SEC to W3Fl-2004-0052 Page 22 of 32 Figure 2.12-51 Waterford-3 Small Break LOCA ECCS Performance Analysis 0.055 ft2/PD Break Cladding Temperature at Hot Spot 2200 .....

1900 1600 LL 0

Li c:

D 1300 w

1000 700 400 0 600 1200 1800 2400 3000 TIME, SEC to W3Fl-2004-0052 Page 23 of 32 Figure 2.12-52 Waterford-3 Small Break LOCA ECCS Performance Analysis 0.06 ft2/PD Break Core Power 1.50 1.25 W

1.00 0

a.

-j 0

0.75 0

N I-0 z 0.50 0.25 0.00 0 100 200 300 400 500 TIME, SEC to W3Fl-2004-0052 Page 24 of 32 Figure 2.12-53 Waterford-3 Small Break LOCA ECCS Performance Analysis 0.06 ft2 /PD Break Inner Vessel Pressure 2400 2000 1600 La U) co CD 1200 CO w

0:

800 400 0

0 360 720 1080 1440 1800 TIME, SEC to W3F1 -2004-0052 Page 25 of 32 Figure 2.12-54 Waterford-3 Small Break LOCA ECCS Performance Analysis 0.06 ft2IPD Break Break Flow Rate 1200 1000 800 V.

w QCO 600

-J U-400 200 0

0 360 720 1080 1440 1800 TIME, SEC to W3Fl-2004-0052 Page 26 of 32 Figure 2.12-55 Waterford-3 Small Break LOCA ECCS Performance Analysis 0.06 ft2 /PD Break Inner Vessel Inlet Flow Rate 50000 40000 30000

-j 2 20000 0

-j 10000 0

-10000 0 360 720 1080 1440 1800 TIME, SEC to W3F1 -2004-0052 Page 27 of 32 Figure 2.12-56 Waterford-3 Small Break LOCA ECCS Performance Analysis 0.06 ft2 IPD Break Inner Vessel Two-Phase Mixture Level 48 40 32

-F ci 24 I TOP OF CORE- - -

Orm 0T 16 BOOM OF CORE. . ..

8 0

0 360 720 1080 1440 1800 TIME, SEC to W3F1 -2004-0052 Page 28 of 32 Figure 2.12-57 Waterford-3 Small Break LOCA ECCS Performance Analysis 0.06 ft2/PD Break Heat Transfer Coefficient at Hot Spot 5

10 4

10 3

IL 0,

10 I

I-2 10 I-M 10 1010 0 360 720 1080 1440 1800 TIME, SEC to W3Fl-2004-0052 Page 29 of 32 Figure 2.12-58 Waterford-3 Small Break LOCA ECCS Performance Analysis 0.06 ft2/PD Break Coolant Temperature at Hot Spot 2000 --..

1700 1400 LI 0

Li 1100 a-w 800 500 200 0 360 720 1080 1440 1800 TIME, SEC to W3F1 -2004-0052 Page 30 of 32 Figure 2.12-59 Waterford-3 Small Break LOCA ECCS Performance Analysis 0.06 ft2/PD Break Cladding Temperature at Hot Spot 2200 1900 1600 Li 0

w n.-

13000 1000 700 400 0 360 720 1080 1440 1800 TIME, SEC to W3Fl-2004-0052 Page 31 of 32 2.12.8 References 2.12-3 Waterford 3 Final Safety Analysis Report through Revision 12-B.

2.12-7 CENPD-134P, COMPERC-ll, A Program for Emergency-Refill-Reflood of the Core, August 1974.

CENPD-1 34P, Supplement 1, COMPERC-II, A Program for Emergency Refill-Reflood of the Core (Modifications), February 1975.

CENPD-1 34, Supplement 2-A, COMPERC-II, A Program for Emergency Refill-Reflood of the Core, June 1985.

2.12-10 CENPD-138P, PARCH, A FORTRAN-IV Digital Program to Evaluate Pool Boiling, Axial Rod and Coolant Heatup, August 1974.

CENPD-1 38P, Supplement 1, PARCH, A FORTRAN-IV Digital Program to Evaluate Pool Boiling, Axial Rod and Coolant Heatup (Modifications), February 1975.

CENPD-1 38, Supplement 2-P, PARCH, A FORTRAN-IV Digital Program to Evaluate Pool Boiling, Axial Rod and Coolant Heatup, January 1977.

2.12-11 CENPD-135P, STRIKIN-Il, A Cylindrical Geometry Fuel Rod Heat Transfer Program, August 1974.

CENPD-1 35P, Supplement 2, STRIKIN-Il, A Cylindrical Geometry Fuel Rod Heat Transfer Program (Modifications), February 1975.

CENPD-1 35, Supplement 4-P, STRIKIN-Il, A Cylindrical Geometry Fuel Rod Heat Transfer Program, August 1976.

CENPD-1 35-P, Supplement 5, STRIKIN-Il, A Cylindrical Geometry Fuel Rod Heat Transfer Program, April, 1977.

2.12-12 CENPD-139-P-A, C-E Fuel Evaluation Model, July 1974.

CEN-161(B)P-A, Improvements to Fuel Evaluation Model, August 1989.

CEN-1 61 (B)-P, Supplement 1-P-A, Improvements to Fuel Evaluation Model, January 1992.

2.12-13 0. D. Parr (NRC) to F. M. Stern (CE), June 13, 1975.

2.12-16 K. Kniel (NRC) to A. E. Scherer (CE), Combustion Engineering Emergency Core Cooling System Evaluation Model, November 12,1976.

2.12-17 R. L. Baer (NRC) to A. E. Scherer (CE), Evaluation of Topical Report CENPD-135 Supplement No. 5, September 6, 1978.

to W3Fl-2004-0052 Page 32 of 32 2.12-18 K. Kniel (NRC) to A. E. Scherer (CE), Evaluation of Topical Report CENPD-138, Supplement 2-P, April 10, 1978.

