05000423/LER-1997-029, :on 970417,design Basis Concern on SGTR Analysis for Main Steam Pressure Relief Bypass Valves, Determined.Caused by Failure to Recognize Design & Single Failure Requirements.Ts Change Request Submitted by 970615

:on 970417,design Basis Concern on SGTR Analysis for Main Steam Pressure Relief Bypass Valves, Determined.Caused by Failure to Recognize Design & Single Failure Requirements.Ts Change Request Submitted by 970615

ML20141G418
Person / Time
Site:	Millstone
Issue date:	05/15/1997
From:	Peschel J NORTHEAST NUCLEAR ENERGY CO.
To:
Shared Package
ML20141G405	List: 05000423/LER-1997-029, :on 970417,design Basis Concern on SGTR Analysis for Main Steam Pressure Relief Bypass Valves, Determined.Caused by Failure to Recognize Design & Single Failure Requirements.Ts Change Request Submitted by 970615
References
LER-97-029, LER-97-29, NUDOCS 9705220358
Download: ML20141G418 (4)

LER-1997-029, on 970417,design Basis Concern on SGTR Analysis for Main Steam Pressure Relief Bypass Valves, Determined.Caused by Failure to Recognize Design & Single Failure Requirements.Ts Change Request Submitted by 970615

Event date:
Report date:
Reporting criterion:	10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition
4231997029R00 - NRC Website

text

1 1

NRC FORM 366 U.s. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO 3150-0104 14 95)

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n'a"%^Lnftfoa'atm8 'n.n'aSi".A"#se ^3;R LlCENSEE EVENT REPORT (LER) lStaV!'c'd?lf**^4tr^a?l%"?Js"s'a'Ta's7#en'c BPc'&%^%C':El"nN UM??s*%'ila* * *

(See reverse for required number of digits /charactersfor each block)

FACIUTY NAME (1)

DOCliET NUMBER (2)

PAGE131 Millstone Nuclear Power Station Unit 3 05000423 1 of 4 TITLE (4)

Design Basis Concern on Steam Generator Tube Rupture (WP.) Analysis for Main Steam Pressure Relief Bypass Valves (MSPRBV)

EVENT DATE (5)

LER NUMBER (6)

REPORT DATE (7) oTHER FACILITIES INVOLVED (8)

MONTH DAY YEAR YEAR SEQUENTIAL REVislON MONTH DAY YEAR-FActuTv Nave DOCKET NUMBER NUMBER NUMBER 04 17 97 97 029 00 05 15 97 OPERATINO 5

THis REPoRTis SUBMITTED PURSUANT To THE REQUIREMENTS oF 10 CFR 5: (Check one or more) (11)

MODE (9) 20.2201(b) 20.2203(aH2)(v) 50.73(aH2Hi) l 50.73(aH2Hvm)

POWER 000 20.2203(aH1) 20.2203(aH3Hi)

X 50.73(aH2)N 50.73(aH2)(x)

~

LEVEL (10) 20.2203(aH2Hi) 20.2203(aH3Hai)

50. 73(aH2Hni) 73.71 20 2203(aH2Hai) 20.2203(aH4) 50.73(aH2)(sv)

OTHER 20.2203(anmn) 50.36(cH1) 50.73(aH2Hv)

Specifv en Abstract below

~

20.2203(a)(2Hiv) 50.36(cH2) 50.73(aH2Hvii)

LICENSEE CONTACT FOR THis LER (12)

NAME TELEPHONE NUMBER enclude Area Cooel J,M. Pesche1, MP3 Nuclear Licensing Manager (860)437-5840 l

COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)

CAJSE SYSTEM COMPONEN T M ANUF ACTURE R FtE PORT ABLE

CAUSE

SYSTEM COMPONENT M ANUF ACTURER RE PORT ABLE TO NPROS TO NPRDS l

SUPPLEMENTAL REPORT EXPECTED (14)

EXPECTED MONTH DAY YEAR YEs X NO SUBMisslON (if yes, complete EXPECTED SUBMisslON DATE).

DATE (15)

ABSTRACT (Limit to 1400 spaces,i.e., approximately 15 smgle-spacedtypewrittenlines) (16)

On March 17,1997, with the unit in Mode 5, it was determined that between February 8,1996 and March 18,1996, the minimum number of Main Steam Pressure Relief Bypass Valve (MSPRBV) flowpaths required to depressurize the Reactor Coolant System (RCS) to satisfy the steam generator margin to overfill analysis may not have been available in the event of a Steam Generator Tube Rupture (SGTR) concurrent with a single failure. Further investigation confirmed on April 17,1997 that the minimum number of MSPRBV flowpaths required after a SGTR combined with a single failure, could not be met.

This condition existed for a period of approximately one month. This event is being reported pursuant to 10CFR50.73(a)(2)(ii)(B) as a condition that resulted in the plant being outside the design basis.

