ML130090215

Issuance of Amendment Change to Suppression Chamber Drywell Leak Rate Test Surveillance Frequency

ML130090215
Person / Time
Site:	Vermont Yankee
Issue date:	01/30/2013
From:	Richard Guzman Plant Licensing Branch 1
To:	Entergy Nuclear Operations, Entergy Nuclear Vermont Yankee
Guzman R, NRR/DORL 415-1030
References
TAC ME7928
Download: ML130090215 (14)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 January 30, 2013 Site Vice President Entergy Nuclear Operations, Inc.

Vermont Yankee Nuclear Power Station P.O. Box 250 Governor Hunt Road Vernon, VT 05354

SUBJECT:

VERMONT YANKEE NUCLEAR POWER STATION -ISSUANCE OF AMENDMENT TO RENEWED FACILITY OPERATING LICENSE RE: CHANGE TO SUPPRESSION CHAMBER-DRYWELL LEAK RATE TEST SURVEILLANCE FREQUENCY (TAC NO. ME7928)

Dear Sir or Madam:

The Nuclear Regulatory Commission (NRC) has issued the enclosed Amendment No. 254 to Renewed Facility Operating License No. DPR-28 for the Vermont Yankee Nuclear Power Station (VYNPS), in response to your application dated February 1, 2012, as supplemented on August 7 and November 20, 2012.

This amendment revises the VYNPS Technical Specification 4.7.A.6.b.3 for performing the drywell-to-suppression chamber leak rate test during an operating cycle instead of during a refueling outage.

A copy of the related Safety Evaluation is enclosed. A Notice of Issuance will be included in the Commission's biweekly Federal Register notice.

Sincerely, Richard V. Guzman, Senior Project Manager Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-271

Enclosures:

1. Amendment No. 254 to License No. DPR-28

2. Safety Evaluation cc w/encls: Distribution via Listserv

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555*0001 ENTERGY NUCLEAR VERMONT YANKEE, LLC AND ENTERGY NUCLEAR OPERATIONS, INC.

DOCKET NO. 50-271 VERMONT YANKEE NUCLEAR POWER STATION AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 254 Renewed License No. DPR-28

1. The Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment filed by Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations, Inc. (the licensee) dated February 1,2012, as supplemented by letters dated August 7 and November 20, 2012, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance: (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

Enclosure 1

-2

2. Accordingly, the license is amended as indicated in the attachment to this license amendment, and paragraph 3.B of the Renewed Facility Operating License No. DPR-28 is hereby amended to read as follows:

(B) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 254, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3. This license amendment is effective as of its date of issuance and shall be implemented within 60 days.

FOR THE NUCLEAR REGULATORY COMMISSION

~cJ..~

George A. Wilson, Chief Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the License and Technical Specifications Date of Issuance: January 30, 2013

ATTACHMENT TO LICENSE AMENDMENT NO. 254 RENEWED FACILITY OPERATING LICENSE NO. DPR-28 DOCKET NO. 50-271 Replace the following page of the Renewed Facility Operating License with the attached revised page. The revised page is identified by amendment number and contains marginal lines indicating the areas of change.

Remove Page Insert Page 3 3 Replace the following pages of Appendix A, Technical Specifications, with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Pages Insert Pages 147 147 149 149 150 150

- 3 D. Entergy Nuclear Operations, Inc., pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use in amounts as required any Byproduct, source, or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components.

E. Entergy Nuclear Operations, Inc., pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not to separate, such byproduct and special nuclear material as may be produced by operation of the facility.

3. This renewed license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations: 10 CFR Part 20, Section 30.34 of 10 CFR Part 30, Section 40.41 of 10 CFR Part 40, Section 50.54 and 50.59 of 10 CFR Part 50, and Section 70.32 of 10 CFR Part 70; 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 is subject to the additional conditions specified below:

A. Maximum Power Level Entergy Nuclear Operations, Inc. is authorized to operate the facility at reactor core power levels not to exceed 1912 megawatts thermal in accordance with the Technical Specifications (Appendix A) appended hereto.

B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 254 are hereby incorporated in the license. Entergy Nuclear Operations, Inc. shall operate the facility in accordance with the Technical Specifications.

