BVY 05-105, Vermont Yankee Nuclear Power Station Correction to Technical Specifications Pages

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Vermont Yankee Nuclear Power Station Correction to Technical Specifications Pages
ML053530181
Person / Time
Site: Vermont Yankee Entergy icon.png
Issue date: 12/12/2005
From: Devincentis J
Entergy Nuclear Northeast, Entergy Nuclear Operations
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
%dam200601, BVY 05-105, TAC MB9091, TAC MC4662, TAC MC5243
Download: ML053530181 (16)
v • d • e


Text

Entergy Nuclear Northeast Entergy Nuclear Operations, Inc.

Vermont Yankee P.O. Box 0500 1185 Old Ferry Road vm%0Fr~ ntf~gyBrattleboro, 01 VT 05302-0500 Tel 802 257 5271 December 12, 2005 Docket No. 50-271 BVY 05-105 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

Subject:

Vermont Yankee Nuclear Power Station License No. DPR-28 Correction to Technical Specifications Pages

References:

1) Letter, USNRC to Entergy, "Vermont Yankee Nuclear Power Station -

Issuance of Amendment Re: Intermediate Range Monitor Surveillance Test Frequencies (TAC No. MB9091)," NVY 05-082, July 7, 2005.

2) Letter, USNRC to Entergy, "Vermont Yankee Nuclear Power Station -

Issuance of Amendment Re: Administrative Changes (TAC No.

MC5243)," NVY 05-099, August 15, 2005.

3) Letter, USNRC to Entergy, "Vermont Yankee Nuclear Power Station -

Issuance of Amendment Re: One-Time Extension of Integrated Leak Rate Test Interval (TAC No. MC4662)," NW 05-108, August 31, 2005.

Entergy Nuclear Operations, Inc. (Entergy) hereby provides corrected Technical Specifications (TS) pages for the Vermont Yankee Nuclear Power Station (VY).

The Nuclear Regulatory Commission (NRC) recently issued License Amendments 225, 226 and 227 (References 1, 2 and 3 respectively) to VY's Facility Operating License which inadvertently did not incorporate prior NRC approved revisions to some of the TS pages.

As discussed with the NRR Project Manager, the inadvertent omission of these changes In the respective License Amendment issued pages was unintentional. These errors were entered into VY's corrective action program. A review of the other changes made by these amendments revealed no other discrepancies.

It is requested that NRC issue the four corrected TS pages, as provided in this letter. to this letter provides a description of the changes. Attachment 2 provides the marked-up version of the current Technical Specification Bases pages. Attachment 3 provides the retyped pages.

A0D

BVY 05-105 Docket 50-271 Page 2 of 2 There are no new commitments being made in this submittal.

If you have any questions or require additional information, please contact me at (802) 258-4236.

Sincerely, Jame M. eVincentis Manager, Licensing Vermont Yankee Nuclear Power Station Attachments (3) cc: Mr. Samuel J. Collins Regional Administrator, Region 1 U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406-1415 Mr. James J. Shea, Project Manager Division of Licensing Project Management Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Mail Stop 08 B1 Washington, DC 20555-0001 USNRC Resident Inspector Vermont Yankee Nuclear Power Station 320 Governor Hunt Road Vernon, VT 05354 Mr. David O'Brien, Commissioner Vermont Department of Public Service 112 State Street, Drawer 20 Montpelier, VT 05620-2601

ATTACHMENT 1 TO BVY 05-105 Vermont Yankee Nuclear Power Station Correction to Technical Specifications Pages DESCRIPTION OF CHANGES ENTERGY NUCLEAR OPERATIONS, INC.

VERMONT YANKEE NUCLEAR POWER STATION DOCKET NO. 50-271

BVY 05-105, Attachment 1 Docket 50-271 Page 1 of 3 Description of Changes The Nuclear Regulatory Commission (NRC) recently issued License Amendments 225, 226 and 227 (References 1, 2 and 3 respectively) to the Technical Specifications (TS) for the Vermont Yankee Nuclear Power Station (VY). The replacement TS pages inadvertently did not incorporate prior NRC approved revisions to some of the TS pages. As discussed with the NRR Project Manager, the inadvertent omission of these changes in the respective License Amendment issued pages was unintentional. The following description of changes provides information about each of the requested corrections.

1. Correction to TS page 27:

License Amendment 225 (Reference 1) inadvertently did not incorporate a change to TS page 27 that was previously approved by License Amendment 219 (Reference 4). In TS page 27, Table 4.1.2, the text 'Every 3 Months (9)," was unintentionally omitted from the Minimum Frequency column for the High Flux Average Power Range Monitor (APRM) Flow Bias. A corrected retyped TS page 27 is provided in Attachment 3.

