05000293/LER-2011-004, For Pilgrim Nuclear Power Station, Regarding Technical Specification Required Shutdown - Drywell to Torus Dp

For Pilgrim Nuclear Power Station, Regarding Technical Specification Required Shutdown - Drywell to Torus Dp

ML11228A028
Person / Time
Site:	Pilgrim
Issue date:	07/12/2011
From:	Rich Smith Entergy Nuclear Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
2.11.044 LER 11-004-00
Download: ML11228A028 (8)

LER-2011-004, For Pilgrim Nuclear Power Station, Regarding Technical Specification Required Shutdown - Drywell to Torus Dp

Event date:
Report date:
Reporting criterion:	10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(viii)(B) 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(ix)(A) 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications
2932011004R00 - NRC Website

text

SEn tergy Entergy Nuclear Operations, Inc.

Pilgrim Nuclear Power Station 600 Rocky Hill Road Plymouth, MA 02360 Robert G. Smith, RE.

Site Vice President July 12, 2011 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555

SUBJECT:

Entergy Nuclear Operations, Inc.

Pilgrim Nuclear Power Station Docket No.: 50-293 License No.: DPR-35 Licensee Event Report 2011-004-00, 'Technical Specification (TS) Required Shutdown - Drywell to Torus DP" LETTER NUMBER: 2.11.044

Dear Sir or Madam:

The enclosed Licensee Event Report (LER) 2011-004-00, "Technical Specification (TS) Required Shutdown

- Drywell to Torus DP" is submitted in accordance with 10 CFR 50.73.

This letter contains no commitments.

Please do not hesitate to contact Mr. Joseph R. Lynch, (508) 830-8403, if there are any questions regarding this submittal.

Sincerely, Robert G. Smith, P.E.

Site Vice-President FXM/fxm Attachment: Licensee Event Report 2011-004-00, "Technical Specification (TS) Required Shutdown -

Drywell to Torus DP" (5 Pages)

PNPS Letter 2.11.044 Page 2 of 2 cc:

Mr. William M. Dean Regional Administrator, Region 1 U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406-1415 INPO Records 700 Galleria Parkway Atlanta, GA 30399-5957 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 11555 Rockville Pike Rockville, MD. 20852 USNRC Senior Resident Inspector Pilgrim Nuclear Power Station

Attachment 1

Letter Number 2.11.044 Licensee Event Report 2011-004-00, "Technical Specification (TS) Required Shutdown - Drywell to Torus DP" (5 pages)

NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 10/31/2013 (10-2010)

Estimated burden per response to comply with this mandatory collection request: 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br />.

Reported lessons learned are incorporated into the licensing process and fed back to industry.

Send comments regarding burden estimate to the FOIA/Privacy Service Branch (T-5 F53), U.S.

Nuclear Regulatory Commission, Washington, DC 20555-0001, or by internet e-mail to LICENSEE EVENT REPORT (LER) infocollects.resource@nrc.gov, and to the Desk Officer, Office ot Information and Regulatory Affairs, NEOB-10202, (3150-0104), Office of Management and Budget, Washington, DC 20503. If a means used to impose an information collection does not display a currently valid OMB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.

3. PAGE Pilgrim Nuclear Power Station 05000293 1 OF5

4. TITLE Technical Specification (TS) Required Shutdown - Drywell to Torus DP

5. EVENT DATE

6. LER NUMBER

7. REPORT DATE

8. OTHER FACILITIES INVOLVED MONTH DAY YEAR YEAR SEQUENTIAL REV MONTH DAY YEAR FACILITY NAME DOCKET NUMBER NUMBER NO

__I N/A 05000 FACILITY NAME DOCKET NUMBER 05 14 2011 2011 004 00 07 12 2011 N/A 05000

9. OPERATING MODE

11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply) 20.2201 (b) 20.2203(a)(3)(i) 50.73(a)(2)(i)(C) 50.73(a)(2)(vii)

