05000271/LER-1997-001, :on 970207,discovered Potential for Postulated Electrical Failure to Affect Primary Containment Integrity.Caused by Failure to Properly Coordinate Cited Design Vulnerability.Controls Implemented

:on 970207,discovered Potential for Postulated Electrical Failure to Affect Primary Containment Integrity.Caused by Failure to Properly Coordinate Cited Design Vulnerability.Controls Implemented

ML20138P683
Person / Time
Site:	Vermont Yankee File:NorthStar Vermont Yankee icon.png
Issue date:	02/27/1997
From:	Maret G VERMONT YANKEE NUCLEAR POWER CORP.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
BVY-97-32, LER-97-001, LER-97-1, NUDOCS 9703050196
Download: ML20138P683 (5)

LER-1997-001, on 970207,discovered Potential for Postulated Electrical Failure to Affect Primary Containment Integrity.Caused by Failure to Properly Coordinate Cited Design Vulnerability.Controls Implemented

Event date:
Report date:
Reporting criterion:	10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(viii) 10 CFR 50.73(a)(2)(ii) 10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(iv), System Actuation 10 CFR 50.73(a)(2)(v), Loss of Safety Function
2711997001R00 - NRC Website

text

Y"ERMONT YANKEE NUCLEAR POWER CORPORATION P.O. Box 157, Governor Hunt Road

/W{

Vernon, Vermont 05354-0157 Y

(802) 257-7711 H.y, R*

February 27,1997 BVY 97-32 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555

Reference:

(a)

License No. DPR-28 (Docket No. 50-271)

Subject:

Reportable Occurrence,No. LER 97-01, Rev. 01 As defined by 10CFR50.73, we are reporting the attached Reportable Occurrence as LER 97-01, Rev. 01.

Sincerely, VERMON YANKEE NUCLEAR POWER CORPORATION j

V 'Ah

. W Gre

. M aret Plant Manager cc:

USNRC Region 1 Administrator USNRC Resident Inspector - VYNPS USNRC Project Manager - VYNPS

@h0 l

9703050196 970227 PDR ADOCK 05000271 S

PDR 050053 ME555555.855 e

- _ _ ~ ~. _ _ _ -.

NRC Form 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB No. 3150-0104 (4-95)

EXPIRES 04/30/98 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS MANDATORY INFORMATION COLLECTION REQUEST: 50.0 HRS. REPORTED LESSONS LEARNED LICENSEE EVENT REPORT (LER)

ARE INCORPORATED INTO THE LICENSING PROCESS AND FED BACK TO INDUSTRY. FORWARD COMMENTS REGARDING BURDEN ESilMATE TO THE INFORMATION AND RECORDS MANAGEMENT BRANCH (T 6 F33), U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555-0001, AND TO THE PAPERWORK REDUCTION PROJECT (3150-0104), OFFICE OF MANAGEMENT AND BUDGET, WASHINGTON, DC 20503.

FACILITY NAME (1)

VERMONT YANKEE NUCLEAR POWER STATION DOCKET NimBER (2)

PAGE (3) 05000271 01 0F 04 TITLE (4) Inadequate design / procedural coordination allows plant operation under conditions where a single postulated electrical f ailure coincident with a LOCA could result in containment overpressure.

EVENT DATE (5)

LER NUMBER (6)

REPORT DATE (7)

OTHER FACILITIES INVOLVED (8)

MONTH DAY YEAR YEAR SEQUENTIAL REVISION MONTH DAY YEAR FACILITY NAME DOCKET NO.(S)

NUMBER NUMBER 05000 02 07 97 97 001 00 02 27 97 N/A OPERATING THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR i: CHECK ONE OR MORE (11)

MODE (9)

N 20.2201(b) 20.2203(a)(2)(v) 50.73(a)(2)(i) 50.73(a)(2)(viii)

POWER 20.2203(a)(1) 20.2203(a)(3)(i) 50.73(a)(2)(ii) 50.73(a)(2)(x)

LEVEL (10) 100 20.2203(a)(2)(i) 20.2203(a)(3)(li) 50.73(a)(2)(iii) 73.71 20.2203(a)(2)(ii) 20.2203(a)(4) 50.73(a)(2)(iv)