Justification: In Entergy's letter to the NRC dated May 21, 2003 (Reference 5), Entergy proposed to revise the functional test frequency for the Intermediate Range Monitors.

Entergy's original submittal, which included a retyped version of TS page 27, was submitted prior to NRC's issuance of License Amendment 219, dated April 14, 2004, and therefore did not incorporate changes made by License Amendment 219. Just prior to NRC's issuance of License Amendment 225, Entergy provided final TS pages that inadvertently did not reflect changes to TS page 27 made by License Amendment 219.

This incorrect version of TS page 27 was then issued in License Amendment 225 and implemented in the VY TS.

2. Correction to TS page 120:

License Amendment 226 (Reference 2) inadvertently did not incorporate a change to TS Section 3.6.D.1 (page 120) that was previously approved by License Amendment 219 (Reference 4). The approved change of text from "both safety valves" to "all safety valves" was unintentionally overlooked on the retyped page 120 provided by License Amendment 226. A corrected retyped TS page 120 is provided in Attachment 3.

Justification: The retyped TS page 120 provided in Entergy's letter to the NRC dated December 6, 2004 (Reference 6) showed the incorrect phrase "both safety valves" rather than the correct phrase "all safety valves." This incorrect version of TS page 120 was then issued in License Amendment 226 and implemented in the VY TS.

3. Correction to TS page 155a:

License Amendment 226 (Reference 2) inadvertently did not incorporate a change to TS Section 4.7.C.1 (page 155a) that was previously approved by License Amendment 223 (Reference 7). The value "1,550 cfm" was inadvertently changed back to "1,500 cfm" on

BVY 05-105, Attachment 1 Docket 50-271 Page 2 of 3 the retyped page 155a provided by License Amendment 226. A corrected retyped TS page 155a is provided in Attachment 3.

Justification: The retyped TS page 155a provided in Entergy's letter to the NRC dated December 6, 2004 (Reference 6) was submitted prior to NRC's issuance of License Amendment 223, dated March 29, 2005 (Reference 7), and therefore did not incorporate changes made by License Amendment 223. Just prior to NRC's issuance of License Amendment 226, Entergy provided final TS pages that inadvertently did not correctly reflect changes to TS page 155a made by License Amendment 223. This incorrect version of TS page 155a was then issued in License Amendment 226 and implemented in the VY TS.

4. Correction to TS page 265:

The retyped page 265 provided by License Amendment 227 (Reference 3) unintentionally omitted a sentence, "The maximum allowable primary containment leakage rate, La, at Pa, shall be 0.8% of primary containment air weight per day.' A corrected retyped TS page 155a is provided in Attachment 3.

Justification: Just prior to NRC's issuance of License Amendment 227, Entergy provided final TS pages to reflect changes made by Amendment 223 (Reference 7). Amendment 223 changes were correctly reflected; however, the subject sentence was inadvertently deleted. This incorrect version of TS page 265 was then issued in License Amendment 227 and implemented in the VY TS.

BVY 05-105, Attachment 1 Docket 50-271 Page 3 of 3

References:

1. Letter, USNRC to Entergy, "Vermont Yankee Nuclear Power Station - Issuance of Amendment Re: Intermediate Range Monitor Surveillance Test Frequencies (TAC No.

MB9091)," Amendment No. 225, NVY 05-082, July 7, 2005.

2. Letter, USNRC to Entergy, "Vermont Yankee Nuclear Power Station - Issuance of Amendment Re: Administrative Changes (TAC No. MC5243)," Amendment No. 226, NW 05-099, August 15, 2005.
3. Letter, USNRC to Entergy, "Vermont Yankee Nuclear Power Station - Issuance of Amendment Re: One-Time Extension of Integrated Leak Rate Test Interval (TAC No.

MC4662)," Amendment No. 227, NW 05-108, August 31, 2005.

4. Letter, USNRC to Entergy, 'Vermont Yankee Nuclear Power Station - Issuance of Amendment Re: Implementation of ARTS/MELLLA (TAC No. MB8070)," Amendment No. 219, NW 04-031, April 14, 2004.
5. Letter, Entergy to USNRC, "Vermont Yankee Nuclear Power Station, License No. DPR-28 (Docket No. 50-271), Technical Specification Proposed Change No. 260, Intermediate Range Monitor Surveillance Test Frequencies," BW 03-49, May 21, 2003.
6. Letter, Entergy to USNRC, "Vermont Yankee Nuclear Power Station, Technical Specification Proposed Change No. 270, Administrative Changes," BW 04-118, December 6, 2004.
7. Letter, USNRC to Entergy, "Vermont Yankee Nuclear Power Station - Issuance of Amendment Re: Alternative Source Term (TAC No. MC0253)," Amendment No. 223, NW 05-045, March 29, 2005.