2.12-20 O. D. Parr (NRC) to F. M. Stern (CE), December 4,1974.

2.12-21 A. C. Thadani (NRC) to A. E. Scherer (CE), Acceptance for Generic Referencing of the Topical Report CEN-161 "Improvements to Fuel Evaluation Model (FATES3)," May 22, 1989.

2.12-22 A. C. Thadani (NRC) to A. E. Scherer (CE), Generic Approval of C-E Fuel Performance Code FATES3B (CEN-161(B)-P, Supplement 1-P) (TAC No. M81769),

November 6, 1991.

2.12-23 CENPD-137P, Calculative Methods for the C-E Small Break LOCA Evaluation Model, August 1974.

CENPD-1 37, Supplement 1-P, Calculative Methods for the C-E Small Break LOCA Evaluation Model, January 1977.

CENPD-1 37, Supplement 2-P-A, Calculative Methods for the ABB C-E Small Break LOCA Evaluation Model, April 1998.

2.12-24 K. Kniel (NRC) to A. E. Scherer (CE), Evaluation of Topical Reports CENPD-133, Supplement 3-P and CENPD-137, Supplement 1-P, September 27, 1977.

2.12-25 T. H. Essig (NRC) to 1.C. Rickard (ABB), Acceptance for Referencing of the Topical Report CENPD-137(P), Supplement 2, Calculative Methods for the C-E Small Break LOCA Evaluation Model" (TAC No. M95687), December 16, 1997.

2.12-26 CENPD-133P, Supplement 1, CEFLASH-4AS, A Computer Program for the Reactor Blowdown Analysis of the Small Break Loss of Coolant Accident, August 1974.

CENPD-1 33, Supplement 3-P, CEFLASH-4AS, A Computer Program for the Reactor Blowdown Analysis of the Small Break Loss of Coolant Accident, January 1977.

2.12-30 B. Houston (Entergy) to Document Control Desk (NRC), Small Break Loss-of-Coolant Accident Emergency Core Cooling System Performance Analysis, Waterford Steam Electric Station, Unit 3, Docket No. 50-382, License No. NPF-38, May 26, 2004.

Attachment 6 W3FI-2004-0052 Miscellaneous Power Uprate Report Changes

Waterford 3 Extended Power Uprate EXECUTIVE

SUMMARY

The core power of Watertord Steam Electric Station, Unit 3 (Waterford 3) Is being Increased In an extended power uprate (EPU) from 3441 megawatts thermal (MWt) to 3716 MWt to increase the facility's electrical power output. The EPU will be implemented in Cycle 14 in Spring 2005. This Powor Uprato Report (PUR) demonstrates acceptable facility operation at the increased power level. This PUR closely follows the guidance of NRC draft Review Standard RS-001, 'Review Standard for Extended Power Uprates.' December 2002. All items in the review standard and all acceptance criteria have been addressed to the extent they are consistent with the Waterford 3 licensing basis.

Plant modifications planned for implementation to support the power uprate include:

  • The high pressure turbine will be upgraded.
  • The generator will be rewound and associated auxiliaries will be provided.
  • Higher capacity main generator output circuit breakers, disconnect switches, and bus work will be installed.
  • Control valves for the heater drain system may be upgraded/replaced as necessary.
  • Instrument and cntrul changes (e.g., lower steam generator pressure trip setpoint) will be made as necessary.
  • elfsystem 0 et valv e reabdequa pac_.
  • Condenser modifications will be performed as required to prevent potential condenser tube vibration.

Methodology changes employed for the EPU include:

  • The CENTS computer code was used rather than CESEC for non-LOCA safety analyses.
  • The 1999 Combustion Engineering Nuclear Power Large-Break Emergency Core Cooling System Evaluation Model was used rather than the June 1985 version of the Evaluation Model.
  • Previously approved leak-betore-break (LBB) methodology was applied to eliminate consideration of mechanical (dynamic) effects of Reactor Coolant System pipe breaks.

Other methodology and plant changes are Identified In the PUR Intruduction.

To ensure protection of the health and safety of plant personnel and the public following power uprate, the Waterford 3 design basis and accident analyses were reviewed with respect to changes to plant operating conditions and characteristics. The analyses and evaluations performed to support the uprate are described in this report. Evaluations of structural integrity and functional performance of plant components and systems demonstrate acceptable performance at uprate conditions. LOCA analyses credit automatic operation of the safety grade ADVs to mitigate the consequences of the small break LOCA. The LOCA analyses x

636dcl1/50 6306.doc-1 1/05/03 xxv

Waterford 3 Extended Power Uprate planned programmed Twd is increased from 5410 F at HZP to 5430 F at HFR Nominal temperature conditions are based upon a nominal RCS flow of 110% of the minimum design flow. Reviews of RCS flow data conducted in association with implementation of the 2002 Appendix K Power Uprate demonstrated that best estimate RCS flow was 110% of design; previously nominal flow was assumed to be 107%.

1.3 EPO APPROACH This EPU application closely follows the guidance of the Office of Nuclear Reactor Regulation (NRR) draft RS-001, 'Review Standard for Extended Power Uprates." December 2002, to the extent that the review standard is consistent with the Waterford 3 licensing basis. FSAR sections that discuss Waterford 3 features addressed in this PUR are identified. In addition, the Arkansas Nuclear One, Unit 2 (ANO-2) EPU application, including Requests for Additional Information (RA19) and as3ociated reaponses, were reviewed for guidance. ANO-2 is also a CE NSSS design plant, and the approximate 7.5 % ANO-2 power uprate compares closely to the approximate 8 % uprate requested for Waterford 3. The ANO-2 application was approved by the NRC's Safety Evaluation Report (SER) dated April 24. 2002.