OTHER 20.2203(a)(2)(lii) 50.36(c)(1)

X 50.73(a)(2)(v)

(Specify in Abstract below or in NRC 20.2203(a)(2)(lv) 50.36(c)(2) 50.73(a)r2)(vii)

Form 366A)

LICENSEE CONTACT FOR THIS LER (12)

NAME TELEPHONE NO. (Include Area Code)

GREGORY A. MARET, PLANT MANAGER 802-257-7711 C(MPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)

CAUSE

SYSTEM COMPONENT MANUFACTURER REPORTABLE

CAUSE

SYSTEM COMPONENT MANUFACTURER REPORTABLE TO NPRDS TO NPRDS NA NO NA NA NA SUPPLEMENTAL REPORT EXPECTED (14)

EXPECTED M0 DAY YEAR SUBMISSION X

YES NO DATE (15) 04 30 97 (If yes, complete EXPECTED SUBMIS$10N DATE)

ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines) (16)

On 02/07/97 during an investigation for an unrelated issue, Vermont Yankee (VY) discovered the potential for a postulated electrical f ailure in the containment isolation control circuitry to challenge primary containment integrity. Plant design ensures that either the drywell vent and purge (VP) outboard isolation valves or the corresponding inboard isolation valves will close as designed given any postulated single failure. However it was recognized that the failure to close at least one of the inboard vent or purge isolation valves could challenge containment integrity. Were a Loss of Coolant Accident (LOCA) to occur, concurrent with the postulated single f ailure in the torus /drywell VP valve control circuitry, while containment inerting/deinerting was in progress, a flow path would be present which would allow a portion of the steam to bypass the available heat sink (suppression pool), potentially overpressurizing the containment pressure vessel. This flow path was possible because VY containment inerting procedures allowed simultaneous openha of the torus and drywell inboard VP paths while inerting and/or deinerting the containment. VY has established administrative <mtrols to preclude simultaneously opening both the torus and drywell vent or purge paths during normal plant operation. Cause c.nalysis ef forts continue. Because plant Technical Specifications only allow the high volume drywell and torus VP paths to be opened with the plant in cold shutdown (or for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> af ter plant startup, or for the 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> preceding shutdown); the affected circuit is tested each operating cycle; and a LOCA must occur coincident with the elce:trical f ailure; the probability that the containment overpressurization could occur is exceedingly low. Therefore this event is not considered to have presented an increased threat to public health or safety.

NRC Form 366 (4-95)

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

-. _ _ _ - _ _ ~

I NR rm 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED OMB N 3 50-0104 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS MANDATORY INFORMATION COLLECTION REQUEST: 50.0 HRS. REPORTED LESSONS LEARNED i

i LICENSEE EVENT REPORT (LER)

ARE INCORPORATED INTO THE LICENSING PROCESS AND FED BACK TO

}

INDUSTRY. FORWARD COMMENTS REGARDING BURDEN ESilMATE TO THE i

INFORMATION AND RECORDS MANAGEMENT BRANCH (T-6 F33), U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555 0001, AND TO THE I

PAPERWORK REDUCTION PROJECT (3150-0104), OFFICE OF MANAGEMENT AND BUDGET, WASHINGTON, DC 20503.

FACILITY NAME (1)

DOCKET InmbER (2)

LER NUMBER (6)

PAGE (3)

YEAR SEoUENTIAL NUMBER REV #

VERMONT YANKEE NUCLEAR POWER CORPORATION 05000271 97 001 00 02 0F 04 4

~

TEXT (If more space is required, use additional copies of NRC Form 366A) (17)

DESCRIPTION OF EVENT

i On 02/07/97 during a cause analysis investigation for an unrelated issue, while operating at 100% power, VY discovered the potential for a postulated electrical failure to affect primary containment (Ells =NH) integrity. It was postulated that a single electrical f ailure which disables the torus and drywell vent and purge (Ells =VB) inboard isolation valves (Ells =lSV) could allow en open flow path f om the drywell to the torus air space partially bypassing the primary containment heat sink.