ATTACHMENT 2 TO BVY 05-105 Vermont Yankee Nuclear Power Station Correction to Technical Specification Pages MARKED-UP PAGES ENTERGY NUCLEAR OPERATIONS, INC.

VERMONT YANKEE NUCLEAR POWER STATION DOCKET NO. 50-271

VYNPS TABLE 4.1.2 SCRAM INSTRUMENT CALIBRATION MINIMUM CALIBRATION FREQUENCIES FOR REACTOR PROTECTION INSTRUMENT!CHANNELS Instrument Channel Group (1) Calibration Standard(4) Minimum Frequency(2 )

High Flux IRM C Standard Voltage Source Once/Operating Cycle Output Signal(7)(10)(11)

High Flux APRM Output Signal B Heat Balance Once Every 7 Days Output Signal (Reduced)

Flow Bias (7) B B

Heat Balance Standard Pressure and Voltage Source R'flnr; cj'e,3 /'.

n Once Every 7 Days M 4 t

-e___

(a) )

(LPRM ND-2-1-104(80)) B(5 ) Using TIP System Every 2,000 MWD/T average core exposure (8)

High Reactor Pressure B Standard Pressure Source On6e/Operating Cycle Turbine Control Valve Fast Closure A Standard Pressure Source Every 3 Months High Drywell Pressure B Standard Pressure Source Once/Operating Cycle High Water Level in Scram Discharge B Water Level Once/Operating Cycle Volume Low Reactor Water Level B Standard Pressure Source Once/Operating Cycle Turbine Stop Valve Closure A (6) Refueling Outage First Stage Turbine Pressure A Pressure Source Every 6 Months and After Permissive (PS-5-14(A-D)) Refueling Main Steam Line Isolation Valve A (6) Refueling Outage Closure Amendment No. -14, GE, 24, i@, i4, -7, 46-4, 418.6, 4, 27

VYNPS 3.6 LIMITING CONDITIONS FOR 4.6 SURVEILLANCE REQUIREMENTS OPERATION D. Safety and Relief Valves D. Safety and Relief Valves

1. During reactor power 1. Operability testing of operating conditions and Safety and Relief Valves whenever the reactor shall be in accordance coolant pressure is with Specification 4.6.E.

greater than 150 psig and The lift point of the a - ------

emperature greater than safety and relief valves 3500 F, ~ safety valves shall be set as specified and at--Teast--three--of -the in Specification 2.2.B.

four relief valves shall be operable.

2. If Specification 3.6.D.1 is not met, initiate an orderly shutdown and the reactor coolant pressure shall be below 150 psig and 350°F within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

E. Structural Integrity and E. Structural Integrity and Operability Testing Operability Testing The structural integrity and 1. Inservice insnention of the operability of the safety-related components safety-related systems and shall be performed in components shall be accordance with maintained at the level Section XI of the ASME required by the original Boiler and Pressure acceptance standards Vessel Code and throughout the life of the applicable. Addenda as plant. required by 10 CFR 50, Section 50.55a(g), except where specific written relief has been granted by the NRC.

Inservice inspection of piping, identified in NRC Generic Letter 88-01, shall be performed in accordance with the staff positions on schedule, methods, and personnel and sample expansion included in the Generic Letter or in accordance with alternate measures approved by NRC Staff.

Amendment No. 1-3, 48, 445, 4-, 4-2-s, 4-39,44G, 172, 1-77, 46, 2-1-9

  • 120 120

VYNPS 3.7 LIMITING CONDITIONS FOR 4.7 SURVEILLANCE REQUIREMENTS OPERATION

i. Suspend movement of irradiated fuel assemblies and the fuel cask in secondary containment; and

- ~ii. Suspend core alterations; and iii. Initiate action to suspend operations with the potential for draining the reactor vessel.

C. Secondary Containment System C. Secondary Containment System

1. Secondary Containment 1. Secondary containment Integrity shall be capability to maintain a maintained during the 0.15 inch of water vacuum following modes or under calm wind conditions: (2<0<5 mph) conditions with a filter train flow
a. Whenever the reactor e of not more than is in the Run Mode, cfm, shall be Startup Mode, or Hot emonstrated at least Shutdown condition*; /quarterly.

or

  • NOTE: The reactor mode switch may be changed to either the Run or Startup/Hot Standby position, and operation not considered to be in the Run Mode or Startup Mode, to allow testing of instrumentation associated with the reactor mode switch interlock functions, provided:
1. Reactor coolant temperature is < 212'F;
2. All control rods remain fully inserted in core cells containing one or more fuel assemblies; and
3. No core alterations are in progress.

