05000285/LER-1989-002, Forwards Corrected LER 89-002 & LER 89-003

Forwards Corrected LER 89-002 & LER 89-003

ML20235U174
Person / Time
Site:	Fort Calhoun
Issue date:	02/28/1989
From:	Morris K OMAHA PUBLIC POWER DISTRICT
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20235U177	List: 05000285/LER-1989-002, Forwards Corrected LER 89-002 & LER 89-003 05000285/LER-1989-003, Forwards LER 89-003
References
LIC-89-239, NUDOCS 8903090084
Download: ML20235U174 (1)

LER-2089-002, Forwards Corrected LER 89-002 & LER 89-003

Event date:
Report date:
2852089002R00 - NRC Website

text

- _ _ _ _ _ _ - _ _ _ _ _

. = , , ' ';,(;P+ ,,.1-

,{ ~ , . . .. ' ]; '

, < g.

b, . - ' ' '

I, .

. s y ?.,$I:T;l'

.f 4.[ M U . i 3 , . ..$ y Cl 1.  : . : . ".' -  ; .- . .. . . , _

Omaha Public Power District 1623 Harney Omaha. Nebraska 68102-2247 402/536-4000 February 28, 1989 LIC-89-239 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Mail Station P1-137 Washington, DC 20555

References:

1. Docket No. 50-285

2. Letter OPPD (K. J. Morris) to NRC (Document Control Desk) dated February 6, 1989, Licensee Event Report 89-002 for the Fort Calhoun Station

3. Letter OPPD (K. J. Morris) to NRC (Document Control Desk) dated February 9, 1989, Licensee Event Report 89-003 for the Fort Calhoun Station Gentlemen:

SUBJECT:

Licensee Event Report 89-002 and Licensee Event Report 89-003 for the Fort Calhoun Station Please find attached Licensee Event Report 89-002 entitled " Surveillance Test ST-RHRS-1 and ST-RHRS-2" and Licensee Event Report 89-003 entitled

" Missed Operability Verification for a Fire Damper" dated February 6, 1989 and February 9, 1989, respectively. Licensee Event Report 89-003, as submitted in reference 3 was inadvertently numbered LER-89-002 on the NRC Form 366. Accordingly, we have attached a correct version of each of these LER's to this letter. The reports did not change in content. If you

! should have any questions concerning this matter, please do not hesitate to contact us.

Sincerely, g .KJ. Morri

. $A l

// Division Manager Nuclear Operations f

KJM/dm &

Attachments

\\

c: R. D. Martin, NRC Regional Administrator P. D. Milano, NRC Project Manager P. H. Harrell, NRC Senior Resident Inspector

INPO Records Center i

American Nuclear Insurers seasotoop po;;___

PDR ADOCK U500uno E. FDC

<s se4 Employmen n ual opportunity

./

Omaha Public Power District 1623 Harney Omaha. Nebraska 68'02 2247 402/536 4000-February 6, 1989 LIC-89-137 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Mail Station P1-137 Washington, DC 20535

Reference:

Docket No. 50-285 Gentlemen:

SUBJECT:

Licensee Event Report 89-002 for the Fort Calhoun Station Please find attached Licensee Event Report 89-002 dated February 6, 1989.

. This report is being submitted per requirements of 10 cro 50.73. _

Sincerely,

& b K. J. Morris Division Manager Nuclear Operations KJM/dm Attschment c: R. D. Martin, NRC Regional Administrator P. D. Milano, NRC Project Manager P. H. Harrell, NRC Senior Resident Inspector INP0 Records Center American Nuclear Insurers s

-.4

_ < - _ = ,e -

NR ee,= ase u s. NUCLEAR REoVLATORv COsassessOse AF# ROVED OMS NO. 318S410s LICENSEE EVENT REPORT (LER) '*"'*:8"

F ACILi'(Y NAME 41)