VY was constructed with a Mark-l Containment, consisting of a drywell which houses the reactor pressure vessel, recirculation system and other major equipment; and a torus, which contains approximately 70,000 cubic feet of water as a heat sink for i

any potential steam leaks. Any steam issuing from a LOCA would normally be forced from the drywell pressure vessel to beneath the water level within the torus (see Figure 1). The condensing action achieved by the water in the torus limits the increase in primary containment pressure to less than the design pressure of the containment pressure vessel.

VY Containment isolation logic (Ells =JE) design ensures that a single component f ailure cannot prevent automatic closure of the potential containment leakage pathways. For the containment vent and purge lines the design ensures that either the i

outboard isolation or the two inboard isolations will close given any postulated single failure. However the cited electrical failure could allow two 18 inch bypass lines connecting the drywell to the torus air space to remain open, thus partially bypassing the water and reducing the condensation of the steam. The bypass can occur via the air purge inboard isolation valves and through L

the drywell and torus inboard vent valves as the same control logic provides isolation of both flow paths.

Upon discovery of this design vulnerability, VY established administrative controls to preclude simultaneously opening both the j

torus and drywell purge and vent valves during normal plant operation. The potential for the single electrical f ailure coincident with a LOCA allowing containment overpressurization is present only when both inboard isolation valves for either the containment vent or purge functions are open simultaneously. Therefore VY has offixed warning tags upon the applicable valve control sivitches prohibiting the cited configuration. VY is in the process of changing plant procedures to preclude the alignment, Technical Specifications do not forbid simultaneous opening of the drywell and torus high volume vent and purge line isolation valves with the plant pressurized. However the Technical Specifications (TS's) restrict such operations to 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> operating windows, allowing 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following startup for containment inerting, and 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to plant shutdown for plant deinerting. These are the only time periods in the past ivhere the single failure vulnerabihty posed a threat to primary containment integrity.

As the time at which the vulnerability existed was such a small fraction of plant operating time, and the affected circuit has been satisfactorily tested once each operating cycle in the past; and a LOCA must occur coincident with the postulated electrical f ailure; the probability that the postulated containment overpressurization could have occurred is extremely low.

i Further it should be recognized that the control circuits and actuating devices affected are of a fail-safe design. That is, they deenergize to actuate. The electrical f ailure postulated is, as an isolated incident, a relatively low probability failure, thus further reducing the probability of the primary containment overpressurization.

CAUSES OF EVENT 1.

The apparent cause of this event was the f ailure to properly coordinate the cited design vulnerability with appropriate administrative controls during initial plant design and licensing efforts. The cause analysis investigation for this event continues.

NRC Form 366 (4-95)

l NRC Form 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3t50-0104 (4-95)

EXPIRES 04/30/93 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THl$ MANDATORY INFORMATION CO'.LECTION REQUEST: 50.0 HRS. REPORTED LESSONS LEARNED LICENSEE EVENT REPORT (LER)

ARE INCORPORATED INTO THE LICENSING PROCESS AND TED BACK TO l

INDUSTRY. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE INFORMATION AND RECORDS MANAGEMENT BRANCH (T-6 F33), U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555-0001, AND TO THE PAPERWORK REDUCTION PROJECT (3150-0104), OFFICE OF MANAGEMENT AND BUDGET, WASHINGTON, DC 20503.

FACILITY NAME (1)

DOCKET NUMBER (2)

LER NUMBER (6)

PAGE (3)

YEAR SEQUENTIAL NUMBER REV #

VERMONT YANKEE NUCLEAR POWER CORPORATION 05000271 97 001 00 03 0F 04 l

TEXT (If more space is required, use additional copies of NRC Form 366A) (17)

ANALYSIS OF EVENT

The design bases of the Primary Containment isolation (PCIS), and Primary Containment Systems (PCS) work together to mitigate the radiological consequences of postulated accidents which could release radioactive steam into the Primary Containment pressure vessel (the drywell). Although two separate systems, PCIS and PCS are designed to function together to ensure that such material released into the drywell would be adequately retained and processed, in this event it was discovered that the configuration of PCIS logic and the associated valve and piping configurations, although consistent with PCIS design bases, introduced a scenario which could result in primary containment overpressurization, which could challenge the primary containment's ability to adequately retain radioactive material and mitigate the radiological consequences of a LOCA. This condition has existed since plant initial construction and licensing.

