BVY 12-055, Technical Specifications Proposed Change 297, Supplement 1, Response to Request for Additional Information

Technical Specifications Proposed Change 297, Supplement 1, Response to Request for Additional Information

ML12223A099
Person / Time
Site:	Vermont Yankee
Issue date:	08/07/2012
From:	Wamser C Entergy Nuclear Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
BVY 12-055
Download: ML12223A099 (11)

Text

Entergy Nuclear Operations, Inc.

Vermont Yankee 10Entergy 320 Governor Hunt Rd.

Vernon, VT 802-257-7711 Christopher J. Wamser Site Vice President BVY 12-055 August 7, 2012 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001

SUBJECT:

Technical Specifications Proposed Change 297, Supplement 1 Response to Request for Additional Information Vermont Yankee Nuclear Power Station Docket No. 50-271 License No. DPR-28

REFERENCES:

1. Letter, Entergy Nuclear Operations, Inc. to USNRC, "Technical Specification Proposed Change No. 297 Suppression Chamber-Drywell Leak Rate Test Surveillance Frequency Change," BVY 12-005, dated February 1,2012

Dear Sir or Madam:

In Reference 1, Entergy Nuclear Operations, Inc. (Entergy) submitted a request for an amendment to the renewed operating license Technical Specifications (TS) for Vermont Yankee (VY) requesting change to the TS related to the drywell to suppression chamber vacuum breakers.

Attachment 1 to this submittal provides Entergy's response to questions provided by NRC staff and Attachment 2 provides revised TS and TS Bases pages reflecting the changes proposed to address NRC questions. The TS Bases page is provided for information only.

This supplement to the original license amendment request does not change the scope or conclusions in the original application, nor does it change Entergy's determination of no significant hazards consideration.

There are no new regulatory commitments being made in this letter.

Should you have any questions concerning this letter or require additional information, please contact Robert Wanczyk at 802-451-3166.

Aoo(

BV 12-055 / Page 2 of 2 I declare under penalty of perjury that the foregoing is true and correct.

Executed on August 7, 2012.

Sincerely, CJW/jmd Attachments: 1. Response to Request for Additional Information

2. Retyped Technical Specifications and Bases Pages cc: William M. Dean Regional Administrator, Region 1 U.S. Nuclear Regulatory Commission 2100 Renaissance Blvd, Suite 100 King of Prussia, PA 19406-2713 Mr. Richard V. Guzman, Project Manager Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Mail Stop O-8-C2 Washington, DC 20555 USNRC Resident Inspector Entergy Nuclear Vermont Yankee, LLC 320 Governor Hunt Road Vernon, Vermont 05354 Ms. Elizabeth Miller Commissioner Vermont Department of Public Service 112 State Street - Drawer 20 Montpelier, Vermont 05620-2601

BVY 12-055 Docket 50-271 Attachment 1 Proposed Change 297, Supplement 1 Response to Request for Additional Information

BVY 12-055 / Attachment 1 / Page 1 of 3 Response to Requests for Additional Information (RAIs)

RAI Number 1 The LAR proposes to revise the technical specification (TS) to perform the drywell-to suppression chamber leakage test during unit operation instead of performing it during refueling outage. The rationale provided in the LAR states: "This 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." Accordingly, the proposed change should be to perform the test just prior (e.g., within a day) to the refueling outage. Please provide reasons for the change to perform the test at any time during the power operation cycle.

Response

The intent of the proposed change was to allow testing anytime during the operating cycle (OC) or during the refuel outage (RFO). Per Vermont Yankee (VY) TS Definition M, the RFO is part of the OC. Our plans, should we elect to perform during the OC would be to perform the surveillance just prior to shutdown to minimize the risk of a forced shutdown and use the information from the test to assess the need for additional maintenance during the outage.

There is no technical reason for the timing of the surveillance which could be performed at any time during the OC (with the 25% interval flexibility). Entergy did not intend to limit performance of the surveillance to just prior to the RFO and thus prevent us from performing during the RFO.

Performance during the RFO may be necessary if work is scheduled to be performed on the vacuum breakers and the surveillance needs to be performed as a post maintenance test.

NUREG-1433 "Standard Technical Specifications General Electric Plants, BWR/4," requires the surveillance every 18 months and even though the Basis states "the 18 month Frequency was developed considering it is prudent that this surveillance be performed during a unit outage..." it is not a requirement of the NUREG that the surveillance be performed when shutdown.

NUREG-1433 also includes a retest frequency of 9 months should there be two consecutive test failures which would require online testing.

Based on this, no revision to the proposed change is deemed necessary.

RAI Number 2 , refers to NUREG-1433 "StandardTechnical Specifications General Electric Plants,BWRI4," Volume 1, Revision 3, Volume 2 of NUREG-1433, "Bases"provides a basis for the surveillance requirement(SR) 3.6.1.1.2 (i.e., performing the drywell-to-suppressionchamber leakage test during unit outage). It states that the 18-month frequency was developed considering that this surveillancebe performed during a uni.t outage: and also, in view of the fact that if the test failed, the component failures that might have affected this test are identified by other Primarycontainment SRs. The TS does allow a same subsequent SR test (aftera test that failed during a unit outage and components were repairedand a successful test was performed during the outage) to verify the containment performanceand assure the components that were repairedare performingsatisfactorily. As proposed,if this SR test is performed during unit operation and it were to fail, please explain how the component(s) that led to the test failure would be identified?

BVY 12-055 / Attachment 1 / Page 2 of 3

Response

The Basis for SR 3.6.1.1.2 in NUREG-1433 states "the 18-month Frequency was developed considering it is prudent that this surveillance be performed during a unit outage and also in view of the fact that component failures that might have affected this test are identified by other primary containment SRs."