DOCKET NUMetR (34 taGE th Fort Calhoun Station Unit No. 1 TITLE tJ o Is I o Io Io! 218 5 1 lOFl 013 Surveillance __ Test ST-RHRS-1 and ST-RHRS-2 EYE 37 DATE(61 l LER NUMSER tel REPORT DATE 17)

OTHER F ACILITIES INVOLVED let MONtM OAv vtAR . vtAR "R* ",;',' '

, gy,g MONTH day vtAR e nciuvv =a=u OOCxET NUM.ERtsi H

o isio g o,o g i ;

0l7 89

~ -

Ol1 8l 9 0l0l 2 0 l0 0l2 0! 6 8l 9 o isio io , o , i ,

TMi8 REPORT to sUSMtTTED PURSUANT TO THE REQUIREMENTS OF 10 CFR th tCheca one er me,e of f** _for. nli 5 a== a .i amaa tus=

==4.in in, a at.ini .0.ni.H:H.i run.i no, 0, 010 ==wa Hai ami Han amuH*

p;s7,7,,a;y, ;;',,,

20AGSteH1Hilll g 50.73teH21HI , 80.73(all2HvHiHA)

J6dAJ 20A0BisH1Hiv) 60.73'eH2Hul 80.731sH2Hel#HS3 20A054eH1Het 50.73teH2Huis to.73teH3Hal LICENSEE CONTACT FOR TMit LER nti NAME TE LEPHONE NUMSER

• REA COQE noon Mo17er. Shift Technical Advisor 402 1 1 5 i 3,3, i 66 i i O0 i COMPLETE ONE LINE FOR EACM COMPONENT FAILURE DESCRIBED IN TM,8 REPORT n38 CAUSL SYSTEM COMPONENT MA C' " E

,TpA CAUS MA C. R SYSTEM COMPONENT RTA LE pq ,

1 1 1 1 1 1 I 3 pag

. l I I I I I f - M' W i  ; ; ; i l l

$ SUPPLEMENTAL REPORT EXPECTED ttei i  ; i ; I I bWi$E'- t MONTM CAV YEAR Yt$ til ven. comoee fMPfCTED suchttsst0N DA TEl NO AouACT,u ...a - . -

..,.e~,.n., l l l Review of Surveillance Tests ST-RHRS-1 and ST-RHRS-2 by System Engineering indicated that these procedures as previously conducted might not have met the intent of Technical Specification 3.16(1). On Januar that these tests, as conducted during previous years,y did7,1989, not meet it the wasTechnical determined Specification requirements and were consequently reportable.

Regarding ST-RHRS-1, investigation has determined that an oversight was made when the surveillance test was revised in 1982. The surveillance test as it was written before the 1982 change, provided for the verification of leakage by pressurizing through a test connection upstream of the check valves. However, the revision changed the location at which the system is hydrostatically pressurized from upstream to downstream of the check valves, thus isolating pressure from the piping in question. To correct the problem, the procedure was changed by the System Engineer to ensure that deficiencies identified during the review were satisfactorily addressed. The revised procedure was successfully completed during the 1988 refueling outage.

Regarding ST-RHRS-2, the isolation valve interlocks were installed in l'175 as part of a modification to protect the Shutdown Cooling (SDC) system iir overpressurization. However, the surveillance test was not changed to reflect the addition of the interlock logic and its operating constraints. To correct the problem and ensure the requirements of the Technical Specifications are satisfied, the procedure was rewritten by the System Engineer. The revised procedure was successfully completed during the 1988 refueling outage.

s UCENSEE EVENT REPORT (LER) TEXT CONTINUATION 4 .aovto ous to me oio.

EXPIRES 9/31/es

,4CsLif Y haast til Docuti humetR 421 LER huustR tai paf,g 43)

n. " %'?,V " 1';* :

FORT CALHOUN STATION UNIT NO. 1 o is jo jo jo ]2 [ 8l5 8l 9 -

0l0j2 -

010 0l 2 0F 0 [3 TEXT i# snore asses a seewed ese .

