05000287/LER-1981-009, Forwards LER 81-009/03L-0.Detailed Event Analysis Encl

Forwards LER 81-009/03L-0.Detailed Event Analysis Encl

ML20009C099
Person / Time
Site:	Oconee
Issue date:	06/05/1981
From:	Parker W DUKE POWER CO.
To:	James O'Reilly NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
Shared Package
ML20009C101	List: 05000287/LER-1981-009, Forwards LER 81-009/03L-0.Detailed Event Analysis Encl 05000287/LER-1981-009-03, /03L-0:on 810507,containment Isolation Valve 3CS-6 Was Discovered in Intermediate Position,Resulting in Loss of Containment Integrity.Cause Undetermined.Valve 3CS-6 Verified Operable
References
NUDOCS 8107200285
Download: ML20009C099 (3)

LER-1981-009, Forwards LER 81-009/03L-0.Detailed Event Analysis Encl

Event date:
Report date:
2871981009R00 - NRC Website

text

~

8 5>

> :E'i.t c Ol*FICIAg-DUKE POWER COMPANY M.w,,

Of>y Powra acimaxo 4cc Sourn Cucacu SuazzT, CHARWTTh.,N;j, pd C

.sa AJ U.* $"

w l L LI AM O. PA R M E R, J R.

June 5, 1981 1,,,,,,,..,,,,70 v e, p...m.

373-4083 Se,au Paoouc'ione g/- 0.2/- OJO Mr. James P. O'Reilly, Director s

/f U. S. Nuclear Regulatory Commission Region II 101 Marietta Street, Suite 3100 Atlanta, Georgia 30303 2' ($

ggg e '3 6

sM Re: Oconee Nuclear Station UAMt,c #

q Docket No. 50-287

Dear Mr. O'Reilly:

Please find attached Reportable Occurrence Report RO-287/81-09. This report is submitted pursuant to Oconee Nuclear Station Technical Spec-ification 6.6.2.1.b(2), which concerns operation in a degraded mode permitted by a limiting condition foe operation, and describes an incident which is considered to be of no significance with respect to its effect on the health and safety of the public.

Very truly yours, bb

.M w

7 William O. Parker, Jr.

JLJ/djs Attachme:nt Mr. Bill Lavallee cc: Director Office of Management & Program Analysis Nuclear Safety Analysis Center U. S. Nuclear Regulet.ory Commission P. O. Box 10412 Washington, D. C. 20535 Palo Alto, California 94303 4I l I

8107200285 810605 PDR ADOCK 05000287

e DUKE POWER COMPANY 3

Oconee Unit 3 Report Number: R0-287/81-09 Report Date: June 5, 1981 Occurrence Date: May 7, 1981 1

Facility: Oconee Unit 3, Senera, South Carolina Identification of Occurrence: Loss of Containment Integrity Resulting In j

Unit Shutdown Conditions Prior to Occurrence: 100% Full Power i

Description of Occurrence: At 0832 hears on May 7, 1981, containment l

isolation valve 3CS-6 was discovered, during a control room tour, to be in i

the intermediate position. An attempt was made to close the valve, however the valve would not close. A review of computer printouts indicated the l

valve opened at 0253 hours0.00293 days <br />0.0703 hours <br />4.183201e-4 weeks <br />9.62665e-5 months <br /> on May 7, 1981. Since 305-5, the redundant isolation valve, was already inoperable, the failure of 3CS-6 resulted in loss of containment integrity. Reactor shutdown commenced. The unit was ohutdown within the time permitted by Technical Specification 3.6.3.

i During the shutdown 3CS-5 was repaired. However, when craft personnel attempted to stroke 3CS-6, the valve operated correctly. An investigation as to the cause of 3CS-6 failure was unable to ascertain the reason for 3CS-6

)

failing to the intermediate position at 0253 on May 7, 1981. Since 3CS-5 had been repaired and 3CS-6 was determined to be functionally operable the 1

unit was restarted and operated until 1445 hours0.0167 days <br />0.401 hours <br />0.00239 weeks <br />5.498225e-4 months <br /> on May 19, 1981 when 3CS-6 was again found to be open. A review of computer printouts revealed the valve opened at 1105 hours0.0128 days <br />0.307 hours <br />0.00183 weeks <br />4.204525e-4 months <br /> en May 19, 1981. Further investigation revealed there was leakage through 3CS-5 which resulted in loss of containment integrity. Reactor. shutdown was again commenced and the unit was shutdown within the time permitted by Technical Specification 3.6.3.