A 1973 analysis was performed for VY to quantify the r' ects of an undetected flow path from the containment drywell directly to the torus air space, partially bypassing the suppression pool. The analysis was performed to determine the potential effects of leaking / ajar torus-to-drywell vacuum breaker check valves (Ells =BF). The analysis determined that such a bypass flow path as small as 0.2 square feet in area coincident with an in-containment leak could cause containment overpressurization.

As previously stated, the potential for this event to manifest itself as an actual containment overpressurization requires several rare conditions / failures to occur simultaneously. The following conditions must be concurrent to threaten containment integrity.

1.

Containment inerting/deinerting must be in progress. This would allow the affected inboard isolation valves to be opened. This condition is only permitted with the plant in a cold shutdown condition or within the 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> preceding a plant shutdown, or for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following plant start up. This limits this condition to approximately one half of one percent of plant operating time.

2.

A LOCA must occur. It should be recognized that this need not be a large LOCA to present a hazard. A small to intermediate break would also pose a threat. However, a LOCA of any significant magnitude is an extremely rare event.

3.

A " hot short" or some similar failure must occur which maintains energized the "deenergize to actuate" isolation and control logics for the affected valves despite a valid isolation signal. Control circuits failing in the energized condition, although not unheard of, are rare. The "deenergize to actuate" design is typically considered a "f ail safe" configuration. This failure must occur between the time of the previous cyclic surveillance and the advent of conditions 1 and 2 above.

i It should also be recognized that the primary containment is one of four fission product barriers specifically designed b retan the radioactive materials associated with the fission process. Other barriers in place include the fuel cladding itself, the reanor coolant pressure boundary, and the secondary containment.

i Safety Sinnificance Due to tN relative rarity of each individual f ailure described above, a scenario which requires that each rare event occur simultaneously is considered of extremely low probability. Therefore this event is not considered to have presented an increased threat to public health or safety.

NRC Form 366 (4 95)

}

,hcF'orm366 U.S. NUCLEAR REGULATot:Y COMMIS$!ON APPR.;VED BY OMB NO. 3150 0104

? (4 95)

EXPIRES 04/30/93 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS MANDATORY INFORMATION COLLECTION REQUESY: 50.0 HRS. REPORTED LESSONS LEARNED LICENSEf EVENT REPORT (LER)

ARE INCORPORATED INTO THE LICENSING PROCESS AND FED BACK TO INDUSTRY. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE INFORMATION AND RECORDS MANAGEMENT BRANCH (T 6 F33), U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555 0001, AND 10 THE 4

PAPERWORK REDUCTION PROJECT (3150 0104), OFFICE OF MANAGEMENT AND BUDGET, WASHINGTON, DC 20503.

J FACILITY NAME (1)

DOCKET IAMBER (2)

LER NUMBER (6)

PAGE (3) l YEAR SEQUENTIAL NUMBER REV #

4 j

VERMONT YANKEE NUCLEAR POWER CORPORATION 05000271 97 001 00 04 of 04 TEXT (if more space is required, use additional copies of NRC Form 366A) (17) l

CORRECTIVE ACTIONS

Immtdiate Actions:

1 1.

Interim plant administrative controls have been implemented to prohibit simultaneously opening the affected l

7 containment inboard isolation valves, precluding the cited mechanism to containment overpressurization (this action is

{

j complete).

I l

2.

A Basis for Maintaining Operation (BMO) was generated which defines the cited design / procedural controls mismatch, I

citing the need for added administrative controls to preclude introducing those conditions which would challenge the containment as described herein. The BMO also cites additional hardware changes which would address the current j

vulnerability (this action is complete).

1 3

1 1

3.

An event report was initiated which requires a formal root cause analysis and corrective action recommendation. The results of this analysis, including long term corrective action recommendations will be issued in a supplement to this Licensee Event Report (expected completion date is 04/30/97).

ADDITIONAL INFORMATION

Several events reported in the past 5 years have involved original plant design and/or licensing issuss. The determination as to which of these are similar to this event will be determined following completion of the cause analysis and communicated in the supplement to this report.

i 1

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Nd W

Inboard Purge M

A a..

i g

outboard Drywell v

A..

W*c"'d h

inboard Purge Tons V w

i

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l T ^' soma Torus y%

i 1

i NRC Form 366 (4 95)