05000336/LER-1996-037, :on 961202,inadequate Surveillance Procedure for Verifying Average Water Temp at Unit 2 Intake Structure Discovered.Caused by Misinterpretation of TS Requirements.Ts Rev Initiated

:on 961202,inadequate Surveillance Procedure for Verifying Average Water Temp at Unit 2 Intake Structure Discovered.Caused by Misinterpretation of TS Requirements.Ts Rev Initiated

ML20133C636
Person / Time
Site:	Millstone
Issue date:	12/31/1996
From:	Laudenat R NORTHEAST UTILITIES
To:
Shared Package
ML20133C603	List: 05000336/LER-1996-037, :on 961202,inadequate Surveillance Procedure for Verifying Average Water Temp at Unit 2 Intake Structure Discovered.Caused by Misinterpretation of TS Requirements.Ts Rev Initiated
References
LER-96-037, LER-96-37, NUDOCS 9701080017
Download: ML20133C636 (3)

LER-1996-037, on 961202,inadequate Surveillance Procedure for Verifying Average Water Temp at Unit 2 Intake Structure Discovered.Caused by Misinterpretation of TS Requirements.Ts Rev Initiated

Event date:
Report date:
Reporting criterion:	10 CFR 50.73(a)(2)(viii) 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications
3361996037R00 - NRC Website

text

__.

..m m -

NRC FORM 366 U.s. RUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3160-0104 i

e EXPIRES 04/30/98

'YoE'#cMcI'a7o'uY'I 5'U$*M"o'E LICENSEE EVENT REPORT (LER)

'" a"!82,ER*:8Wa'a tiLTu'n'<"h"as"3R (M ut s'

u I

(See reverse for re uired number of E' N'a LEU EM"EI*'0""

o""YoS!E l

digits / characters or each block)

N^ '

"T ^N

" o' 8

FOCIUTY NAME 0)

DOCKET NUMBER (2)

PAGE (3)

Millstone Nuclear Power Station Unit 2 05000336 1OF3 i

l i

TITLE 44) j inadequate Surveillance Procedure for Verifying Average Water Temperature at the Unit 2 Intake Structure EVENT DATE (5)

LER NUM8ER (6)

REPORT DATE (7) oTHER FACILITIES INv0LVED (8)

]

SE

^L E $ N MONTH DAY YEAR YEAR MONTH DAY YEAR NU NU R

" "^"'

12 02 96 96

-- 037 --

00 12 31 96 OPERATING THIS REPORT IS SU8MITTED PURGUANT TO THE REQUIREMENTS OF 10 CFR l' (Check one or more) (11)

MODE (9) 5 20.2201(b) 20.2203(a)(2)(v)

X So.73(a)(2)(i) 50.73(a)(2)(viii)

POWER 20.2203(a)(1) 20.2203(a)(3)(i)

So.73(a)(2)(ii)

So.73(a)(2)(x)

LEVEL (10) 20.2203(a)(2)(i) 20.2203(a)(3)(ii)

So.73(a)(2)(iii) 73.71 20.2203(a)(2)(ii) 20.2203(a)(4)

So.73(a)(2)(iv)

OTHER 20.2203(a)(2)(iii)

So.36(c)(1) 5n.73(aH2)(v)

SpecgAbstr t elow 20.2203(a)(2)(iv)

So.36(c)(2)

So.73(a)(2)(vii)

LICENSEE CONTACT FOR THIS LER (12)

NAME TELEPHONE NUMBER linclude Area Codel i

l R. T. Laudenat, MP2 Nuclear Licensing Manager (860) 444 5248 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)

"T^

T^

CAUSE

SYSTEM COMPONENT MANUFACTURER

CAUSE

SYSTEM COMPONENT MANUFACTURER PRDS pRD SUPPLEMENTAL REPORT EXPECTED (14)

EXPECTED MONTH DAY YEAR YES SUBMISSION X NO DATE (15) j (If yes, complete EXPECTED SUBMISSION DATE).

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

On December 2,1996, it was identified that surveillance procedure (SP) 2619A, " Control Room Shift Checks," did not adequately refiert the requirements of Technical Specification Surveillance Requirement 4.7.11 for determining ultimate heat sink [BS) temperature in Modes 1,2,3, and 4. Although contpliance to the Technical Specification Surveillance Requirement was not met in this event, the basis for the Technical Specification requirement was met by ensuring that the service water temperature did not exceed the design limit of 75 degrees F.

The cause of this event was that the surveillance was prepared to meet the intent nf Technical Specification 3.7.11, rather than to comply with the Technical Specification.

NO immediate actions were required as a result of this event. Technical Specification 3.7.11 does not apply in the current plant Mode. As a result of this event, a Technical Specification revision will be initiated to remove the requirement to tcke the average temperature of the ultimate heat sink at the Unit 2 intake structure. This revision will be subm;aed to the NRC by March 31,1997, 9701000017 961231 PDR ADOCK 05000336 S

PCR WRC FORM 366 (4-95)

-.U.S. NUCLEAR REGULATORY COMMisslON (4-93)

UCENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME (1)

DOCKET LER NUMBER i6)

PAGE (3)

SEQUENTIAL REVISION YEAR NUMBER NUMBER 2OF3 Millstone Nuclear Power Station Unit 2 05000336 96 037 -

00 TEXT (11 more space is required, use additional copies of NRC Form 366A) (17) 1.

Description of Event

On December 2,1996, it was identified that surveillance procadure (SP) 2619A, " Control Room Shift Checks,"

did not adequately reflect the requirements of Technical Specification Surveillance Requirement 4.7.11 for determining ultimate heat sink [BS) temperature in Modes 1,2,3, and 4. At the time of discovery of this event, the unit was in Mode 5 at 0 percent power.

Techn; cal Specification 3.7.11 was added by Amendment 145 (issued June 12,1990) and requires that the ultimate heat sink be operable with an average water temperature of less than or equal to 75 degrees Fahrenheit (F) at the Unit 2 intake structure [MK). This change reflected the assumption made for the ultimate heat sink in the design basis analysis for maximum service water system [Bl] temperature. The associated surveillance requirement (4.7.11) states the following:

The ultimate heat sink shall be determined OPERABLE:

a. At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by verifying the average water temperature at the Unit 2 intake structure to be within limits.

b.

At least once per 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> by verifying the average water temperature at the Unit 2 intake structure to be within limits when the average water temperature exceeds 70'F.

SP 2619A utilizes the higher of one temperature instrument at the Unit 2 intake structure or the average circulating water system [KE] inlet waterbox temperature until the temperature exceeds 70 degrees F. (The circulating water system takes its suction at the Unit 2 intake structure.) Above 70 degrees F, temperature measurements are recorded from local indicators in the service water header inside the turbine building [NM).

The one temperature instrument at the Unit 2 intake structure does not provide an average water temperature and the local service water system header temperature instruments are not located at the Unit 2 intake structure.

Therefore, the requirements of Technical Specification 4.7.11 were not being met by SP 2619A. Previous revisions to SP 2619A also utilized the circulating water system inlet waterbox temperature without considering the one temperature instrument at the Unit 2 intake structure.

This event is being reported in accordance with 10 CFR 50.73(a)(2)(i)(B), any operation or condition prohibited by the plant's Technical Specifications.

11. Cause of Event

The cause of the event was that the surveillance was prepared to meet the intent of Technical Specification 3.7.11, but failed to comply with the Technical Specification i imiting Condition for Operation (LCO).

The Ultimate Heat Sink Technical Specification was initiated in 1990 to ensure that the Service Water inlet temperatures would be monitored in order to ensure that the maximum temperature used in the accident analysis would not be exceeded. Although the surveillance met the intent of this requirement, the conclusion that it met the LCO was inappropriate.

Ill. Analysis of Event The ultimate heat sink consists of the Long Island Sound and provides the cooling water necessary to ensure that heat removal capacity exists for normal cooldown and for the mitigation of design basis accidents. The service water system takes its suction from the ultimate heat sink at the intake structure and provides safety related N*.C FORM 366A (4-9M

WRC FORM 366A U.s. NUCLEAR REGULATORY CoMMISslON (4 95) e LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME (1)

DOCKET LER NUMBER i6)

PAGE (3) i SEQUENTIAL REVISION YEAR NUMBER NUMBER 3OF3 Millstone Nuclear Power Station Unit 2 05000336 96 - 037 --

00 1

TEXT fit more space is required, use additional copies of NRC Form 366A) (17) cooling water for several systems including the reactor building closed cooling water system [CC] and the diesel generators [EK]. The maximum ultimate heat sink temperature limit in Technical Specification 3.7.11 is based upon the design analysis for the service water system during accident conditions which assumes a maximum of 75 degrees F service water temperature.

Although compliance with the Technical Specification Surveillance requirement was not met in this event, the basis for the Technical Specification requirement was met by ensuring that the service water temperature did not exceed the design limit of 75 degrees F. Therefore, this event was not safety significant.

This event demonstrates the need to review Technical Specification surveillance procedures. Technical Specification surveillance requirements will be verified as part of a program to support the response to Notice of 4

Violation (NOV) 336/96-08-07.

IV. Corrective Action

No immediate actions were required as a result of this event since Technical Specification 3.7.11 does not apply in the current plant Mode. As a result of this event, a Technical Specification revision will be initiated to remove the requirement to take the average temperature of the ultimate heat sink at the Unit 2 intake structure. Included in this is a revision of the Technical Specification Bases to identify instrumentation to be used to perform the Technical Specification surveillance. This revision will be submitted to the NRC by March 31,1997.

Technical Specification surveillance procedures wili be reviewed to ensure compliance with Technical Specification surveillance requirements as part of the Millstone Unit No. 2 Operational Readiness Plan. The review will initially focus on Technical Specification surveillance procedures required for Mode 6 and defueled.

Surveillance procedures required for subsequent mode changes will be reviewed prior to mode entry.

(The above commitment was previously sent to the NRC in the response to NOV 336/96-08-07, NNECO Commitment No. B16076-2.)

V.

Additional information

Similar Events Previous LERs that involve deficient surveillance procedures include:

LER 96-023-00: Failure to Perform Technical Specifications Surveillances on Certain Contaimnent Isolation Valves LER-96-024-00: Response Time Testing of RPS and ESAS Failed to include Response Time of SPEC 200 Electronics LER 96-025-00: Enclosure Building Filtration Actuation Signal / Auxiliary Exhaust Actuation Signal Interlock Not Tested Periodically LER-96-026-00: Incomplete Technical Specification Required Surveillance - Valve Lineups Inside Containment LER 96-035-00: Failure to Perform Periodic Surveillance Testing for interlock Function Associated with the Main Steam Isolation System Function of the Engineered Safeguards Actuation System Energy Industry Identification System (Ells) codes are identified in the text as [XX].

MRC FORM 366A (4-95)

]