C. Reports Entergy Nuclear Operations, Inc. shall make reports in accordance with the requirements of the Technical Specifications.

D. This paragraph deleted by Amendment No. 226.

E. Environmental Conditions Pursuant to the Initial Decision of the presiding Atomic Safety and Licensing Board issued February 27, 1973, the following conditions for the protection of the environment are incorporated herein:

1. This paragraph deleted by Amendment No. 206, October 22, 2001.

2. This paragraph deleted by Amendment 131, 10/07/91.

Renewed Facility Operating License No. DPR-28 Amendment No.~, 254

VYNPS 3.7 LIMITING CONDITIONS FOR 4.7 SURVEILLANCE REQUIREMENTS OPERATION at normal cooldown rates if the torus water temperature exceeds 120°F.

e. Minimum Water Volume

- 68,000 cubic feet

f. Maximum Water Volume

- 70,000 cubic feet

2. Primary containment 2. Primary Containment integrity shall be Surveillances maintained at all times when the reactor is a. The primary containment critical or when the integrity shall be reactor water temperature demonstrated as required is above 212°F and fuel is by the Primary in the reactor vessel Containment Leakage Rate except while performing Testing Program (PCLRTP).

low power physics tests at atmospheric pressure at b. Once every 18 months, a power levels not to exceed drywell to suppression 5 Mw(t). chamber leak rate test shall demonstrate that

3. If a portion of a system with an initial that is considered to be differential pressure of an extension of primary not less than 1.0 psi, containment is to be the differential pressure opened, isolate the decay rate shall not affected penetration flow exceed the equivalent of path by use of at least the leakage rate through one closed and deactivated a 1-inch orifice. Should automatic valve, closed there be two consecutive manual valve or blind test failures the test flange. frequency shall be changed to once every 9

4. Whenever primary months until two containment integrity is consecutive tests pass.

required:

3. (Blank)

a. The leakage rate from anyone main steam 4. In accordance with the isolation valve (MSIV) PCLRTP, verify that the shall not exceed 62 following leakage rates are scfh at 44 pSig (Pa); within acceptable limits:

b. The combined leakage a. The leakage rate through rate from the main each MSIV; steam pathways shall not exceed 124 scfh at b. The combined leakage rate 44 psig (Pa); and for the main steam pathways; and

c. The combined leakage rate from the secondary c. The combined leakage rate containment bypass for the secondary pathways shall not containment bypass exceed 5 scfh at 44 pathways.

psig (Pa) *.

Amendment No. §G, ~, i-Q, 4+&, ~ 254 147

VYNPS I

3.7 LIMITING CONDITIONS FOR 4.7 SURVEILLANCE REQUIREMENTS OPERATION line is verified to be closed and conditions required by 3.7.0.2 are met.

6. Pressure Suppression 6. Pressure Suppression Chamber - Drywell Vacuum Chamber - Drywell Vacuum Breakers Breakers

a. When primary a. Periodic Operability containment is Tests required, all suppression ,:hamber Operability testing

- drywell vacuum* of the vacuum breakers shall be breakers shall be in operable except accordance with during testing and Specification 4.6.E as stated in and within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Specifications after any discharge 3.7.A.6.b and c, of steam to the below. Suppression suppression chamber chamber - drywell from the vacuum breakers safety/relief valves shall be considered and within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> operable if: following an operation that (1) The valve is causes any of the demonstrated vacuum breakers to to open fully open. Operability with the of the corresponding applied force position switches at all valve and position positions not indicators and exceeding alarms shall be that verified monthly and equivalent to following any 0.5 psi maintenance.

acting on the suppression b. Refueling Outage chamber face Test of the valve disk. (1). All suppression (2) The val ve can chamber be closed by drywell gravity, when vacuum released breaker after being position opened by indication remote or and alarm manual means, systems shall to within not be calibrated greater than and the functionally equivalent of tested.

0.05 inch at all points (2) Deleted along the seal surface of the disk.

