ML20043C099

LER 90-014-00:on 900427,investigation Revealed That Component Cooling Water Piping to Reactor Coolant Pump Seal Coolers Could Be Targets of High Energy Line Break.Safety Analysis for Operability completed.W/900529 Ltr

ML20043C099
Person / Time
Site:	Fort Calhoun
Issue date:	05/29/1990
From:	Gates W, Lehman L OMAHA PUBLIC POWER DISTRICT
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
REF-GTECI-A-02, REF-GTECI-RV, TASK-A-02, TASK-A-2, TASK-OR LER-90-014, LER-90-14, LIC-90-0399, LIC-90-399, NUDOCS 9006040038
Download: ML20043C099 (5)

Text

r--

g d

s.

. . Lt .

Omaha Public Power District I

444 South 16th Street Mall Omaha, Nebraska 68102-2247 402/636-2000

( May 29, 1990 LIC-90-0399 i

.I U. S. Nuclear Regulatory Commission Attn Document Control Desk Mail Station P1-137-

,j Washington, DC 20555

Reference:

Docket No. 50-285  !

! Gentlemen -

Subject:

Licensee Event Report 90-14 for the Fort Calhoun Station.

Please find attached Licensee Event Report 90-14 dated May 29, 1990.

. This report is being submitted pursuant to requirements of 10 CFR -i

i. 50.73(a)(2)(ii)(B). '

i If you should have any questions, please contact me.

Sincerely, 2

W. G. Gates

! Division Manager i

Nuclear Operations

( i l WGG/ tem Attachment c: R. D. Martin, NRC Regional Administrator A. Bournia, NRC Project Manager a P. H. Harrell, NRC Senior Resident Inspector '

INP0 Records Center i American Nuclear Insurers -

?

I a

9006040038 900529

, PDR ADOCK 05000283 ,

t i S PDC .

15-St24 Employmen th Equal opportunity l l

.- - .. . . . . . - . - . - - - - - . - -- - ---~ - - -- - - --- - -

4 .

" 38' U S. NUCL612 $.EEUbaTOAV COnesi&460N C,fD APrMOVED OMB NO 3160410e e ExPints 440/97

- " *^ " ** ' **^

LICENSEE EVENT REPORT (LER) '!**, Wo' M ^" m' i n N '. M"' ""C ^'

'!M".':5E"f!

%!",,'i"'.Ji"m'ao*N '

Eis'd#0A Ofici '

OF MANAGEMENT ANO suDGET. WASHINGTON.DC 20503.

8 AC4Lif y seatet 11t 'DOCEST NURISLR(21 PAGEi3i Fort Calhoun Station Unit No. 1 o i s I o I o I o i 21815 1 lorl 014 181L4 tes comnonent Coolina Water _ Containment Isolation Valves Outside Design Basis IVENT DATS (St Lgm Ngugga (gg mgpony paTt (7) oTHgh F ACiUTit$ INVOLVED it)

MONTH DAY vtam vtAn " M U 'n @w"M MONTH DAY vtAR ' AC' Liv v m Auss pocaat NvMesmisi i

N o is to lo g o; i ;

~ ~

Ol4 2l 7 90 9l0 0l1 l 4 0l0 0l5 2 l9 9 l 0, o isto lo t oi l l-TMiG R$PQRT 83 $Uge44TT80 PunsuANT TO THE RhOuisitutNTS 0810 CFR 5 ,Ca.c4 eae .e *ere e r tae rettowiat, titi OPE R ATite,O 88005 M 5 20 a0*ini 0 0.i.i .0.T3 H H i 73 tiini Powsn to 405teH1H4 90.384all16 90.73teH2Het 73.71tal

~ ~ - ~

'IEOP 01010 =0 =NH$ H,H umNim u nNH Hai ovai'i ,s < f, ,a A* e .<r d

20 408taH1HHH 90.734aH2Hd 90.73tsH2Hv.dHA) JW6AJ N H, H.,

1 .0 n NH.,~,

.0 nl.H H .H.,

30 406 tait 1 Hut E734eH2Huu E73(aH2Hal LICENott CONT ACT FOR TMit Lim (121 N A.M t TELEPMQNg NUMSER AntACoot

{

Larry Lehman, Shift Technical Advisor 410 12 51313 l- t 61812 I 9 ,

Cowan oNi uni roa iACn cowoNENT , Aituai oiaCaisio N Twie alPoaf tm CAU s sv tim CowoN NT agga,TAggi c,v,, ,y ,n u co ,oN,,T

}

• ^3l AC. MANg;AC- n47aTAgg.