Both of these events constitute operation in a degraded mode permitted by Technical Specification 3.6.3 and are thus reportable pursuant to Technical Spec-ification 6.6.2.1 b(2).

Apparent Cause of Occurrence: The apparent cause of the failure of 3CS-6 which occurred at 0253 hours0.00293 days <br />0.0703 hours <br />4.183201e-4 weeks <br />9.62665e-5 months <br /> on May 7, 1981 was not determined at that time.

a However, following the failure which occurred on May 19, 1981, it was deter-mined the cause was two fold. First, valves 3CF-22 and 3CF-23 (core flood I

tank manual drain valves) were leaking past the seat. These valves drain to

[

the quench tank drain line which is isolated by 3CS-5 and 3CS-6.

Second, valve 3CS-124 which is a relief valve upstream of 3CS-5 and 3CS-6 was " gagged" i

l closed. The combination of these two occurrences resulted in pressurization of the quench tank drain line to approximately 450 psig. Valve 3CS-6 is pneumatically operated and requires air pressure to opsn but closes on spring pressure. Approximately 175 psig system pressure is required to over-come the spring within the operator. An extensive document search was l

unable to determine why or when 3CS-124 was " gagged", however it is believed l

it was gagged during construction and/or original system hydro testing.

I r

I w

sr-r+-,-.

.,14 e w n wa - -

- ewm.,,~,--n,-n--,.,-e-ww.as--e-,,,e,-me-,,,-

n.--,,-m,-e-w--,,,,a,+

--envn n,.-e

,v--.

,~-----n---mw.,--.

.w-mv.--

0 O

O e

Analysis or Occurrenec: Both valves 3CS-5 and 3CS-6 are containment isolation valves whose safety function is to isolate against a 59 psig post loce con-tainment atmosphere. Relief valve 3CS-124 its designed to prevent overpressuriza-tion of the quench tcnk drain line by higher pressure sources which draf't into it.

The simultanious thru leakage of valves 3CF-22 and 3CF-23 combined with the " gagging" of 3CS-124 resulted in pressures which forced flow thru 3CS-5 and forced 3CS-6 to open. However, had a loca occurred and the core flood tanks discharged, the pressure would have reduced and, 3CS-5 would have sealed against the post loca containvient pressure of 59 psig and 3CS-6 would have closed also sealing against the post loca atmospheric pressure. Additionally on both events described in-this-repor.t the-unit.was.

shutdown within the time frame specified by the Technical Specifications.

Thus, this incident is considered to be of no significance with respect to safe operation, and the health and safety of the public were not affected.

Corrective Action

For the event of May 7, 1981 valve 3CS-5 was repaired and tested and valve 3CS-6 was determined to be functionally operable. For the event of May 19, 1981 the following actions were'taken:

1.

A modification was performed to temporarily remove a.section._ __.

of the core flood tank drain lines and pipe caps were installed on both open ends of the piping.

2.

3CS-124 was "ungagged", removed from the line, tested, reset and reinstalled.

3.

An inspection and evaluation of the section of piping which was overpressurized determined that the piping and valves were still f'inctional.

4.

The identical relief valves on Units 1 and 2 ( 1CS-124, and 2CS-124 ) will be inspected at the next opportunity.

• MWe'*

m e".nein

=

,.eg,a ge W

u.=

= ->

^ * - - * - -

..em e.ew. e =,

p.

hMe t-