N 20.2201(d) 20.2203(a)(3)(ii) 50.73(a)(2)(ii)(A) 50.73(a)(2)(viii)(A) 20.2203(a)(1) 20.2203(a)(4) 50.73(a)(2)(ii)(B) 50.73(a)(2)(viii)(B) 20.2203(a)(2)(i) 50.36(c)(1)(i)(A) 50.73(a)(2)(iii) 50.73(a)(2)(ix)(A) 20.2203(a)(2)(ii) 50.36(c)(1 )(ii)(A) 50.73(a)(2)(iv)(A) 50.73(a)(2)(x) 1 O L20.2203(a)(2)(iii) 50.36(c)(2) 50.73(a)(2)(v)(A) 73.71 (a)(4) 20.2203(a)(2)(iv) 50.46(a)(3)(ii) 50.73(a)(2)(v)(B) 73.71 (a)(5) 14%

20.2203(a)(2)(v) j 50.73(a)(2)(i)(A) 50.73(a)(2)(v)(C)

OTHER 20.2203(a)(2)(vi)

]

50.73(a)(2)(i)(B) 50.73(a)(2)(v)(D)

Specify in Abstract below or in

BACKGROUND:

The safety objective of the Primary Containment System (PCS) is to provide the capability, in conjunction with other safeguard features, to limit the release of fission products in the event of a design basis accident so that offsite doses would not exceed the guidelines set forth in 10 CFR 100. The PCS design employs a low leakage suppression containment system that houses the Reactor Vessel (RV), the Reactor Recirculation System loops, and other branch connections of the Reactor Primary System.

The PCS is designed to withstand the forces from any size breach of the nuclear system primary barrier up to and including an instantaneous circumferential break of the reactor recirculation piping, and provides a holdup time for decay of any radioactive material released. The PCS also stores sufficient water to condense the steam released as a result of a breach in the nuclear system primary barrier and to supply the Core Standby Cooling Systems (CSCS).

The Venting and Vacuum Relief System is part of the PCS design. The purpose is to equalize the pressure between the Drywell and the Torus and the Torus and the Reactor Building so the structural integrity of containment is maintained. Drywell to Torus vacuum breakers and Torus to Reactor Building vacuum breakers are provided for this purpose.

The Vacuum Relief System for the Drywell to the Torus consists of ten (10) vacuum relief valves (i.e., vacuum breakers). Vacuum breakers (X-201A thru K) are passive, normally closed and are required to open to relieve excessive Drywell to Torus differential pressure. These vacuum breakers are sized to limit the differential pressure between the Drywell and Torus during post accident Drywell cooling operations to the design limit of 2.0 psig (i.e., the external design pressure). The vacuum relief function can be ensured with two vacuum relief valves secured in the closed position and eight open valves. Instrumentation is provided to monitor the position status of the valves.

The Drywell to Torus vacuum breakers were also analyzed in safety analyses to remain closed when not needed for vacuum relief. These valves must remain closed to ensure the following functions are satisfied:

1. maintain a 1.17 psid differential pressure between the Drywell (DW) and Torus (or Wetwell, WW) airspace; and

2. limit steam bypass leakage from the Drywell to the Torus airspace.

The DW-WW differential pressure requirement was implemented during the Mark I Program and its purpose is to reduce containment loads and stress resulting from the suppression pool swell caused by Loss of Coolant Accident (LOCA) blowdown. The normal design requirement for LOCA blowdown loads requires stresses to remain below Service Level A limits from ASME Section II1. Per TS 3.7.A.8.a.i requirements, the 1.17 psid differential pressure between the DW-WW airspace must be established when the mode switch is in RUN and within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after Core Thermal Power (CTP) is greater than 15% following startup.U.S. NUCLEAR REGULATORY COMMISSION (10-2010)

LICENSEE EVENT REPORT (LER)

CONTINUATION SHEET

1. FACILITY NAME

2. DOCKET

6. LER NUMBER

3. PAGE YEAR SEQUENTIAL REV Pilgrim Nuclear Power Station 05000293 NUMBER NO.

3 OF 5 2011 -

004 00 The purpose of limiting steam bypass leakage is to protect the pressure suppression function of Primary Containment by preventing excessive leakage from the vent system directly to the WW airspace. Direct leakage to the WW airspace increases WW temperature and pressure which in turn raises Drywell pressure during a LOCA. The analysis performed for the limiting scenario (a small break) is described in FSAR Section 5.2.4.11 and TS 3.7 Bases. The analytical maximum allowable bypass area is approximately 0.2 ft2 [Ref. FSAR 5.2]. This is equivalent to a 6 inch diameter orifice. This function is ensured by TS 3.7.A.4 and associated surveillances.