Amendment No. ~, ~ 254 149

VYNPS 3.7 *LIMITING CONDITIONS FOR 4.7 SURVEILLANCE REQUIREMENTS OPtRATION (3) The position alarm system will annunciate in the control room if the valve opening exceeds the equivalent of 0.05 inch at all points along the seal surface of the disk. (3) Deleted

b. Up to two (2) of the ten (10) suppression chamber - drywell vacuum breakers may be determined to be inoperable provided i. Oxygen Concentration that they are secured, or known to The primary containment be, in the closed oxygen concentration shall position. be measured and recorded on a weekly basis.

c. Reactor operation may continue for fifteen (15) days provided that at least one position alarm circuit for each vacuum breaker is operable and each suppression chamber

- drywell vacuum breaker is physically verified to be closed immediately and daily thereafter.

7. Oxygen Concentration

a. The primary containment atmosphere shall be reduced to less than 4 percent oxygen by volume with nitrogen gas while in the RUN MODE during the time period:

i.From 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after thermal power is greater than 15% rated thermal power following startup, to Amendment No. ~, ~, ~, ~ 254 150

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 254 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-28 ENTERGY NUCLEAR VERMONT YANKEE, LLC AND ENTERGY NUCLEAR OPERATIONS, INC.

VERMONT YANKEE NUCLEAR POWER STATION DOCKET NO. 50-271

1.0 INTRODUCTION

By letter dated February 1, 2012 (Reference 1), as supplemented by letters dated August 7, 2012 (Reference 2), and November 20,2012 (Reference 4), Entergy Nuclear Operations, Inc.

(Entergy, the licensee) submitted a license amendment request (LAR) for changes to the Vermont Yankee Nuclear Power Station (vyNPS) Technical Specifications (TSs).

The supplemental letters dated August 7 and November 20, 2012, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the Nuclear Regulatory Commission (NRC) staff's original proposed no significant hazards consideration determination as published in the Federal Register on April 3, 2012 (77 FR 20074).

The proposed change would revise TS 4.7.A.6.b.3 for performing the drywell-to-suppression chamber leak rate test during an operating cycle (OC) instead of during a refueling outage (RFO). VYNPS is on an 18-month OC that is defined as the interval between the end of an RFO and the next subsequent RFO. This TS change would allow the surveillance specified in TS 4.7.A.6.b.3 to be performed during the OC which is inclusive of the RFO period. Additionally, the revision would renumber and re-title the TSs, accordingly, to support the proposed TS change.

2.0 REGULATORY EVALUATION

The regulatory requirements and guidance which the NRC staff considered in assessing the proposed TS change are as follows:

As stated in Title 10 of the Code of Federal Regulations (10 CFR). Section 50.36(c)(2)(i), the "Limiting conditions for operation are the lowest functional capability or performance levels of equipment required for safe operation of the facility. When a limiting condition for operation of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any remedial action Enclosure 2

-2 permitted by the technical specification ..." Criterion 3 of 10 CFR 50.3S(c)(2)(ii) requires a limiting condition for operation (LCO) to be established for a structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design-basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

General Design Criterion (GDC) 1S, "Containment design," requires that the containment and associated systems be designed to establish an essentially leak tight barrier against the uncontrolled release of radioactivity to the environment, and to assure that the containment design conditions important to safety are not exceeded as long as postulated accident conditions require.

3.0 TECHNICAL EVALUATION

3.1 Licensee's Proposed Change The licensee's specific proposed changes to the TS in Reference 1 are as follows:

• Revise the title of TS 4.7.A.S.b from "Refueling Outage Tests" to "Refueling Outage Test"

• Delete TS 4.7.A.S.b.2

• Renumber TS 4.7.A.S.b.3 to TS 4.7.A.S.c and TS 4.7.A.S.c.1 and add the title "Operating Cycle Test" to TS 4.7.A.S.c to allow the test to be performed during the OC.

3.2 NRC Staff Evaluation Steam blowdown into the drywell from a design-basis accident loss-of-coolant accident can bypass the suppression pool and end up in the suppression chamber airspace through leak paths between the drywell and the suppression chamber. The leak test required in TS 4.7.A.S, "Pressure Suppression Chamber - Drywell Vacuum Breakers," between the drywell and the suppression chamber, measures the drywell-to-suppression chamber differential pressure to ensure that the leakage paths that would bypass the suppression pool are within allowable limits.