Amendment No. 44-4, 7, 4-97, , -1 23 55a 155a

VYNPS Report for-the period of the report in which any change to the ODCM was made. Each change shall be identified by markings in the margin of the affected pages, clearly

-indicating the area of the page that was changed, and shall indicate the date (e.g., month/year) the change was implemented.

C. PRIMARY CONTAINMENT LEAKAGE RATE TESTING PROGRAM A program shall be established to implement the leakage rate testing of the primary containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, entitled "Pertoxmance..Based Containment Leak-Test Program," dated September 1995, as modified by the following:

  • The first Type A test after the April 1995 Type A test shall be performed no later than April 2010. (This is an exception to Section 9.2.3 of NEI 94-01, Rev. 0, "Industry Guideline for Implementing Performance-Based Option of 10CFR50, Appendix J.")

Option B; (2) Section 6.4.4 of ANSI/ANS 56.8-1994; and (3)

Section 10.2 of NEI 94-01, Rev. 0.

Option B; (2) Section 3.2 of ANSI/ANS 56.8-1994; and (3)

Sections 8.0 and 9.0 of NEI 94-01, Rev. 0.

The peak calculated containment internal pressure for the design basis loss of coolant accident, Pa, is 44 psig.

Leakage rate acceptance criteria are:

1. Primary containment leakage rate acceptance criterion
  • 1.0 La.
2. The as-left primary containment integrated leakage rate test (Type A test) acceptance criterion is
  • 0.75 La.
3. The combined local leakage rate test acceptance criterion for.

Type B and Type C tests (excluding the leakage contributions from the main steam pathways) is 5 0.6 La, calculated on a maximum pathway basis, prior to entering a mode of operation where primary containment integrity is required.

4. The combined local leakage rate test acceptance criterion for Type B and Type C tests (excluding the leakage contributions from the main steam pathways) is < 0.6 La, calculated on a minimum pathway basis, at all times when primary containment integrity is

\ required.

C Se m~x \o,,<;lc pr;>X¢ CdvAr, VA tN e~lyn-G c- fae Lo A ,5v b eo, .,f% c q.\^ >> r Amendment No. 151, 152, 171, 215, 2Q4, -22r2 265

Attachment 3 to BVY 05-105 Docket 50-271 Listing of Affected Technical Specifications Pages Replace the Vermont Yankee Nuclear Power Station Technical Specifications page listed below with the revised page. The revised page contains a vertical line in the margin indicating the area of change.

Remove Insert 27 27 120 120 155a 155a 265 265

TABLE 4.1.2 SCRAM INSTRUMENT CALIBRATION MINIMUM CALIBRATION FREQUENCIES FOR REACTOR PROTECTION INSTRUMENT CHANNELS Instrument Channel Group (1) Calibration Standard(4) Minimum Frequency(2)

High Flux IRM Output Signal(7)(10)(11) C Standard Voltage Source Once/Operating Cycle High Flux APRM Output Signal B Heat Balance Output Signal (Reduced) (7) Once Every 7 Days B Heat Balance Once Every 7 Days Flow Bias B Standard Pressure and Voltage . Refueling Outage l Source Every 3 Months (9)

LPRM (LPRM ND-2-l-104(80))

B(5 ) Using TIP System Every 2,000 MWD/T average core exposure (8)

High Reactor Pressure B Standard Pressure Source Once/Operating Cycle Turbine Control Valve Fast Closure A Standard Pressure Source Every 3 Months High Drywell Pressure B Standard Pressure Source Once/Operating Cycle High Water Level in Scram Discharge B Water Level Volume Once/Operating Cycle Low Reactor Water Level B Standard Pressure Source Once/Operating Cycle Turbine Stop Valve Closure A (6) Refueling Outage First Stage Turbine Pressure A Pressure Source Every 6 Months and After Permissive (PS-5-14(A-D))

Refueling Main Steam Line Isolation Valve A (6)

Closure Refueling Outage Amendment No. 14, Go., 2, "8, .4-, 46, 444, m,19.1, A1, n9 222 27

VYNPS 3.6 LIMITING CONDITIONS FOR 4.6 SURVEILLANCE REQUIREMENTS OPERATION D. Safety and Relief Valves D. Safety and Relief Valves

1. During reactor power 1. Operability testing of operating conditions and Safety and Relief Valves whenever the reactor shall be in accordance coolant pressure is with Specification 4.6.E.