TS 3.7.A.4 specifies the operability requirements for the Drywell to Torus vacuum breakers. Essentially, the vacuum breakers are required to be operable when Primary Containment is required except during testing and other certain conditions. Primary Containment is required when the reactor is critical or when the reactor vessel water temperature is greater than 212 degrees F except when performing certain low power tests. TS 3.7.A.5 specifies that if TS 3.7.A.1 thru 4 requirements can not be met, an orderly shutdown shall be initiated and the reactor shall be in Cold Shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Drywell to Torus vacuum breaker opening force testing was performed during Refueling Outage (RFO) 18 to verify the force necessary to open each vacuum breaker was within acceptance limits. This testing revealed that vacuum breakers X201 D, X-201 G, and X-201 J required adjustment and rework in order to meet opening force test acceptance criteria.

EVENT DESCRIPTION

At 0245 hours0.00284 days <br />0.0681 hours <br />4.050926e-4 weeks <br />9.32225e-5 months <br /> on May 14, 2011, with the plant operating at approximately 14% power and the mode switch in RUN, Pilgrim Nuclear Power Station (PNPS) commenced a controlled shutdown of the reactor due to inoperable Drywell to Suppression Chamber (Torus) Vacuum Breakers. The Drywell to Torus vacuum breakers were declared inoperable due to inability to set the conditions necessary to demonstrate that Technical Specification limits for Drywell to Torus differential pressure decay rate were satisfied.

This event was initially reported to the NRC via Event Report #46852 on 5/14/2011 pursuant to 10 CFR 50.72(b)(2)(i).

CAUSE

The direct cause of the event was determined to be improper sealing of three (3) Drywell to Torus vacuum breakers (X-201 D, X-201 G, and X-201 J) due to improper magnet to striker plate clearance. The root cause of the event identified that the vacuum breakers were incorrectly adjusted during maintenance in RFO 18 because the procedure lacked necessary instruction.

EXTENT OF CONDITION:

An extent of condition was performed on similar vacuum breakers and associated procedures. This review evaluated the Reactor Building to Torus vacuum breakers X-212A/B and the eight (8) safety relief valve discharge line vacuum reliefs (VRV-261-97A/B/C/D and VRV-261-98A/B/C/D). No similar concerns were noted.

CORRECTIVE ACTIONS

Completed Actions:

1. Correctly adjusted the breaker strike plate to magnet clearances on the X-201D, X-201G and X201J Drywell to Torus vacuum breakers. Proper valve sealing was verified after this adjustment.U.S. NUCLEAR REGULATORY COMMISSION (10-2010)

LICENSEE EVENT REPORT (LER)

CONTINUATION SHEET

1. FACILITY NAME

2. DOCKET

6. LER NUMBER

3. PAGE YEAR SEQUENTIAL REV Pilgrim Nuclear Power Station 05000293 NUMBER NO.

4 OF 5 2011-004 00 Open Actions:

1. Develop a new procedure or modify the existing maintenance procedure to provide enhanced instruction for maintaining and adjusting vacuum breakers.

2. Review the Root Cause Analysis during Continuing Training and use as a case study for all maintenance disciplines.

This event and the associated corrective actions were entered into the Site Corrective Action Program.

ASSESSMENT OF SAFETY CONSEQUENCES

The event posed no threat to public health and safety.