The licensee proposes to revise the TSs to perform the drywell-to-suppression chamber leak test on an operating cycle basis so that the test can be performed either during a refueling outage or at another time during the operating cycle (i.e., on-line or during a non-refueling shutdown). The licensee states in its LAR that the proposed TS change: "would allow performance of the test just prior to a refuel outage to confirm the operability of the pressure suppression function of the primary containment and assess the need for maintenance during the refueling outage."

Based on this rationale, the NRC staff understands that the licensee proposes to perform the surveillance requirement (SR) test just prior to (e.g., within a day of) the RFO. In a request for additional information (RAI) dated June 27,2012 (ML 1230SA499), the licensee was requested to provide reasons for performing the test at any time during the OC. In its response (Reference 2), the licensee stated that the intent is to perform the test at anytime during the OC including the RFO. However, the licensee would plan to perform the test just prior to a planned shutdown in order to minimize the risk of a forced shutdown and use the test results to assess the need for

- 3 maintenance during the RFO. Performing this test as a post-maintenance test during the RFO may be necessary if case work is performed to fix the leak paths between the drywell and the suppression chamber.

The licensee referred to the equivalent SR in the Standard Technical Specification (STS)

(Reference 3) for performance of the drywell-to-suppression chamber leak rate test once per 18 months, and stated that the STS does not preclude performing the test at other times during the OC. In the NRC staff's RAI, it is noted that according to STS SR 3.6.1.1.2, two consecutive test failures would indicate unexpected containment degradation, requiring an increased surveillance frequency, until the situation is remedied as evidenced by passing two consecutive tests. Under the condition of failure of two consecutive tests, the STS SR 3.6.1.1.2 requires the test to be performed at an increased frequency of 9 months until two consecutive tests pass. The licensee was also requested in the RAI to provide an explanation for why the increased surveillance frequency requirement of the STS was not being included in the proposed amendment. The licensee stated in its RAI response that it agreed with the NRC staff's observation and proposed to add the increased SR frequency, consistent with the STS, of nine 9 months. Subsequently, the licensee submitted revised TS pages reflecting this change in its supplemental letter (Reference 2).

For the licensee to perform the surveillance test as proposed in TS 4.7.A.6.c in Reference 1, the NRC staff notes that the licensee did not provide a proposed TS action to address the case of a failed surveillance test and the containment being declared inoperable during plant operation. In the NRC staff's RAI, the licensee was requested to specify a TS action that would be required to be taken under this LCO. In its RAI response, the licensee stated that it agreed with the NRC staff's observation, and submitted a revision to its proposed TS pages, addressing the NRC staff's concern (Reference 2). Additionally, the licensee proposed to relocate the SR to TS Section 4.7.A.2, which coincides with the primary containment TS requirements. Therefore, should the test fail, TS 3.7.A.2 would not be satisfied and TS 3.7.A.8 would be entered as the action statement. The required action proposed is an orderly reactor shutdown to cold conditions within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in case the SR is not met. The NRC staff considers the licensee's proposed action as acceptable.

The NRC staff also notes that the proposed change originally did not include a specified frequency for performing the test, which would allow the test to be performed anytime between 18 month and 36 month intervals. In an RAI dated November 1, 2012 (ML12306A499), the NRC staff indicated to the licensee that if the plant is brought to power without performing the test during an RFO, omission of the specified 18-month frequency would allow the plant to enter a MODE where the containment integrity is required without having met the SR. This would be in conflict with SR 4.0.1. The NRC staff also noted in the RAI that in the STS (Reference 3), the frequency for SR 3.6.1.1.2 is 18 months and is required to be met in Modes 1, 2, and 3. This test frequency prevents the plant from entering from Mode 5 to Modes 1, 2, or 3 without verifying containment integrity. The licensee was requested to explain how the proposed TS change ensures that the SR is met consistently every 18 months. In its RAI response, the licensee stated that it agreed with the NRC staff's observation and proposed to include the frequency of the drywell-to-suppression chamber leak rate test of once every 18 months, consistent with SR 3.6.1.1.2 in Reference 3. The licensee submitted revised TS pages reflecting this change in its supplemental letter (Reference 4).