This statement indicates that it is prudent because performing the surveillance during the outage eliminates the potential for an unnecessary shutdown due to a loss of primary containment integrity and also that outage testing (e.g., 10 CFR 50.55a Inservice testing, 10 CFR 50 Appendix J testing) may repair components in advance of the test and make a successful test more likely.

If the surveillance fails with the plant operating, the plant would be shutdown, the Corrective Action Program (CAP) entered and the cause of the failure determined. If the surveillance fails with the plant shutdown, the CAP will be entered and the cause of the failure determined. In both cases the cause is identified and corrected. Additionally, if the surveillance fails, a retest would be performed to verify that the primary containment was operable prior to start-up.

Based on this, no revision to the proposed change is deemed necessary.

RAI Number 3 In the case that the SR test as proposed in SR 4.7.A.6.c were to fail when performed during unit operation, the containment would become inoperable. The NRC staff notes that there are no proposed actions as part this specification that prescribes required actions to be taken under this condition. Please propose the "Action" that is required to be taken under the limiting condition for operation.

Response

VY agrees with this observation. NUREG 1433 includes this SR in Section 3.6.1.1 "Primary Containment' where the VY custom TS include the SR in TS Section 4.7.A.6 related to vacuum breaker testing requirements.

To address this VY proposes to relocate the SR to Section 4.7.A.2 to coincide with the primary containment requirements of the TS. 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. Attachment 2 provides revised TS pages reflecting this change.

RAI Number 4 The Standard Technical Specifications (STS) NUREG-1433 Volume 2, "Bases" for SR 3.6.1.1.2 notes that 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. The STS Volume 1, SR 3.6.1.1.2 requires the test to be performed at an increased frequency of nine (9) month frequency until two consecutive tests pass. Please explain why the increased surveillance frequency requirement of the STS not included in the proposed amendment.

BVY 12-055 / Attachment 1 / Page 3 of 3

Response

VY agrees and proposes to add the STS surveillance requirement to the proposed change. provides revised TS and TS Bases pages reflecting this change.

BVY 12-055 Docket 50-271 Attachment 2 Proposed Change 297, Supplement 1 Retyped Technical Specification and Bases Pages

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. At least once per power levels not to exceed Operating Cycle, a 5 Mw(t). drywell to suppression chamber leak rate test

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

4. Whenever primary changed to once every 9 containment integrity is months until two required: consecutive tests pass.

a. The leakage rate from 3. (Blank) any one main steam isolation valve (MSIV) 4. In accordance with the shall not exceed 62 PCLRTP, verify that the scfh at 44 psig (Pa); following leakage rates are within acceptable limits:

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

c. The combined leakage pathways; and rate from the secondary containment bypass c. The combined leakage rate pathways shall not for the secondary exceed 5 scfh at 44 containment bypass psig (Pa). pathways.

Amendment No. G3,45-, 4-3, --74-, 2-2-3 147

VYNPS I

3.7 LIMITING CONDITIONS FOR 4.7 SURVEILLANCE REQUIREMENTS OPERATION line is verified to be closed and conditions required by 3.7.D.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 chamber 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 valve 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. 1-2_, 2-31 149

VYNPS 3.7 LIMITING CONDITIONS FOR 4.7 SURVEILLANCE REQUIREMENTS OPERATION (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 7. 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. 4- 24-&, 2- L-,31 150

VYNPS BASES: 4.7 (Cont'd)

Each operating cycle, a leak rate test shall be performed to verify that significant leakage flow paths do not exist between the drywell and suppression chamber. The drywell pressure will be increased by at least 1 psi with respect to the suppression chamber pressure and held constant. The 2 psig set point will not be exceeded. The subsequent suppression chamber pressure transient (if any) will be monitored with a sensitive pressure gauge. If the drywell pressure cannot be increased by 1 psi over the suppression chamber pressure it would be because a significant leakage path exists; in this event the leakage source will be identified and eliminated before power operation is resumed. If the drywell pressure can be increased by 1 psi over the suppression chamber the rate of change of the suppression chamber pressure must not exceed a rate equivalent to the rate of leakage from the drywell through a 1-inch orifice. In the event the rate of change exceeds this value then the plant will be shut down, if operating, the source of leakage will be identified and eliminated before power operation is resumed. Two consecutive test failures, however, would indicate unexpected primary containment degradation; in this event, increasing the frequency to once every 9 months is required until the situation is remedied as evidenced by passing two consecutive tests.

The drywell-suppression chamber vacuum breakers are exercised in accordance with Specification 4.6.E, following termination of discharge of steam into the suppression chamber from the safety/relief valves and following any operation that causes the vacuum breakers to open. This monitoring of valve operability is intended to assure that valve operability and position indication system performance does not degrade between refueling inspections.

When a vacuum breaker valve is exercised through an opening-closing cycle, the position indicating lights are designed to function as follows:

Full Closed 2 White - On (Closed to <0.050" open)

Open 2 White - Off

(>0.050" open to full open)

Experience has shown that a weekly measurement of the oxygen concentration in the primary containment assures adequate surveillance of the primary containment atmosphere.

B. and C. Standby Gas Treatment System and Secondary Containment System Initiating reactor building isolation and operation of the standby gas treatment system to maintain at least a 0.15 inch of water vacuum within the secondary containment provides an adequate test of the operation of the reactor building isolation valves, leakage tightness of the reactor building, and performance of the standby gas treatment system. The testing of reactor building automatic ventilation system isolation valves in accordance with Technical Specification 4.6.E demonstrates the operability of these valves. In addition, functional testing of initiating sensors and associated trip channels demonstrates the capability for automatic actuation. Periodic testing gives sufficient confidence of reactor building integrity and standby gas treatment system performance capability.

Amendment No. 4-24, 144, 2-31 169