The safety objective of the Primary Containment System (PCS) is to provide the capability, in conjunction with other safeguard features, to limit the release of fission products in the event of a design basis accident so that offsite doses would not exceed the guidelines set forth in 10 CFR 100. The degraded condition where three Drywell to Torus vacuum breakers did not fully close did not preclude capability of the PCS to perform it's required function.

The failure of three Drywell to Torus vacuum breakers to seal closed does not affect capability of these valves to equalize pressure between the Drywell and the Torus. Therefore the vacuum breaker function necessary to protect the structural integrity of the containment from external pressure was maintained.

The condition where Drywell to Torus differential pressure could not be maintained at 1.17 psid without assistance from the Drywell and Torus Ventilation System fans did not impact plant safety. Safety analysis and TS 3.7.A.8 requirements do not require the 1.17 psid differential pressure to be maintained until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after Core Thermal Power is raised above 15%. This TS Limiting Condition for Operation (LCO) was not exceeded.

The steam bypass leakage limit from the Drywell to the Torus airspace is based on analyses that are described in the FSAR and TS Bases. The purpose of limiting bypass leakage is to protect the pressure suppression function of Primary Containment by preventing excessive steam leakage during a LOCA from the Drywell vent system directly to the WW airspace. Direct steam leakage to the WW airspace increases WW temperature and pressure which in turn raises Drywell pressure during a LOCA. The analysis performed for the limiting scenario (a small break) is described in FSAR Section 5.2.4.11 and TS 3.7 Bases. The analytical maximum allowable bypass area is 0.2 ft2 which is equivalent to a 6 inch diameter orifice. Based on the FSAR analysis, reactor operation is permissible if the bypass area does not exceed the allowable bypass area. The equivalent leak path opening between the Drywell and Torus was estimated to be less than a 2 inch diameter pipe and the reactor was operated at reduced power levels. This indicates that the leak path was within the analytical maximum allowed value described in the FSAR and TS Bases. However, since the drywell to suppression chamber leakage did not satisfy the more conservative TS 4.7.A.4.b.4 limit (i.e., does not exceed differential pressure decay rate which would occur through a 1 inch diameter opening without the addition of air or nitrogen), the vacuum breakers were declared inoperable and the reactor was shutdown per TS 3.7.A.5.

For risk assessment purposes it was conservatively assumed that at least one Torus to Drywell vacuum breaker was unavailable, fully stuck open, and allowed bypass of the pressure suppression function in the event of a large break LOCA. An incremental core damage probability (ICDP) of 8.28E-8 was calculated. Since the ICDP was less than 1 E-7, the impact on external events and large early release metrics is not considered significant.U.S. NUCLEAR REGULATORY COMMISSION (10-2010)

LICENSEE EVENT REPORT (LER)

CONTINUATION SHEET

1. FACILITY NAME

2. DOCKET

6. LER NUMBER

3. PAGE YEAR SEQUENTIAL REV Pilgrim Nuclear Power Station 05000293 NUMBER NO.

5 OF 5 2011 -

004 00 The vacuum breakers were repaired in accordance with an approved repair plan and there is no long term negative effect to the vacuum breakers.

REPORTABILITY

This LER is submitted pursuant to the requirements of 50.73(a)(2)(i)(A) because a Technical Specification required shutdown was completed.

SIMILAR EVENTS

A review was conducted of Pilgrim Station Licensee Event Reports (LERs) submitted since January 1994.

The review focused on LERs which involved the Drywell to Torus vacuum breakers and requirements to shutdown the plant. The review identified that several LERs were issued that involved the Drywell to Torus vacuum breakers but none involved failure of the vacuum breaker to close resulting in a TS plant shutdown.

The following LERs were reviewed: LER 96-01 involved a pressure switch on the Reactor BuUding to Torus vacuum breakers; LERs 95-05 and 95-04 involved inadvertent opening of the vacuum breakers during plant operation; and LER 94-07 involved instrument lines associated with Drywell and Torus dP monitoring instrumentation.

FAILED COMPONENT IDENTIFICATION:

The following EIIS codes are applicable to this report:

COMPONENTS CODES Breaker, Vacuum VACB SYSTEMS CODES Containment Vacuum Relief System BF