Entergy plans to implement the Waterford 3 EPU in one increment. Completion of plant modifications necessary to implement the EPU is planned prior to the end of refueling outage 13 in spring 2005. The plant will be operated at 3716 MWt starting in cycle 14.

1.4 PLANT MODIFICATIONS Among the plant modifications planned for implementation prior to the end of refueling outage 13 to permit implementing the EPU in cycle 14 are:

  • The high pressure turbine will be upgraded.
  • The generator will be rewound and associated auxiliaries will be provided.
  • Higher capacity main generator output circuit breakers, disconnect switches, and bus work will be installed.
  • Control valves for the heater drain system may be upgraded/replared as necacssry.
  • Instrument and control changes (e.g., lower steam generator pressure trip setpoint) will be made as necessary.
  • Condenser modifications will be performed as required to prevent potential condenser tube vibration.

Evaluations of these modifications are provided in Section 2.0 of this report 6306.doc-1 11/05/03 1-2

Waterford 3 Extended Power Uprate parameters. Therefore. the evaluation prior to Cycle 14 will account for the changes in temperature on crack initiation and growth and the reduction u the allowed primary-to-secondary leakage rate propose for ec ni i cation Section 3.4.5.2, 'Operational Leakage.' Utilizing this information with the inspection results will determine EOG conditions which are compared to the performance criteria for burst and leakage. This will determine if the plant can operate full cycle or will be limited based on the projected degradation.

Implementation of the NEI 97-06 program ensures that changes in the conditions of the steam generator tubing that are the result of power uprate will be identified and addressed.

Repair techniques (e.g., plugs, sleeves, etc.) are addressed in Section 2.2.2.1.4.6.2 of this report.

Conclusions Steam generator tube integrity will continue to be maintained in accordance with NEI 97-06 following the EPU.

2.1.10 Steam Generator Blowdown System The Steam Generator Blowdown System (SGBS) is described in FSAR Section 10.4.8.

The SGRS is designed to fulfill the following requirements:

a) To maintain the steam generator shell side water chemistry by maintaining continuous blowdown during normal plant operating conditions.

b) To direct the blowdown to the blowdown demineralizers for further treatment.

c) To achieve and maintain the chemistry requirements of the water inventory in the Condensate and Feedwater Systems (CFWSs) prior to introduction of feedwater into the steam generators during plant start-up and operation.

The EPU roview examined the effect of the higher foodwater flow rato on the SGBS.

Evaluation Typical operational blowdown rates during normal operation are approximately 1% of current feedwater flow. The blowdown system is sized to handle 2% of the original rated flow or 650 gpm. Although the feedwater flow will be increasing as a result of EPU, the capacity of the SGBS under EPU conditions will still be adequate to maintain chemistry In the secondary system.

Conchirlon There are no design basis changes in the SGBS. The system will continue to meet the requirements of GDC 14 (secondary water chemistry), and satisfactorily perform the functions listed above under EPU conditions.

6306.doc-l 1/05103 2.1-13

Waterford 3 Extended Power Uprate for prG-uprate was 5.5 ksi vs. an allowable of 29.1 ksi. The uprate will increase the AP across the deflector plate by approximately 20%. Since the stress on the plate is directly proportional to the AP, the stresses will also increase approximately 20% to a value of 6.6 ksi. However, this stress is well below the Code limit of 29.1 ksi. Thus, the flow deflector is acceptable for the EPU.

2.2.2.1.4.3.5 Bottom Blowdown Pipe The bottom blowdown pipe evaluation for pre-uprate Is not affected by the lower secondary side pressure or the lower secondary side temperature for power uprate. Therefore. the previous analysis of the bottom blowdown pipe is fully applicable to EPU.

2.2.2.1.4.3.6 Tube and Tube Supports The pre-uprate evaluations of the tube and tube supports are based on design parameters that do not change for EPU. Parameters that affect the interaction between the tubes and the vertical grid supports are the coefficient of thermal expansion and the average temperatures between the reactor coolant fluid within the tube and the external steam temperature. Because the T," temperature for power uprate is lower than the temperature used in the AOR for pre-uprate, and the rmains the same, it is concluded that the pre-uprate conditions bound the PU conditions.

y~sTeX; he r;ma ry /i(

2.2.2.1.4.3.7 Tube and Tubesheet Weldj=0oq.o Zcc91 The variables used in the calculation of stresses in this component are:

(1) The differential primary pressure-to-secondary pressure (2) The mean tubesheet temperature-to-primary temperature The change in temperature for EPU represents a very small change in the mean tubesheet temperature and therefore an insignificant change in stress. The changes In primary-to-secondary pressure. however. were evaluated in detail for EPU for the following loadings:

A) Design B) Primary leak test C) Primary hydro D) Secondary hydro E) Secondary leak test F) Normal operation For EPU conditions, only loading conditions A and F change. Using a conservative multiplier of 1.147, which is based on the maximum ratio of power uprate to pre-uprate delta between primary and secondary presurea end is applied to the maximum atreaa for loadings A and F 221 63W.dw-I 1105103 6306.doc-1 1/05/03 2.2-13

Waterford 3 Extended Power Uprate reactor. This system also furnishes power for a safe shutdown of the plant during normal and emergency conditions. Should an accident occur and preferred power be lost in either or both divisions, this system acts as a source of power for safety related DC loads and static uninterruptible power supplies (SUPSs) until emergency power is available.

The non-safety related function of the system is to power the non-safety related loads required to ensure instrumentation and control capability to monitor and maintain the plant status during startup, shutdown. normal, and emergency plant operation. Loads which are sensitive to voltage loss/fluctuations andlor frequency fluctuations are served by the DC or SUPS systems.

Examples include computer, communications, and security equipment. In addition, loads which are critical upon loss of offsite power, such as turbine bearing lube oil pumps, are fed from these systems.