The cause of this event was a failure to recognize design and single failure requirements necessitating that the MSPRBV be included in the Technical Specifications A Technical Specification change request establishing a requirement for control of the MSPRBVs will be submitted prior to June 15,1997, 9705220358 970515 PDR ADOCK 050004E3 S

PDR

NRC FoRIVI 366 A u.s. NUCLEAR REGULATORY Commission (4 95)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION F ACILITY NAME (1)

DOCKET NUMBER (2)

LER NUMBER (6)

PAGE (3)

YEAR SEQUENTIAL REVISloN Millstone Nuclear Power Station Unit 3 05000423 NUMBER NUMBER 2 of 4 97 029 00 TEXT (11more spaceis required use additionalcopies of NRC form 366A) (17) 1.

Description of Event

On March 17,1997, with the unit in Mode 5, it was determined that between February 8,1996 and March 18,1996, the minimum number of Main Steam Pressure Relief Bypass Valve (MSPRBV) flowpaths required to depressurize the Reactor Coolant System (RCS) to satisfy the steam generator margin to overfill analysis may not have been available in the event of a Steam Generator Tube Rupture (SGTR) concurrent with a single failure. Further investigation confirmed on April 17,1997 that the minimum number of MSPRBV flowpaths required after a SGTR combined with a single failure, could not be met.

The remotely controlled MSPRBVs dump steam directly to the atmosphere using only safety grade components. The current SGTR margin to overfill analysis assumes a single active component failure and a ruptured Steam Generator, and demonstrates that RCS depressurization can be accomplished using two out of the four MSPRBV flowpaths. To ensu

,,iat two depressurization flowpaths are available under accident conditinns four MSPRBV paths are required.

The MSPRBVs are not controlled under the unit Technical Specifications.

A review of historical operating records has shown that none of the four MSPRBV pathways had been declared inoperable while the unit was operating in Mcde 1. However, while performing valve inspections between February 8, 1996 and March 18,1996, two upstream block valves (3 MSS *MOV18A & 3 MSS *MOV18B) were de-energized and closed. These valves are normally maintained open, and when de-energized will

• fail-as-is" (refer to Figure 1 for typical va!ve arrangement). As a result, when the upstream block valves were closed and de-energized for the performance of the valve inspection, the two related MSPRBV flowpaths became unavailable. When sinole failure entena is considered, the minimum two flowpaths for RCS depressurization requirement could not be met. This condition existed for a period of approximately one month and is being reported pursuant to 10CFR50.73(a)(2)(ii)(B) as a condition that resulted in the plant being outside the design basis.

II.

Cause of Event

The cause of this event was a failure to recognize design and single failure requirements necessitating that the MSPRBV be included in the Technical Specifications.

Ill, Analysis of Event This event is significant in that the unit operated for approximately one month outside the design basis. There were no adverse consequences as a result of this event in that it was not necessary to perform a RCS depressurization to support SGTR mitigation.

IV, Corrective Action A Technical Specification change request establishing a requirement for control of the MSPRBVs will be submitted prior to June 15,1997.-

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b NRC FoRWI 366A U.s. NUCLEAR REGULATORY Commission 1

r4-95)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME (1)

DOCKET NUMBER (2)

LER NUMBER (6)

PAGE (3)

YEAR SEQUENTIAL REVISION Millstone Nuclear Power Station Unit 3 05000423 NUMBER NUMBER 3 of 4 97 029 00 TEXT (11 more space is required. use additionalcopies o.

.nm 366A) (17)

V.

Additional Information

l None

Similar Events

1 LER 96-013-01 " Residual Heat Removal Svstem Desian Deficiency Due to NonconservativeOriainal Desian Assumptiorf This LER involved a design deficiency in the Residual Heat Removal System ( RHS ) that was outside i

the design basis of the plant. A loss of control air could cause the RHS control valves to fail open.

i LEP 96-006-00 " Plant Shutdewn Reauired by Technical Specifications for Auxi;,ary Feedwater Containment isolation Valves Declared Inoperable "

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i Manuf acturer Data Ells System Code Main / Reheat Steam System..

..SB Ells ComDonent Code Relief Valve......

................RV a

j 1

NRC FORS 366A usi U.S. NUCLEAR REGULATORY COMMISSION LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION F ACILITY NAME (1)

DOCKET NUMBER (2)

LER NUMBER (6)

PAGE (3)

YEAR SEQUENTIAL REVISION i

Millstone Nuclear Power Station Unit 3 05000423 NUMBER NUMBER 4 Of 4 97 029 00 TEXT lif more space is required, use additionalcopies of NRC Form 366A) (17)

Fiaure _1_

Reactor Coolant System Decay Heat Removal Via Main Steam Diagram a

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