ML19332D492

Responds to Generic Ltr 89-21 Re Status of Implementation of USI Requirements.Requested Info Re Util Fort Calhoun Station & Supplemental Info Re Four Specific USIs Encl

ML19332D492
Person / Time
Site:	Fort Calhoun
Issue date:	11/28/1989
From:	Morris K OMAHA PUBLIC POWER DISTRICT
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
REF-GTECI-MI, REF-GTECI-SC, TASK-***, TASK-OR GL-89-21, LIC-89-1095, NUDOCS 8912040041
Download: ML19332D492 (7)

Text

_ _ . _ _ _ . . _ _ _ . . _ _ - . - _ . _ _ ___

, . ~.

3.- q

l. .

Omaha Public Power District 1623 Harney Omaha, Nebraska 68102 2247  !

402/536 4000  :

i November 28, 1989 LIC-89-1095 i i

U. S. Nuclear Regulatory Commission r Attn: Document Control Desk Mail Station PI-137 '

t' Washington, D2, 20555 l

References:

1. Docket No. 50 285

2. Letter from NRC (J. G. Partlow) dated October 19, 1989 I (Generic Letter 89 21) ,

Gentlemen:

SUBJECT:

Response to Request for Information Concerning Status of #

Implementation of Unresolved Safety Issue (USI) Requirements Reference 2 requested addressees to provide information on the implementation status of USIs for which the NRC has achieved a final technical resolution.

-The requested information with regard to Omaha Public Power District's (OPPD)  ;

Fort Calhoun Station is provided in Attachment 1. In addition, supplemental -

information is provided in Attachment 2 regarding four specific USIs that have l been discussed with OPPD's NRC Project Manager. If you have any questions,

~

please contact me, i Sincerely, j K. . Morris  ;

Division Manager l Nuclear Operations  !

l KJM/pje c: LeBoeuf, Lamb, Leiby & MacRae R. D. Martin, NRC Regional Administrator A. Bournia, NRC Project Manager P. H. Harrell, NRC Senior Resident Inspector Md20ggg[k$ , g5 1 w

\

as sm imnioumengeguy onnonunnu

,. o

,-4:

3ttacilmet 1 ,.

Pbet mlhrust Statist Status for Ikitesolved Safety Immans with a Final Tetfirdnel paurentitut

• IJSI Ntunber Title Status /Date* Dea 74ts A-1 Water N NC ' 3Nur IMEEHB93, Revision 1, necommary actions aN by NLEEG-0737 requi. _ 16.

A-2 Asy=.-cir.Ic Blowdown E 03/15/90 See Attaciument 2 toads on nasctor Primary Or'. ant Systems A-3 W _m= Steen Gaa.zaWx TLibe Integrity E Nestint$ouse HEs only A-4 CE Stessa Generator Tibe C 06/18/85 Desolved by OPPD response to Generic Integrity Letter 85-02 (Referetum OPPD letter LIC-85-259 dated 6/18/85)

A-5 B&W Steam Generator R BtM BEs only Tube Integrity A-6 Mark I Contairusent R Mark I-EBEs only Short-Tern Fi wa.ama A-7 Mark I IIng-Teria R Mark I-IBEs only Frwtama A-8 Mark II Containment R Mark II IBes only Pool Dynastic innrk i

• C - Complete '

NC - No Changes Necessary NA - Not Applicable I - Incanplete E - Evaluating Actions Required

Burt Calhoun Station Statmas for TRresolved Safety T-- with a Final Sedwnirm1 metab*itut ,

USI Number Title Status /Date* P =mits

• A-9 Artir ipated Transients C 08/11/89 Besolved by 1EIC Safety Evaluation dated Without Scram 08/11/89. (OPPD reeptmded to SER confi ^wsy iten via IZC-89-901 dated 08/22/89).

A-10 bet Feedwater Nor,zle NA Nets only CrackimJ A-11 Reactor Vessel Material C 04/25/84 See Attactment 2 Totaja ====

A-12 Fracture Tot.jas of NA DEts with u noction permits Steam Ge-u.atur and issued after octr*=r 1983 only.

Baac+nt- Coolant hup

^

Swtw A-17 Systees Interactions E 03/08/90 See Attadament 2 A-24 Qualification of Class C 02/15/85 Resolved by IEC Safety Evaluation dated IE Safety-Related 02/15/85. (164C letter dated 01/17/86 Egli tuarit resolved follcnMp confinnatory items).

A-26 Reactor Vessel Pressure I 12/31/90 See Attachment 2 Transient Prd*4cn A-31 Residnal Heat Renoval NA Operating Licenses iemanad after 01/01/79 shutdown Requim - s only.

A-36 Otmtrol of Heavy Taark C 12/30/85 Resolved by MRC letter dated 12/30/85.

Near Spent Rael (Prior resolution of Ihase I drv,==rited by NRC Safety Evaluatist dated 05/22/84)

• C - Ocmplete NC - No Changes Nmwry NA - Not Applicable I - Inoceplete E - Evaluating Mdans Required

o Port on1houn Station Status for Ikitesolved Safety Ts===== witin a 1*inal Tactademi immenh*4,=3 ,

USI Number Title Status /Date* psenrts A-39 Detenmination of SPX MA B Es only Pool Dynastic innris arrl Pressure Transients A-40 Seisunic Design MC Per NWEG-1233, requirements of IEI A-46 Criteria otwer IEI A-40; no separate action is

! required. (See IEI A-46) .