.greater than 150 psig and The lift point of the temperature greater than safety and relief valves I 350 0F, all safety valves shall be set as specified and at least three of the in Specification 2.2.B.

four relief valves shall be operable.

2. If Specification 3.6.D.1 is not met, initiate an orderly shutdown and the reactor coolant pressure shall be below 150 psig and 350'F within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

E. Structural Integrity and E. Structural Integrity and Operability Testing Operability Testing The structural integrity and 1. Inservice inspection of the operability of the safety-related components safety-related systems and shall be performed in components shall be accordance with maintained at the level Section XI of the ASME required by the original Boiler and Pressure acceptance standards Vessel Code and throughout the life of the applicable Addenda as plant. required by 10 CFR 50, Section 50.55a(g), except where specific written relief has been granted by the NRC.

Inservice inspection of piping, identified in NRC Generic Letter 88-01, shall be performed in accordance with the staff positions on schedule, methods, and personnel and sample expansion included in the Generic Letter or in accordance with alternate measures approved by NRC Staff Amendment No. A4s, 4A,-4-s, 4 2, 3 4.4, 14, 4-74, 177 4", 2-4, 6-2-0 120

VYNPS 3.7 LIMITING CONDITIONS FOR 4.7 SURVEILLANCE REQUIREMENTS OPERATION

i. Suspend movement of irradiated fuel assemblies and the fuel cask in secondary containment; and ii. Suspend core alterations; and iii. Initiate action to suspend operations with the potential for draining the reactor vessel.

C. Secondary Containment System C. Secondary Containment System

1. Secondary Containment 1. Secondary containment Integrity shall be capability to maintain a maintained during the 0.15 inch of water vacuum following modes or under calm wind conditions: (2<Z<5 mph) conditions with a filter train flow
a. Whenever the reactor rate of not more than is in the Run Mode, 1,550 cfm, shall be I Startup Mode, or Hot demonstrated at least Shutdown condition*; quarterly.

or

  • NOTE: The reactor mode switch may be changed to either the Run or Startup/Hot Standby position, and operation not considered to be in the Run Mode or Startup Mode, to allow testing of instrumentation associated with the reactor mode switch interlock functions, provided:
1. Reactor coolant temperature is < 212'F;
2. All control rods remain fully inserted in core cells containing one or more fuel assemblies; and
3. No core alterations are in progress.

Amendment No. 444, 4-4-, 4-94, Q-24, 2-2-6 155a

VYNPS Report for the period of the report in which any change to the ODCM was made. Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page that was changed, and shall indicate the date (e.g., month/year) the change was implemented.

C. PRIMARY CONTAINMENT LEAKAGE RATE TESTING PROGRAM A program shall be established to implement the leakage rate testing of the primary containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, entitled

'Performance Based Containment Leak-Test Program," dated September 1995, as modified by-the following:

  • The first Type A test after the April 1995 Type A test shall be performed no later than April 2010. (This is an exception to Section 9.2.3 of NEI 94-01, Rev. 0, "Industry Guideline for Implementing Performance-Based Option of 10CFR50, Appendix J.")

Option B; (2) Section 6.4.4 of ANSI/ANS 56.8-1994; and (3)

Section 10.2 of NEI 94-01, Rev. 0.

Option B; (2) Section 3.2 of ANSI/ANS 56.8-1994; and (3)

Sections 8.0 and 9.0 of NEI 94-01, Rev. 0.

The peak calculated containment internal pressure for the design basis loss of coolant accident, Pa, is 44 psig.

The maximum allowable primary containment leakage rate, La, at Pa, shall be 0.8% of primary containment air weight per day.

Leakage rate acceptance criteria are:

1. Primary containment leakage rate acceptance criterion
  • 1.0 La.
2. The as-left primary containment integrated leakage rate test (Type A test) acceptance criterion is
  • 0.75 La.
3. The combined local leakage rate test acceptance criterion for Type B and Type C tests (excluding the leakage contributions from the main steam pathways) is
  • 0.6 La, calculated on a maximum pathway basis, prior to entering a mode of operation where primary containment integrity is required.
4. The combined local leakage rate test acceptance criterion for Type B and Type C tests (excluding the leakage contributions from the main steam pathways) is
  • 0.6 La, calculated on a minimum pathway basis, at all times when primary containment integrity is required.

Amendment No. 414 ,r 62 44 715 24 r4 23, 2426 265

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