I I I I I I I I i I i i i I I i l I I i l I I I I I I I SUPPLEMENTAL REPORT EMPGCTtO t14) MONTM OAv VGAR Sv8 Mis $ ion T] ve ,,, . . ^""

surcreo sv.wssioN oA re, 7 No 018 1I5 910 A T . AC T ,L .-.. , . . , . . . . ,, .. . -, ,.a , n .,

OmahaPublicPowerDistrict(0 PPD)initiatedaninvestigationoftheComponent CoolingWater(CCW)linespenetratingthecontainmentwallinresponse.to concerns recognized in NRC Information Notice 89-055. The results of this investigation revealed that the CCW piping to the Reactor Coolant Pump seal coolers could be targets of a High Energy Line Break-(HELB) on the Reactor Coolant System hot leg. A HELB inside containment could negate one containment isolation barrier, leaving only the outboard containment isolation valve to mitigate potential radiological releases and thus susce)tible to single failure. This is a condition outside the plant design ) asis.

The problem was reported to the NRC pursuant to 10 CFR 50.72(b)(2)(i) at 1627 on April 27, 1990.

Short term corrective action was completion of a safety Analysis for Operability to justify operation. Resolution of Unresolved Safety Issue A-2 responses is expected to allow the CCW valves and piping to be exempted from HELB targeting. j

/

s

/

NRC Perm att (649) e

/

/_ . ..

7 FCpassesA u 8. KuCL(2.R kEOULAThY hitBION EMPIRES 4/30/92 LICENSEE EVENT REPORT (LER) C ISat'Mio'#8'UcWo"'!?o*uT' .0 f'lA' ERw'A %

TEXT CONTINUATION M"4"4'#i o'2%iNM"RIUe'#^,'Z"'"MRA PAPER O Rt U T N J (3 60 04 O IC QP MANAGEMENT AND OUDGET. WASHINGTON. DC 20603, P AC8kl(7 8sA884 01 00CKti nut 00ER (2) (gg gygggR f61 PA00 (31 n*= " % W ." .

?!:::

Fort Calhoun Station Unit No. 1 o ls lo j o j o l 218 l 5 9 l0 -

0l1l4 -

0l0 0 12 oF 0 l4 ftKT g mee asses a assessest, ese ainssener Mtc pwm was (17)  !

Fort Calhoun Station Reactor Coolant Pumps utilize Com>onent Cooling Water (CCW) as a heat sink for cooling the pump seals and lu)ricating oil. The'CCW piping is neither ) art of the Reactor Coolant Pressure Boundary nor does it communicate with tie containment atmosphere; therefore, it is classified as a -

closed system. The CCW piping inside and outside containment was designed to '

the requirements of ANSI B31.7, Class I Seismic.

The CCW lines to and from the Reactor Coolant Pumps penetrate the' containment wall. The two CCW lines are isolated by containment isolation valves, HCV-438A-and HCV-438C on the inboard side and'HCV-4388 and-HCV-438D on the outboard. '

side. Each of these valves fail in the open position to prevent a loss of l cooling to the Reactor Coolant Pump seals should valve failure occur. The outboard valves (HCV-438B and D) each have a 1000 hour0.0116 days <br />0.278 hours <br />0.00165 weeks <br />3.805e-4 months <br /> nitrogen' backup to '

maintain closure of-the valves during a LOCA with a lossiof: Instrument Air.

The closed system classification defined in 10 CFR 50 Appendix A, Criterion 57, allows the inboard containment isolation valves to be replaced by closed barrier piping, provided that it is protected from Loss of-Coolant Accident-(LOCA)conditionsanditsdynamiceffects.

l Omaha Public Power District (OPPD)' initiated an investigation of the CCW lines-penetrating the containment wall in response to NRC Information Notice 89-55.

I Design Engineering personnel walked down the CCW linesLduring the'1990 refueling outage to determine'if any of the piping might be=affected-by.a LOCA. The results of this investigation revealed that the CCW piping.to the Reactor Coolant Pump seal coolers could be targets of the dynamic effects-(missiles, whip, jets)ofaHighEnergyLineBreak(HELB)oftheReactor Coolant System hot leg. The HELB scenario assumes that targeting will' affect allcomponentswithin10pipediametersoftheRCShotleg(320' inches). The CCW lines to the Reactor Coolant Pump seals run within 48 inches.of-the RCS hot leg. A HELB inside containment causing-the failure of.CCW piping would negate one containment isolation barrier and leave the system with only3the outboard I containment isolation valve to mitigate potential radiological releases.

Since no documentation exists indicating that the lines.have been designed to withstand a HELB, failure of the CCW lines is assumed to occur during a LOCA/HELB due to pipe whip. The CCW lines at this point would no longer be

considered part of a closed system. Hence, the containment isolation valves .

I must comply with 10 CFR 50 Appendix A, Criterion 56. This means that both inside valves HCV-4388 and D must fail in the closed position or be maintained >

l in the closed position during a LOCA to provide containment isolationiand l- prevent containment bypass due to a single failure. Because these valves do l

not presently meet Criterion 56, they are considered to be outside the design' basis for containment isolation valves in open systems. Investigation of the effects of a LOCA on other closed system piping in containment is ongoing.

f feAC Feren 306A (649)

~

peau assa u.s. s:veta An ia rutarony cosasesioN -

' Ept7.18 4/30/92 LICENSEE EVENT REPORT (LER) f,%"t'#o',fi8[*,eM'M o*j'JWT' ,*o'?.'J!