Evaluation The 125-V C distribution s tem is disc ed in FSAR Section 8.3.2. and was evaluated for component operation or battery duty cycles due to te EPU.

Conclusion The 125-VDC distribution system will continue to function as designed. Adequate separation exists, and the system has the capability to continue to supply adequate power to both safety and non-safety equipment. The system will continue to meet the requirements of GDC 17 following implementation of the EPU.

2.3.5 Station Blackout The station blackout evaluation is discussed in FSAR Section 8.1A.

Station blackout encompasses the complete loss-of-offsite electric power concurrent with a turbine trip and failure of the onsite emergency AC power system. Plant equipment power from the DC onsite power system and steam from the SGs will be used to remove decay heat from the reactor core.

Evaluation The turbine-driven Emergency Feedwater System (EFS) pump, the atmospheric dump valves.

and DC electric power from Class 1E batteries are available for reactor core decay heat removal during the 4-hour event. The turbine-driven pump draws water from the condensate storage pool and pumps it into the SGs. Steam from the SGs is used for operating the pump's turbine. The batteries provide electric power to the steam supply valves and the EFS flow control valves. The suction lines to the turbine-driven pump are aligned to the condensate storage pool.

The atmospheric dump valves (ADVs) are used for bleeding steam from the SGs. Decay heat frnm the reafctor mrm is removed by feeding and hIeding the SGs. Approximately 82,200 gallons of water is needed to remove decay heat for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> in accordance with 2.3-7 1105)03 6306.doc.i 1/05103 6306.doc-I 2.3-7

Waterford 3 Extended Power Uprate 2.4 INSTRUMENTATION AND CONTROLS 2.4.1 Reactor Protection. Safety Features Actuation, and Control Systems The Plant Protection System (PPS) is composed of two sub-systems: the Reactor Protection System (RPS) and the Engineered Safety Features Actuation System (ESFAS).

The RPS is designed to trip the reactor by de-energizing the control element drive mechanism (CEDM) coils whenever any monitored condition reaches a trip setpoint. For each measured variable, the RPS uses a 2 of 4 channel logic arrangement, wtn each channel electrically and physically separated to ensure no loss of functionality with a single failure. The RPS is designed to protect the reactor core and the reactor pressure boundary during defined anticipated operational occurrenccs, and to ascist in mitigating tho consoquencos of certain design basis accidents. The RPS also includes the Core Protection Calculator System (CPCs),

the reactor trip switchgear (RTSG) and the Reactor Coolant Pump Shaft Speed Sensor System (RCPSSSS). The RPS is discussed in FSAR Section 7.2.

The ESFAS is designed to initiate safety features whenever any monitored condition reaches a trip setpoint. Like the RPS, the ESFAS uses a 2-of-4 channel logic arrangement, with each channel electrically and physically separated to ensure no loss of functionality with a single failure. The ESFAS is designed to mitigate the consequences of certain design basis accidents (DBAs). particularly by protecting the containment building integrity. The ESFAS is discussed in FSAR Section 7.3.

Systems required for safe shutdown are described in FSAR Section 7.4. Instrumentation required to monitor, control, and provide interlocks for these systems is described in FSAR Sections 7.4, 7.5. and 7.6. The EPU affects the process ranges of some of the instrumentation described in these sections.

E. KA=D err. n_

Waterford 3 provides anticipated transient without scram (ATWS) mitigation systems required by IOCFR50.62 as described in FSAR Section 7.B.

2.4.1.1 Evaluation The PPS is affected by EPU. The low steam generator pressure trip setpoint has decreased.

The maximum allowable linear power level - high trip setpoints with inoperable steam line safety valves are adjusted accordingly for EPU conditions. No other setpoints affecting the PPS will change due to EPU.

The EPU redefines 100% power level, affecting changes to the neutron flux to percent power correlation of the ex-core nuclear instruments. The ex-core nuclear instruments provide signals to the PPS for high linear power level and high logarithmic power level reactor trips. The correlation Is ochiovod by frequent secondary calorimetric calibrations, which norrnaiizo the flux signals to calorimetric power.

2.4-1 1(05/03 6306.doc-1 1105103 6306.doc-1 2.4-1

Waterford 3 Extended Power Uprate Power uprate will result in different operating conditions in the SGs, which affects the calibration of the SG level instruments. The SG low and high level limits will remain protected.

The core protection calculators (CPCs) receive input signals, Including neutron flux from the ex-core instruments, and calculate local power density (LPD) and departure from nucleate boiling ratio (DNBR). Power uprate will affect the ex-core signals as addressed above, but changes to the neutron flux to power correlation will be corrected by the calorimetrics. As part of EPIJ, the rPC differential cold leg tempprature trip Rrtpoint was reduced to compensate for more negative MTC and the higher EPU power level, as discussed in Section 2A3.9.1.A. As pait of EPU, CPC response times have been clarified in the Technical Requirements Manual (TRM).

The changes to ranges on instrumentation supporting safe shutdown will not change the instrument functions as described in FSAR Sections 7.4 and 7.5. The setpoints for shutdown system interlocks and plant processes having significant impact on plant safety described In FSAR Section3 7.4 and 7.6 are not affected by the EPU.

The EPU does not change the design of AIWS mitigation systems as described in FSAR Section 7.8. However, because the EPU lowers steam generator and main steam operating pressures, the diverse emergency feedwater actuation signal (DEFAS) permissive setpoint will be lowered; this change ensures that the configuration for the Emergency Feedwater System (EFS) actuation described in Section 7.8 is maintained under EPU conditions. Additionally, other actuation setpoints remain such that ATWS mitigation systems will not be actuated before initiation of the PPS.

The EPU does not change the separation, redundancy, or diversity of the above instrumentation and controls as described in the applicable sections of the FSAR.