In addition, the licensee proposed to: (1) change the title of TS 4.7.A.6.b from "Refueling Outage Tests" to "Refueling Outage Test;" (2) renumber TS 4.7.A.6.b.3 to TS 4.7.A.6.c and TS 4.7.A.6.c.1; and (3) add the title, "Operating Cycle Test" to TS 4.7.A.6.c., to support the

-4 proposed TS change. Additionally, the licensee proposed to delete TS 4.7.A.6.b.2 since the provision was deleted by a prior amendment. The NRC staff has confirmed that these non technical, administrative changes to TS 4.7.A.6, "Pressure Suppression Chamber - Drywell Vacuum Breakers," are appropriate, editorial in nature, and involve the reorganization or reformatting of requirements without affecting the technical content of TS 4.7.A.6. Therefore, the NRC staff finds these changes acceptable.

3.3 Conclusion The NRC staff concludes that the licensee has provided adequate justification for performing the drywell-to-suppression chamber leak rate surveillance test at any time during the OC with an 18-month frequency. The NRC staff also concludes that the proposed changes meet the requirements of 10 CFR 50.36 and GDC 16, and are acceptable.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Vermont State official was notified of the proposed issuance of the amendment. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and changes SRs.

The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration, and there has been no public comment on such finding (April 3, 2012 (77 FR 20074>>. Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.

6.0 CONCLUSION

The Commission has concluded, based on the conSiderations discussed above that (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) there is reasonable assurance that such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

7.0 REFERENCES

1. Letter from Entergy Nuclear Operations, Inc to NRC, "TS Proposed Change No. 297, Suppression Chamber-Drywell Leak Rate Test Surveillance Frequency Change Vermont Yankee Nuclear Power Station, Docket No. 50-271, License No. DPR-28," dated February 1, 2012 (ADAMS Accession No. ML12037A064).

2. Letter from Entergy Nuclear Operations, Inc to NRC, "TS Proposed Change 297, Supplement 1, Response to Request for Additional Information, Vermont Yankee

~ 5~

Nuclear Power Station, Docket No. 50-271, License No. DPR-28," dated August 7,2012 (ADAMS Accession No. ML12223A099).

3. NUREG-1433 "Standard Technical Specifications General Electric BWRJ4 Plants,"

Volume 1, Revision 4, dated April 2012 (ADAMS Accession No. ML12104A192).

4. Letter from Entergy Nuclear Operations, Inc to NRC, "Technical Specifications Proposed Change 297, Supplement 2, Response to Request for Additional Information, Vermont Yankee Nuclear Power Station, Docket No. 50-271, License No. DPR-28," dated November 20,2012 (ADAMS Accession No. ML12331A290).

Principal Contributor: A. Sallman, NRR/DSS/SCVB Date: January 30, 2013

January 30, 2013 Site Vice President Entergy Nuclear Operations, Inc.

Vermont Yankee Nuclear Power Station P.O. Box 250 Governor Hunt Road Vernon, VT 05354

SUBJECT:

VERMONT YANKEE NUCLEAR POWER STATION -ISSUANCE OF AMENDMENT TO RENEWED FACILITY OPERATING LICENSE RE: CHANGE TO SUPPRESSION CHAMBER-DRYWELL LEAK RATE TEST SURVEILLANCE FREQUENCY (TAC NO. ME7928)

Dear Sir or Madam:

The Nuclear Regulatory Commission (NRC) has issued the enclosed Amendment No. 254 to Renewed Facility Operating License No. DPR-28 for the Vermont Yankee Nuclear Power Station (VYNPS), in response to your application dated February 1, 2012, as supplemented on August 7 and November 20, 2012.

This amendment revises the VYNPS Technical Specification 4.7.A.6.b.3 for performing the drywell-to-suppression chamber leak rate test during an operating cycle instead of during a refueling outage.

A copy of the related Safety Evaluation is enclosed. A Notice of Issuance will be included in the Commission's biweekly Federal Register notice.

Sincerely, IRA!

Richard V. Guzman, Senior Project Manager Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-271

Enclosures:

1. Amendment No. 254 to License No. DPR-28

2. Safety Evaluation cc w/encls: Distribution via Listserv DISTRIBUTION:

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