_f haC # eve Assrali171 Review of Surveillance Tests ST-RHRS-1 and ST-RHRS-2 by System Engineering indicated that these procedures as previously conducted might not have met the intent of Technical Specification 3.16(1). On January 7, 1989, it was determined that these tests, as conducted during previous years, did not meet the Technical Specification requirements and were consequently reportable.

Review of ST-RHRS-1, " Recirculation Heat Removal Hydrostatic Test" revealed that the test did not satisfy the requirements of Technical Specification 3.16(1)b. The Technical Specification requires that piping from valves HCV-383-3 and HCV-383-4 to the discharge isolation valves of the safety injection pumps and containment spray pumps be hydrostatically tested at no less than 100 psig. Due to the way the test was written, the sections of piping from check valve SI-159 to HCV-383-3 and from check valve SI-160 to HCV-383-4, were isolated from system pressure by the check valves, and therefore cannot be verified to have been hydrostatically tested to 100 psig in previous tests.

An oversight was made when the surveillance test was revised in 1982. The surveillance test as it was written before the 1982 change provided for the verification of leakage by pressurizing through a test connection upstream of the check valves. However, the revision changed the location at which the system is hydrostatically pressurized from upstream to downstream of the check valves, thus isolating pressure from the piping in question.

To correct the problem, the procedure was changed by the System Engineer to ensure that deficiencies identified during the review were satisfactorily addressed. The revised procedure was successfully completed during the 1988 refueling outage. '

Technical Specification 3.16(1)a requires that the portion of the shutdown cooling (SDC) system that is outside containment be tested at 250 psig each refueling outage. ST-RHRS-2, " Recirculation Heat Removal Leak Rate," is the procedure used to satisfy this requirement. Due to the way in which the surveillance test was performed, it is unlikely that the SDC system was pressurized at 250 psig when the surveillance test was conducted in previous years.

The surveillance test was designed to utilize normal SDC system operation to pressurize SDC piping. An initial condition for the performance of the surveillance test is the requirement that Operating Instruction 01-SC-1,

" Initiation of Shutdown Cooling" be completed. Initial conditions for the initiation of SDC is for RCS pressure to be below 250 psia (235 psig) and 300 degrees F. The interlock logic on SDC primary isolation valves HCV-347 and HCV-348 prevents the valves from opening or initiates a close signal when RCS pressure is equal to or greater than 235 psig. This setpoint is checked by Calibration Procedure CP-118 which provides the setpoint tolerance as 235 (+0,

-16) psig. Based on this setpoint band the SDC interlocks would prevent SDC from being initiated when RCS pressure is equal to or greater than 250 psia (235 psig). Based on this information, it appears that the SDC .uction header could not have been pressurized to 250 psig during previous tests.

.g.o o..

. o s o,o i ...un.e ..

=

meC ,eem Amea

" ' ' U e. NUCLE An meGULATOaY COnelseessoas UCENSEE EVENT REPORT (LER) TEXT CONTINUATION amoveo o o sino oio.

enem em,,e f actLair masse gis oocati muassen us tan muassen con paes est vtan sepsman's L

.Ev k

FORT CALH0'UN STATION UNIT N0. 1 rar- - . m e , , m,,m 0 l510 lo lo !2 l 8!5 8l9 -

0l0 l2 -

O!0 0l3 or 0l3_

The isolation valve interlocks were installed in 1975 as part of a modification to protect the SDC system from overpressurization. However, the surveillance test was not changed to reflect the addition of the interlock logic and its

! operating constraints.

l To correct the problem and ensure the requirements of the Technical Specifications are satisfied, the procedure was rewritten by the System Engineer. The revised procedure was successfully completed during the 1988 refueling outage.

l 4

FCAM Me

_ _ - __ ____ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _