05000293/LER-1981-043-01, /01T-0:on 810713,outer Drywell Personnel Access Door Discovered to Have Leak Rate Exceeding Tech Spec Limits During Sp.Caused by Leaking Seal Due to Normal Inservice Degradation.Leak Rate Test Procedures Revised

/01T-0:on 810713,outer Drywell Personnel Access Door Discovered to Have Leak Rate Exceeding Tech Spec Limits During Sp.Caused by Leaking Seal Due to Normal Inservice Degradation.Leak Rate Test Procedures Revised

ML20010G277
Person / Time
Site:	Pilgrim
Issue date:	09/03/1981
From:	Whitney G BOSTON EDISON CO.
To:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
Shared Package
ML20010G262	List: 05000293/LER-1981-043, Forwards LER 81-043/01T-0 05000293/LER-1981-043-01, /01T-0:on 810713,outer Drywell Personnel Access Door Discovered to Have Leak Rate Exceeding Tech Spec Limits During Sp.Caused by Leaking Seal Due to Normal Inservice Degradation.Leak Rate Test Procedures Revised
References
LER-81-043-01T, LER-81-43-1T, NUDOCS 8109150515
Download: ML20010G277 (2)

LER-1981-043, /01T-0:on 810713,outer Drywell Personnel Access Door Discovered to Have Leak Rate Exceeding Tech Spec Limits During Sp.Caused by Leaking Seal Due to Normal Inservice Degradation.Leak Rate Test Procedures Revised

Event date:
Report date:
2931981043R01 - NRC Website

text

U. 5. NUCLE AR REGULATORY COMMISSloN NRC FORM 366 *

(7 77)

LICENSEE EVENT REPORT CONTROL BLOCK: l l

l l

l l

lh (PLE ASE PRINT oR TYPE ALL R EoulRED INFoRMATioN) lo l1l lM l lP lP lS l 1l@l 0 l0 l-l 0 l0 l0 l0 l 0 l -l 0l 0 l@[ 4 l 1 l 1 l 1 l 1 l@[ l lg 7

8 9 LICENSEE CODE 14 15 LICENSE NUMBER 25 26 LICENSE TYPE JO 57 CAT b8 CON'T lolil s$ReE dl 01510 l-101219 l3 @l Ol 71 Il 3l 811 @l 0 l 9 l 013 l 8 l1 }@

7 8

60 El DOCKET NUMBEp 68 69 EVENT DATE 74 75 REPORT DATE 80 EVENT DESCRIPTION AND PROB ABLE CONSEQUENCES h to l2l l On July 13, 1981 during reactor startup, while conducting surveillance procedure l

No. 8.7.1.7, the outer drywell personnel access door was discovered to have a leak l

lo lal l g o l4 j l rate in excess of T.S. limits. The inner door seal had been proven to be intact.

l The seal was replaced in-kind and a satisfactory leak rate test c.onducted at 1335 l

g o l5l j lo is l l hrs. the same day.

See Attachment l

l lc l7l l 1

l 0 l8 l l 80 E

CODE SUSC E COMPONENT CODE SLB ODE SU E

loisl

'l S I A lh [Ejh [Bjh l V I E l S l S l E l L lh dh W h 7

8 9

10 ft_

32 13 18 19 20 SiOUE NTI AL OCCURRENCE REPORT REVISION r_

LE R/RO EVENTYEAR st EPORT NO.

CODE TYPE N O.

h,aEg l218l 1l

[- _j l 0l 4l 3l l/l l0 l1 l lTl l---]

J 22 23 24 26 27 28 29 30 31 32 V

A N ACT ON ON PL NT MET D HOURS

$8 1 D

FOR 8.

SUPPLI R MAN FAC URER Q{34G_j@ d@

(_Z_jh l7 Ol01010l [Y_j@

lN l@' l Nl@

lCl3l11Ol@

33 35 36 3

40 44 42 43 44 47 CAUSE DESCRIPT3ON AND CORRECTIVE ACTIONS 47 l 3 g o l l The cause of the seal leakage has been determined to be normal in-service seal l

, ;i j l degradation. A change to procedure 2.1.1, which requires a local leak rate test priori to Reactor C ulant Temperature exceeding 212 F. has been made to preclude recurrence l i Ii 12 I I gi;3; l of this event.

l I

l1 lAll 7

8 9 STA

% POWER OTHER STATUS 5O RY DISCOVERY DESCRIPTION l

[_XJ h l 01 Ol 1l h l Startuo I

W hl surveillnnce Test 1 5 A.CTivlTY CO TENT Q

RELEASED OF RELE ASE AMOUNT OF ACTIVITY D LOCATION OF RELE ASE N.A.

l l

N.A.

l l1 l6 l [ Z_] h [Zj@l 7

8 9 10 11 44 45 80 PERSONNEL EXPOSURES NUMBER TYPE

DESCRIPTION

N.A.

l l 1 l 71 l 0 l 0 l 0 lhl Z hl PE RSONNE L INJURIES oEsCRiPriON@

NUMBER N.A.

l l1 l 8 l l 0 l 0 l 0 l@l 80 7

8 9 11 12 LOSS OF OR DAM AGE TO FACILITY TYPE DESCr8PTION N.A.

l

[_Z_jhl

• 1 9 80 7

8 9 10 NRC USE ONLY

"'"'" @ 8109150515 810903 l

hDRADOCK05000 N.A.

l lllllllllllll l2l0l 68 69 80 3 7

8 9 10 617-746-7900 G. G. Witney PHONE:

NAME OF PREPARER

\\

e

'.s BOSTON EDISON COMPANY PILGRIM NUCLEAR POWER STATION l'

DOCKET NO. 50-293 Attachment to LER 81-043/01T-0 On July 15, 1981, the Operations Review Committee (ORC) determined that degradation of the outer drywell access door seal was-reportable per 10 CFR 50 Appendix J.

This determination was based on available information. However, the ORC re-quested the On-site Safety Group Leader to prepare a detailed report of the criteria for testing the drywell door seals. On August 19, 1981 the ORC reviewed this report and determined an immediate event report be issued since operating procedure 2.1.1 permitted (and operating personnel performed) a drywell entrance for a routine, scheduled inspection after reactor coolant temperature exceeded 2120 F and before a satisfactory surveillance test of the drywell access.

door seals. To preclude a recurrence of this event, procedure 2.1.1 has been

[;

revised to require a local leak rate test of the door seals prior to a coolant temperature in excess of 2120F as well as after the required drywell inspection.

t l

\\