A-42 Pipe Cracks in Boiling NA B Gs only j water neactors A-43 Oontainument E-r_.p.cy NC Per Generic Istter 85-22 dated 12/3/85, Sump Performance no requin._A. for licensed plants j A-44 Station Blackout I 12/31/89 0:x1dtacting test to validate masangtimis in Station Blachaut Ooping Assessment (Reference OPPD letter LIC-89-572 dated

6/15/89) l A-45 Shutdown Decay Heat I 12/01/93 To be addressed by Indivirhal Plant nanoval aequi-is Examination required by Generic Istter 88-20 dated 11/29/89 (Reference OPPD ,

letter UC-89-930 dated 10/31/89)  !

A-46 Saimnic QulifiWen of E ** ** Actions and actiarble 4 L:24.on NEC i Equipment in Operating review of Seisamic Omlification Utility Plants Group (SQUG) Generic Imple.-:=iustion i Procedure (Reference OPPD letter MC-89-661 dated 7/20/89)

• C - Ccanplete '

NC - No Onnges N-mry -

NA - Not Applicable I - Inccarplete E - Evaluating Wdons Required

. . _ .- . . _ . _. __ _ , , . . . _ _ - _ ~ . . _ _,

A-- ,'

~

4 . --

Ptzt Calhourt Stat.icn Stattus for IR1 resolved Safety Ta== with a Final TInthrdcal muureattrun ,

i l 'IEI Number Title Status /pate* Maarts A-47 Safety Implication of E 03/19/90 Plarmed response to Generic Tm8+mr 89-19 Control Systems

~

dated 9/20/89 vill atMress IEI A-47

, A-48 Hya vy== Control Measures NA 14Rs with ice condener

, and Effects of Hyavy=== cantairuumts and Bets only Burns on Safety Fey 2i=arrt

{

A-49 Pra==2rized 'Dwmmal Shodt C 03/05/89 Addrs== mad by IE4C Safety Evaluatimt dated i

3/5/87 (See also USI A-26 regardirty i =*==F=rit correspedence related to 4

ceneric Intter 89-n ans > fantory Giih 1.99 34evisicr12) 4 a

I i

i I

2 i

(

• C - Omplete NC - No Ctm.i- @ N _ry

NA - Not Applicable

. I - Inomplete l E - Evaluating Actions Required i

i l

I Attachment 2

' +-

. Supplemental Information USI A 2 Asymmetric Blowdown Loads on Reactor Primary Coolant Systems In letter LIC 80-0082 dated June 30, 1980, OPPD indicated the intention to utilize a method involving leak detection to identify a potential throughwall flaw before a double-ended failure would occur. In Generic Letter 84 04

" Safety Evaluation of Westinghouse Topical Reports Dealing with Elimination of Postulated Pipe Breaks in PWR Primary Main loops", NRC indicated that sufficient leak detection capability would be an acceptable response to this issue. According to Regulatory Guide 1.45, a leakage detection system capable of detecting a one gallon per minute (GPM) leak within four hours of initiation is acceptable. Existing equipment allows detection of such a leak at Fort Calhoun Station within four hours, with operator monitoring of Volume Control Tank inventory. OPPD has not, however, verified whether the associated alarm can be relied upon to identify a one gpm leak within four hours without operator monitoring. This has been identified as an open item in OPPD's design basis reconstitution project. A schedule for resolution of the open item will be provided by March 15, 1990.

USI A 11 Rg3ctor Vessel Material Touahness The central issue in this USI is to address the NRC staff concern that some vessels were projected to have belt line materials with Charpy upper shelf energy less than 50 ft - lb. Technical Specification Table 3-7 provides the capsule removal schedule for the Technical Specification 3.3 surveillance program to monitor radiation-induced changes in the mechanical and impact properties of the reactor vessel materials. The analysis of the most recently removed irradiated reactor vessel surveillance specimen (reference Omaha Public Power District, Fort Calhoun Station Unit No.1, Evaluation of Irradiated Capsule W 265, March 1984, transmitted to NRC via OPPD letter LIC-84-124 dated April 25, 1984) states "The average upper shelf impact energy after irradiation ranged from 59 ft - lb for the weld material to 109 ft - lb for the base metal (longitudinal)." It goes on to say "Using this conservative prediction, the upper shelf energy of the plates will remain above 82 ft - lbs during the i design life of the vessel and the weld shelf energy will remain above 50 ft -

I lb." OPPD therefore concludes USI A-ll has been satisfied.

USI A-17 Systems Interactions Th s USI was resolved by the NRC with the issuance of Generic Letter 89-18

" Systems Interactions In Nuclear Power Plants". This Generic tetter contains no requirements, but its contents are being reviewed against current OPPD activities related to the Probabilistic Risk Assessment (PRA) being conductr.d by OPPD (reference NRC Generic Letter 88 20 dated November 23, 1988, and OPPD letter LIC-89-930 dated October 31,1989). Completion of OPPD's review of Generic Letter 89-18 is expected by March 8, 1990.

1

(. . .-

Attachment 2 (Continued)

(

USI A 26 Reactor Versel Pressure Transient Protection OPPD's resolution of this USI involves completion of actions planned in response to NRC Generic Letter 88 11 'NRC Position On Radiation Embrittlement s

of Reactor Vessel Materials and Its Impact on Plant Operations". OPPD responded to the Generic Letter with a letter dated October 28, 1988 (reference LIC-88-950). By letter dated March 24, 1989, NRC accepted OPPD's response to the Generic Letter and requested that OPPD provide a summary of the results of a low Temperature Overpressure Protecticn (LTOP) analysis to be conducted by Combustion Engineering for OPPD, including any OPPD evaluation of this analysis. The requested LTOP information will be provided by December 31, 1990.

l l

l l