' TEXT CONTINUATION . iu""A*4'.'l1"^23?nN!"U/c'."^.'!# L"t MRA O!?#?,"TJt"n=to"~^30!aWom'10??? l op uANAosusNT ANo suoott. WASHINGTON, DC 20603.

_ l

. Aciutv NAast m oocnet Numeen m Lan numeen sei PAot tai vtAR .

N[,"4  : . $,eb m Fort Calhoun Station Unit No. 1 o l510 l0 l0 l 21815 9 l0 -

0l 1l 4 -

0l0 0 l3 - 0F 0 14 i text a- . w. u w =c i 3su wim

. I FortCalhounStationwasinRefuelingShutdown(Mode 5)whentheCCWvalves were determined to be outside the design basis at~1445 hours on A)ril.27, 1990. The problem was re pursuant to 10 CFR 50.72(ported as " degraded b)(2)(1)at1627onApril 27,while 1990. shutdown" This LER toistie NRC submitted'pursuantto10CFR~50.73(a)(2)(ii)(B).

The potential impact on safety is a possible breach of containmen't as a result of a LOCA concurrent with a loss of DC Bus 2 or Instrument Air (IA). The-radiological consequences of such a potential breach could be in excess of 10' ,

CFR 100 guidelines. For this reason-a probabilistic argument was completed to ,

determine the probability of' containment-bypass via the CCW lines as a result:

of the LOCA dynamic impact.' The event considered as models the failure of IA, DC Buses, backup nitrogen supply to HCV-438B and D, solenoid valves, regulators,-autoclose signals, manual closure by control switches,'and-local manual closure of HCV-438B and D. The probabilistic argument is specific for this scenario. It is not indicative of core damage or radiological release .

frequency.

The conclusion of the analysis is a 1.51E-8 per year event probability of containment bypass via the CCW. lines initiated by LOCA' This is only the .

probability that. containment bypass will occur and does not include conditions 4 that would cause severe core damage and fuel failure. -A severe accident assumes gross failure of fuel. .The screening criterion for Generic Letter-88-20 states that a probability of IE-8 per year is acceptable for severe accidents. For the breach of'CCW line event-, the probability of additional failures that cause containment bypass to be classified as a severe accident-would reduce the 1.51E-8 )er year probability to around 1E-11 per year, a- J significant margin from tie Generic Letter screening value.

Therefore, the issue of a potential bypass in.the'CCW system concurrent.with' LOCA and failures as modeled in the probabilistic based argument was found to be of a sufficiently low frequency such that it is not considered'a.significant concern. SafetyAnalysisforOperability(SA0)90-08,.approvedonMay12, 1990, addresses the safety. significance of this problem and provides-justification for the operability of the CCW piping and valves throughout the upcoming fuel cycle.

NIC Form 30BA (6491

~ _ - - - . - . _.

,,,, , p A unvCt Amuvutoar coM

,,,,,y,,,,,,,,,,,,,,

• EXPrAES 4/30/92

](

LICENSEE EVENT REPORT (LER) '#Et','Ro',f28'"aM",cl;'&o**il,' .#'l'l'n'* ,"o'."J"'! )

. TEXT CONTINUATION 1,""n'*48,"iT M2d n'0?!Mf/c'"*4'# '".' O'EUA j 1 PAPE OK & TION C l'3 60 04 i ' 'l OF MANAGEM(NT ANO SUDGET, WASHINGTON. DC 20503.

l F ACILITV feAMt nl DOCKli VURSER (21 LE A 8tVMBER (Si PA06131'

" a'  :

uwaYa Fort Calhoun Station Unit No.1 0 l5 t o l o l e 12 l 815- 9l0 -

0l1l4 -

0l0 0l4 oF 0l4 I rm ,u - e m ~ me - m., ,nm As corrective action, OPPD plans to finalize resoltition of Unresolved Safety.

si _ . Issue (USI) A-2 by utilizing Leak Betore Break (LBB) methodology as per Generic' i i' Letter 84-04.- .This resolution is expected to eliminate the requirement for the-RCS HELB postulation and the resultant dynamic effects. Acceptance by the NRC of this resolution will allow the CCW valves and piping to be exempted,from consideration as a HELB target...When this occurs, the plant desjgn basis:in- ,

the USAR will be appropriately revised in'the 1991 annual-update. ('

.J Additional review.into the reason for this condition will be. conducted. 'A Root.

Cause Analysis will be performed and the results will be submitted in a- '

supplement to this LER. '

Other Licensee Event Reports which have been submitted addressing design deficiencies are LER's90-003, 90-005,90-007,- 90-009,- 89-007,89-009,: 89-014,89-015, 89-024,88-009, 88-019,88-020, 88-032,88-033, 87-018, and 79-021.

l .t i

l

}

e

.f l

l l

l

-l NRC Fenn ag8A (640) .

I