2.4.1.2 Conclusion The EPU configuration has been analyzed to a lower SG pressure. Changes to the Plant Protection System due to lower SG pressure will be implemented for EPU.

No setpoint changes to safe shutdown interlocks or alarms are required.

CPC constants will be updated to be consistent with the new definition of 100% power and other requirements of the EPU and cycle-specific analysis.

The EPU does not change the safety functions or design such as separation. redundancy. and diversity of the instrumentation as described in FSAR Sections 7.2, 7.3, 7.4, 7.5, 7.6, and 7.8.

The changes resulting from the EPU are, therefore, consistent with the licensing basis and comply with acceptance criteria related to 10CFR49, 10CFR50.55a, 10CFR62, 10CFR50 App.AGOCs 1,2,3.4,10,12,13,19,20,21,22,23,24,25, 29and 10CF5bOApp. R.

6306doc- 110/03 .4-6306.doc-1 1/05/03 2.4-2

Waterford 3 Extended Power Uprate b) Superheats the HP turbine exhaust to Improve the thermodynamic efficiency of the steam cycle using the MSSS as the heat source.

c) Limits the steam flow to the LP turbine following a load rejection thereby protecting the main turbine from an overspeed condition.

The EPU review examined the effects of extended power uprate (most notably increased stparm mass flow) on the system's capability to transport steam to the power conversion system, provide heat sink capacity, supply steam to drive the EFPT, and withstand adverse dynamic loads. Overpressure protection capabilities were also reviewed with respect to new power uprate conditions.

Evaluation The design temperatures and prcccurcs of tho MSSS arc not impacted by the EPU. Operating temperatures and pressures for the MSSS from the SG to the turbine have been reduced at normal operating power levels, while steam mass flow has been increased (which results in a higher steam velocity). MSSS flow at normal operating power levels has increased by approximately 8.6%. In order to accept the higher MSSS flows for EPU, the HP turbine steam path will be replaced. Pressures and temperatures at some intermediate points In the power cycle (most notably extraction steam) have increased slightly due to lower pressure losses and improved turbine efficIencIes. The MSSS piping has been evaluated and has been shown to be acceptable in its current configuration. Lines with increased velocities above design guidelines are infrequently operated or have been recommended for inclusion in the Flow Accelerated Corrosion Program. The MSSS piping has been evaluated for increased LOCA movements at SG nozzles and transient loading as a result of a turbine trip. The results indicate both piping and nozzles are acceptable.

Closure times for MSIVs have been evaluated and have been determined to be unaffected by the EPU.

The ADVs arc required to maintain reactor coolant temperature during LOOP, and are al3o credited for the small-break LOCA (SBLOCA) event. ADV capacities are incorporated into the accident and transient analyses. These analyses indicate that sufficient heat re'ection capability exists. The ADV opening setpoint is being lowered to less than or psia.

pal The main steam safety relief valve capacity has been reviewed and has nd to be sufficient to pass the maximum rated plant steam flow at set pressure plus accumulation (3%).

The main steam safety relief valves, therefore, have sufficient capacity at EPU conditions to perform their design basis function.

Hot-zero power (HZP) steam pressures are reduced slightly by the EPU, however, no changes to EFW pump nowrates or operating conditions are required, nor are any EFW turbine operating conditions adversely affected. Decay heat has increased, increasing the EFW volume required, however, the flow rate is within the design capacity of the EFW pumps.

The MSSS provides steam for the feedwater pump turbines during startup. The EPU requires that the feedwater flow (and thus steam flow to the feedwater pump turbines) be increased to meet the new steam now requirements. No modifications are required to the turbine since the rated horsepower of the turbine exceeds the calculated horsepower requirements (to deliver 2.5-32 1105/03 6306.doc-1 1/05/03 n3o6.doc-1 2.5-32

Waterford 3 Extended Power Uprate the required feedwater flow) at EPU conditions. The Feedwater Control System setpoints will be modified slightly to increase pump speed at a lower demand. The remaining current Feedwater Control System setpoints were used in the evaluation and were found to be acceptable at EPU conditions.

The MSSS design bases are described in FSAR Section 10.3.1. EPU conditions do not affect the existing design basis of the MSSS or alter the method In which it performs these functions.

An additional safety related function has been added, in that the ADVo are crodited for secondary pressure control during the small break LOCA event. The sizing of these valves Is adequate for EPU conditions.

The MSSS has been evaluated with respect to I OCFR5O. Appendix A. GDC 34, Residual Heat Removal. The MSSS is capable of providing heat sink capacity and pressure relief capability and supplying steam to the steam-driven EFW pump at EPU conditions.

No additional internally generated missiles or pipe break locations are identified in the MSSS as a result of evaluation at EPU conditions, nor have any additional effects from natural phenomena such as tornadoes. earthquakes, or hurricanes been postulated. Evaluiation of missile generation and protection Is addressed in Section 2.5.12 of this report. The MSSS piping analysis incorporates the effects of dynamic loads due to LOCA and sudden turbine stop valve closure. Based upon the above; continued compliance with IOCFR50, Appendix A, GUC; 2, Design Bases for Protection Against Natural Phenomena and GDC 4, Environmental and Missile Design Bases is assured.

it = Y MSR and othor cystom pi ing and The turbine steam valve closure time following the detection of a turbine overspeed condition as described in FSAR Section 10.2.2.2.9 is not impacted by the EPU.

Tho Stoam Bypass System (SBS) is discussed in Section 2.5.6.3.

Conclusion Safety-related and non-safety related components and sub-systems in the MSSS as desg'be in the Waterfo d 3 FSAR are adequately size and designed (with ADV setpoinVgo modifications e acon( to perform their intended function at EPU conditions for normal, transient, and accident conditions. The MSSS will maintain its ability to transport steam to the power conversion system, provide heat sink capacity, supply steam to steam-driven safety pumps, and withstand steam hammer. The system continues to comply with the requirements of GOCs 2. 4 and 34.

2.5.6.2 Maln Condenser The main condenser is discussed in FSAR Section 10.4.1 and FSAR Table 10.4-1. The three-shell, single-pressure, main condenser provides a continuous heat sink for the exhaust from the three tandem-compound LP turbines and for miscellaneous flows, drains, and vents during normal plant operation.

2.5-33 MA6.oc-1 1/05103 6306.doc-1 1/05/03 2.5-33

Waterford 3 Extended Power Uprate 2.9 HUMAN PERFORMANCE 2.9.1 Channns in Emrnemncy and Abnormal Operatinn Procedures Existing procedures will adequately cover emergency scenarios, abnormal occurrences, or normal operations. New procedures are not expected to be required. No new operator actions, as a result of the proposed EPU, are anticipated.

Setpoints that are mentioned in emergency operating procedures or off normals will be changed to now uprato voluc3.

Specific, non-setpoint, changes to existing steps in emergency and off-normal operating procedures for power uprate follow:

The time window to establish simultaneous hot and cold leg injection in the loss-of-coolant accident (LOCA) emergency operating procedure will be changed from 2 to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to 2 to 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />.

As discussed in Section 2.5.2.4, Entergy is submitting a separate license amendment request to eliminate the Technical Specificatinn requirements for combustible gas control in containment. Resulting Emergency Operating Procedures (EOP) changes, regarding operation of the combustible gas control system, will be made during the implementation phase following approval and issuance of the associated amendment.

Conclusion The change in the hot/cold leg injection window between 2-3 hours provides no challenge to operator performance and will be satisfactorily implemented by the procedure change process and operator training.

2.9.2 Changes to Operator Actions Sensitive to Power Uprate ThoAtmo3phcric Dump Volvc (ADV) controllril have 'a Technical Specification-mand atwet Operators will be required to perform channel chkcks once.per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to. ensure ADV automatic actuation operability when operating above 70% rated thermal power Conclusion Thtse diannel checks will have no impact on operator response times.

2.9.3 Changes to Control Room Controls. Displays and Alarms The instrument loops affected by the extended power uprate (EPU) are listed in Table 2.9-1 below. The associated changes to the control room controls, displays, and alarms affected by the FlPU are described in the table below.

2.9-i 6306.doc.1 6306.Woc-1 1/05/03 1/05/03 2.9-1

Waterford 3 Extended Power Uprate Operators have been involved in the conceptual design, detailed design, and review of the modification packages making the Indicated changes. Multiple layers of operator involvement are in place to assure and reinforce operator cognizance of these changes.

Tppis stoteVmod f~~n that up~grdoo onch 4o ADV rolerm fro n ana digital t f-containe digital propditional int derivativ (PID) unit hese controUfrs will bepangd to iprove the a~ceracy of setSg the ADV Pftrol setpo0 .

Conclusion See the conclusion to 2.9.4.

2.9.4 Changes to the Qualified Safety Parameter Display System Planned changes to the Qualified Safety Parameter Display System (QSPDS) are summarized in Table 2.9-1 below. As discussed in Section 2.9.3 above, multiple layers of operator Involvement are in place to assure and reinforce operator cognizance of these changes.

Table 2.9-1 Summary of Changes Description Before EPU After EPU RPS Channel Trip SG Alarm Setpoint: 764 psia Alarm Setpoint: 662 psia Pressure Feedwater Mass Flow Indicated Range: Indicated Range:

0 - 8.0 MPPH 0 - 9.0 MPPH Main Steam Mass Indicated Range: Indicated Range:

Flow 0 - 8.0 MPPH 0 - 9.0 MPPH ADV Control Analog with setpoint adjustment resolution of 60 psi. ust ent r luti of s@

Setpoint: 1050 psig SBCS Controller Indicated Range: Indicated Range:

Process Range 800- 1050 psia 750- 1050 psia QSPDS Cold Leg Alarm Setpoint: A.arm Setpoint:

Temperature 558°F Color banding will change for the following indicated parameters: A. /7 7 z a) Pressurizer pressure narrow ranger 54+

b) Pressurizer pressure wide range Ant4 wi7 h ,.veA/0 S .n Ja<we c) RCScoldlegtemperature res alI te f 1i f; 1;li;-i_

d) RCS hot leg temperature / Ahr plet'7> WO.;a4r :m 4 r.

e) SIT levels Se4- °; vi t IT ?f 2 P e5 ; 3 2.9-2 6306.doc.1 6am6 1/05/03 bc-1 1/05/03 2.9-2

Waterford 3 Extended Power Uprate Conclusion The changes summarized in Table 2.9-1 will provide the control room operator the information required to achieve and maintain safe shutdown of the reactor during normal and accident conditions. Training and tasting of the operators regarding the use of the changed displays is addressed in Section 2.9.5 of this report ensuring that control room operators will use the instrumentation reliably The control room configuration after implementation of the EPU will therefore remain consistent with GDC 19. and SRP Section 18.

2.9.5 Changes to the Operator Training Program and the Control Room Simulator 2.9.5.1 Operation Training The training staff will provide training as determined by the Operations Training Review Group.

Prior to EPU Implementation, the simulator will be upgraded to provide operator training at EPU conditions. Operators will receive training using the systematic approach to training process. This training will cover procedure changes, new or revised technical specifications/safety analysis, and equipment modifications for the EPU. The topics will include items such as high pressure turbine upgrade, )atmospheric dump valvee and instrument range/alarm changes. Classroom training/tAosting and simulator troining/ttin with the new equipment/instrument changes installed will ensure the operators understand e changes to plant systems.

Startup training will be conducted prior to the conclusion of the outage for the operations crews with emphasis on core reload, positive moderator temperature coefficient, reactor engineeving interface, and teamwork skills.

2.9.5.2 Simulator Modification Plant modifications including the power uprate modifications and refueling outage plant modifications are reviewed for impact on the simulator. Plant modifications that affect the primary, secondary, control systems logic or dynamic response are incorporated into the simulator. H-ardware control board changes trat affect the simulator physical fndelity are incorporated into the simulator. Plant computer and display changes are updated on the simulator. New or modified instructor station malfunction, remote functions, and overrides along with panel and piping graphics are updated to reflect the modifications.

Verification and validation testing is performed on the simulator modifications. After the process model modifications are implemented and core reload parameters are installed, simulator initial conditions are established at various power levels and core life. Simulator testing occurs after the 100% power initial conditions have been established.

) i 4 iC ;cfr4-o Co 2.9-3 6308.doc-1 1/05/03 6306.doc1 1/05/03 2.9-3

Waterford 3 Extended Power Uprate Result: The ECCS performance analysis calculated a maximum cladding oxidation of times the total cladding thickness before oxidation for thea Criterion 3: Maximum Hydrogen Generation: The calculated total amount of hydrogen generated from the chemical reaction of the cladding with water or steam shall not exceed 0.01 times the hypothetical amount that would be generated If all of the metal in the cladding cylinders surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react.

Rooult: The ECCS performance analysal calculated a maximum hydrogen generation of less than 0.01 times the hypothetical amount for the 0.8 DEG/PD break.

Criterion 4: Coolable Geometry: Calculated changes In core geometry shall be such that the core remains amenable to cooling.

Result: The cladding swelling and rupture models used in the ECCS performance analysis account for the offects of changes In core goomotry that would occur if cladding rupture is calculated to occur. Adequate core cooling was demonstrated for the changes In core geometry that were calculated to occur as a result of cladding rupture. In addition, the transient analysis was performed to a time when cladding temperatures were decreasing and the RCS was depressurized, thereby precluding any further cladding deformation. Therefore, a coolable geometry was demonstrated.

Criterion 5: Long-Term Cooling: After any calculated successful initial operation of the ECOS, the calculated core temperature shall be maintained at an acceptably low value and decay heat shall be removed for the extended period of time required by the long-lived radioactivity remaining In the core.

Result: The LBLOCA and SBLOCA ECCS performance analyses demonstrated that the Watorford 3 ECCS successfully maintains the fuel cladding tomporaturo at an acceptably low value in the short term. Subsequently, for the extended period of time required by the long-lived radioactivity remaining in the core, the ECCS continues to supply sufficient cooling waler from the refueling water storage pool and then from the sump to remove decay heat and maintain the core temperature at an acceptably low value. In addition, at the appropriate time, the operator realigns a HPSI pump for simultaneous hot and cold leg injection in order to maintain the core bore acid concentration below the solubility limit.

2.12-83 63O5Aklc-1

=36.dac-1 1105103 1105103 2.12-83

Waterford 3 Extended Power Uprate rapid flow reduction causes the pressure difference (AP) across the affected loop steam generator (SG) to drop below the low flow differential pressure trip setpoint.

The reactor trip produces an automatic turbine trip. Following the turbine trip, there is an assumed failure to transfer to offsite power, thus a loss-of-non-emergency-offsfte power (LOOP) is experienced. This results in a loss of power (LOOP) to the remaining RCPs, steam bypass valves (SBVs), the PPCS and the PLCS. The main impact is that the remaining 3 RCPs begin coastdown, resulting in further loss of flow at a time in which the core is already experiencing possible DNB. This makes the immediate DNB situation more adverse.

This evaluation conservatively analyzed LOOP occurring coincident with turbine trip. DNBR degradation is terminated when the mitigating effects of scram CEA Insertion dominate the flow coastdown resulting from the combined effects of the initiating seized rotorlsheared shaft event and the subsequent LOOP.

Table 2.13.3.3.1-1 contains the initial conditions and assumptions used for RCP seized rotor/sheared shaft with LOOP event.

2.13.3.3.1.5 Radiological Consequences The radiological consequences for the RCP seized rotorlsheared shaft were calculated assuming that the radioisotopes in the gas gap of the pins that experience DNB was immediately mixed with the RCS for release. Releases for site boundary doses were calculated accounting for the carry over of activity to tho secondary Soytem via SG tube Ickago paths.

The allowed fuel failure limit for the RCP seized rotor/sheared shaft was back-calculated to determine the extent of fuel failure which would result in the regulatory limit for the event.

Cycle-specific fuel failure evaluations for power uprate cores will be performed to ensure that this fuel failure limit will not be exceeded. For the RCP seized rotor/sheared shaft, the fuel failure limit to meet the SRP Guidance of a small fraction of 10CFR100 was determined to be 8.0% of the fuel pins.

The radiological consequences resulting from these fuel failure results are:

I 2-Hour EAB I 8-Hour LPZ l Thyroid < 30 rem <: 30 rem Aep;s &,jro1 3 -se35 Whole Body <gem it<rem M4ety 7, ZtOq 2.13.3.3.1.6 Results i I iiŽ The combined impact of a more adverse flow coastdown and a later time to reach the credited RPS trip condition results in the RCP sheared shaft event being more limiting than the RCP seized rotor event. LOOP occumng coincident with turbine trip was determined to be more limiting that an LOOP occurring shortly after turbine trip. Hence, the most limiting event for this analysis (i.e., resulting in more adverse DNBR degradation) is the RCP sheared shaft event with LOOP at time of trip. A sequence of events for this limiting case is shown in Table 2.13.3.3.1-2.

2.13-256 6306 6306-2.doc-1 1/05/03 2.doc-1 1/05/03 2.13-256 W3FI-2003-0074 Page 2 of 5 PUR Commitment Type Scheduled Section (Check One) Completion One-Time Continuing Date (If Action Compliance Required) 2.3.2.2 Due to the EPU, the existing breakers' X End of continuous load and short circuit RF13 interrupting ratings will be exceeded, as

. he4l as he associated mechanical T

disconnect switches on either side of each breaker. New breakers and disconnects are required to meet 1333 MVA and wiobe powerland oter RFi1 rements o e 2.4.1.1 hver,because te stEP rsteatN X Endof generator and main steam operating RF13 pressures, the diverse emergency fpp"MArnr leuation signal (DEFAS) andcyle sefpoint permissive will be lowered; peif analysi. _____R___________

2.4.1.2 The EPU configuration has been analyzed X End of to a lower steam generator pressure. These RF13 PPS chantes will be Imnlemented for EPU.

2.4.1f2 CPC constants will be updated to be X End of consistent with the new definition of 100% RF13 power and other requirements of the EPU and cycle specific analpsis.

2.4.2.1 The EPU requires adjustments to NSSS X power control systems setpoints and paramotersAcoin to provide proper control system during Cycle performnance for the EPU operating 14

_conditions.

2.4.2.1 The COLSS constants that are based on X End of the reactor thermal power tng RF13 instrumentation uncertainties will be modified as necessary as pa rt of the reload fuel design process. These constants will be calculated and implemented as part of

_the reload fuel design process._

2.5.1.2 The EPU will employ the same measures End of that have been taken for existing valves to RF13 IDrevent missile aeneration.

Attachment 8 W3Fl -2003-0074 Page 3 of 5 PUR Commitment Type Scheduled Section (Check One) Completion One-Time Continuing Date (If Action Compliance Required) 2.5.2.4 Based on this rule change to 10 CFR 50.44, X 12/31/03 Entergy will be submitting a separate license amendment requoct to climinato the Waterford 3 technical specification requirements for combustible gas control in containment. This license amendment request will be submitted by the end of 2003.

2.5.3.1 In response to Generic Letter 2003-01, X 9/30/04 Control Room Habitability, Entergy has committed to complete the requested evaluation prior to the end of September 2004. This evaluation will include a validation of the Inleakage assumptions made in the dose consequence analyses.

The results of this evaluation will determine further appropriate actions, if any, that must be taken to resolve this Issue. (Reference commitment A26565) 2.5.5.3 These higher heat loads will increase the X End ot temperature of the CCWS return flow in RF13 some of the CCWS piping sections. The impact of those higher tomporaturos on the CCW piping, supports and components will be evaluated. Aso, the impact of these higher temperatures on the shutdown cooling heat exchanger room cooler will be evaluated.

2.5.6.1 In order to accept the higher MSSS flows X End of for EPU, the HP turbine steam path will be RF1 3 replaced.

2.5.6.1 The feedwater control system setpoints will X End of be modified slightly to increase pump speed r1 at a lower demand.

Th SR safe alves We.5,1 un Iersi _

C 2.5.6.2 Measures will be implemented as End of necessary to prevent potential condenser RF13 tube vibration under power uprate conditions.

W3FI-2003-0074 Page 4 of 5 PUR Commitment Type Scheduled Section (Check One) Completion One-Time Continuing Date (If Action Compliance Required) 2.5.8.1 Entergy will perform additional testing to X End of reconfirm the acceptability of the fuel oil RF13 consumption rates utilized in the fuel oil usage camculation prior to implementing lhe EPU.

2.9.1 Sptpoints that amr mpntioned in emergpncy X End of operating procedures or off normal RF13 procedures will be changed to new uprate values.

2.9.1 The time window to establish simultaneous X End of hot and cold leg injection in the LOCA RF13 emergency operating procedure will be chagried from 2 and 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to 2 and 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />.

2.9.1 As discussed In Section 2.5.2.4, Entergy is X 'Ff >a ays submitting a separate license amendment Jo l request to eliminate the Technical Iot/ O ta- iss nce Specification requirements for combustible mb ble gas control In containment. Resulting EOP )a r t g contro changes, regarding operation of the combustible gas control system, will be ( mn nt made during the implementation phase following approval and issuance of the associated amendment.

2.9.4 Color banding will change for the following X End of indicated parameters: RF13 a) Pressurizer pressure narrow range b) Pressurizer pressure wide range c) RCS cold leg temperature d) RCS hot leg temperature e) SIT levels 2.9.5.1 Prior to EPU implementation, the simulator X Beginning of will be upgraded to provide operator last operator training at EPU conditions. training cycle prior

._ _to RF13 W3Fl-2003-0074 Page 5 of 5 PUR Commitment Type Scheduled Section (CheckOne) Completion One-Time Continuing Date (If Action Compliance Required) 2.9.5.1 The training staff will provide training as X End of determined by the Operations Training RF13 Review Group. Operators will receive training using the systematic approach to training process. This training will cover procedure changes, new or revised technical specifications / safety analysis, and equipment modifications for the EPU.

The topics will include items such as high pressure turbine u de, t nospheric oi4- aedjq4s1eht Adump valve an instrument range I alarm change lassroom training /* -

testing and simulator trai esting with e .. of " '}e ISy the new equipmentinstrume es f A installed will ensure the operators understand the changes to plant systems.

Startup training will be conducted prior to the conclusion of the outage for the operations crews with emphasis on core reload, positive moderator temperature coefficient, reactor engineering interface, and teamwork skills.

2.9.5.3 The EPU simulator modification, design, X Beginning of analysis, and test data will be evaluated last operator based on, but is not limfted to, the following: training a) Engineering report modification cycle prior packages to RF3 b) Design and analysis data including heat balance, balance-of-plant (BOP) flow changes, and plant systems set point changes c) Updated plant and physics data book d) Engineering reports on system response and new accident analyses e) Best-estimate data 2.9.5.4 The simulator will be compared to actual X During plant data. Cycle 14 2.10 A Cycle 14 power ascension test program X Startup after will be performed as described in Section RF13 2.10.

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