ML19210A211
| ML19210A211 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 10/24/1975 |
| From: | Arnold R METROPOLITAN EDISON CO. |
| To: | |
| References | |
| 75-08, 75-8, GQL-1638, NUDOCS 7910240927 | |
| Download: ML19210A211 (4) | |
Text
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NRC C RIBUTION FOR PART 50 DOCKEl ATE RI AL (TEMPORARY FORM)
CONTROL NO: i nn, FILE: INCIDE Tr REPonT rn.
FROM: Metropolitan Edison Co.
DATE OF DOC DATE REC'D LTR TWX RPT OTHER Reading, Pa R.C. Arnold 10-24-75 10-2e-79 vvvy TO:
ORIG CC OTHER SENT AEC POR vvy J.P.0'Reilly 1-sign d l
SENT LOCAL PDR xxxx CLASS UNCLASS PROPINFO INPUT NO CYS REC'D DOCKET NO:
xxxxx 1
50-289 DESCRIPTION:
ENCLOSURES:
Ltr trans the following:
Nonroutine 30-Day Report #75-08 concerning Reactor Bui'iding temperatur.es in excess of tho~se expected during normal operation PLANT NAMh5ree Mile Island #1 FOR ACTION /INFORMATION 10-31-75 JGB BUTLER (L)
SCHWENCER (L) ZIEMANN (L)
REGAN (E)
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-j CLARK (L)
STOLZ (L)
DICKER (E)
LE AR (L)
W/ Cooies W/ Copies W/ Cooies W/ Copies PARR (L)
VASSALLO (L)
KNIGHTON (E)
SPELS W/ Copies W/ Copies W/ Copics W/ Copies KNIEL (L)
PURPLE (L)
YOUNGBLOOD (E)
W/ Copics W/ Ccpics W/ Copics W/ Copies INTERNAL DISTRIBUTION
% FIL D TECH REVIEW JENTON LIC ASST A/T IND NBU F0H pSCHROEDER yGRIMES R. DIGGS (L)
B R AITM AN f OGC, RCOM P-50GA #MACCARY G AMMILL H. GE ARIN (L)
SALTZMAN GOSSICK/ST AF F KNIGHT KASTN E R E. GOU LBOUR NE (L)
ME LTZ CASE PAWLICKl BALLARD P. KREUTZER (E) f GI AM BUSSO SHAO SPANG LE R J. LEE (L)
PLANS BOYD
MCDONALD MOORE (L)
- fHOUSTON ENVIRO S. REED (E)
CHAPMAN DEYC UNG (L)
,.NOVAK MULLER M. SERVICE (L)
DUBE (Ltr)
SKOVHOLT (L)
,.ROSS DICKER S. SHEPPARD (L)
E. COUPE GOLLER (L) (Ltr) lPPOLITO KNIGHTON M. SLATER (E)
PETERSON f
P. CO LLINS
/TEDESCO YOUNGBLOOD H. SMITH (L)
HARTFIELD (2)
DENISE J. CCLLINS REGAN S. TEETS (L)
KLECKER REG OPR LAIN AS PROJECT LDR G. WILLI AMS (E)
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BENAROYA V. WI LSON (L)
- WlGGINTON M /?E
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F. WILLI.e:S
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STEELE M. DUNCAN M
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roern nn noanss METROPOLITAN EDISON COMPANY DCST CFFICE sCX 542 REAclNG. PENNSYLVANI A 19603 TELEPHCNE 215 - 929-360:
October Eh, 1975 c'h.
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Director of Nuclear Reactor Regulation
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U.S. Nuclear Regulatcry Cen=ission Washington, D.C.
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' A(i T J. F. O'Reilly, Director U.S. Nuclear Regulatory Cc==ission, Regicn I 631 Fark Avenue King of Prussia, PA 19LC6
Dear Sirs:
N.;.
Eccket No. 50-289 Operating License No. DPR-50 Three Mile Island Nuclear Station, Unit 1 (SI-1)
Nonroutine 30-Oay Repcrt 75-08 In accordance with Section 6.7.2.b.2 cf the Technical Specifications for NI-l enclosed please find Ncnrcutine 30-Eay Repcrt 75-C8 vhich deals with Reactor Ruilding temperatures in excess of these expected during normal cperation.
We trust this submittal satisfies the repcrting requirements stated abcVe and any concerns ycu may have.
If you shculd have any questiens please contact me.
Sincerely, f
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R. C./ arnold W
7 CJ Vice Fresident G'
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Eescription of Variance During the surser =cnths Reacter Euilding (RE) internal temperatures in excess c
cf the 110 F sustained terr.erature assumed in the FSAR have been ext.erienced.
Average ter eratures ar,t.reachine 130"F have been experienced abcve elevation e
0 320 feet and 110 F-120"? below elevaticn 320 feet on da~s when wet bulb s
te=perature has been greater than approximately 75'F.
2.
Desirnation cf An_tarent Cause of Variance It has been determined that the RE heat 1 cad is nearly twice as high as the heat load as criginally predicted by E & W.
Althcugh the RE industrial ccclers are rated for twice the desien (E & W) heat lead they have not perferred up to their design rating.
A RE heat load of twice the design heat lead is suspected to be caused by possible chimney effects created by gaps in scme cystem insulation. Convection air ficw caused by the chimney effect could cause an increase in RE heat lead.
Tvc areas of concern with regard to these increased temperatures are the compenents within the EE that experience these temperatures and the EE structrual integrity.
With regard to RE structural integrity, the Reactor Euilding Containment Scell has been reanalyced fcr the additicnal thermal stress createi by the operating-c temperatures experienced (i.e. Il0*F to 120 F belcw the 320' elevation and 130*
F above elevation 320').
The Working Stress Design methed of ACI31c-c3 was used as specified in the FSAR. The calculated stresses were fcund to exceed the allowable stress values for the reinforcing steel permitted at the time that the RE was criginally designed. Ecwever for the identical reinforcing steel used for constructicn of the RE, Secticn II: Division 2 cf the ASME Eciler and Fressure Vessel Ccde (Jan. 1, 1975), permits a higher allcwable stress. Using this higher allcwable stress calculated stress values were within allevable limits.
In additicn, the centainment shell has been checked using the alternate design method of ACI316-63, the Ultimate Strength Design. Using the Ultimate Strength Design methed, the = cst critical secticn under the gcVerning lead cc=binaticn was determined to have a 1.63 facter cf safety.
The "Reacter Centainment Euilding First Tenden Surveillance Test One Year After S.I.T.." discussed the effect of the higher te=peratures on the tendens. As indicated in Section 51 cf this repcrt, "the effect cf a 1ccal high ambient temperature en the vertical tenicns is negligible. The hecp tendens would experience apprcximately a 3.1 percent greater icss of prestress force, which e.
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It should also te noted that during the times that RE temperatures rise abcVe 11I F the outside ambient temperature is well above the 21.hc? temperature used for RE design stress analysis. The design thermal stress therefore was based cn a 0
thermal stress represented by abcut a 00 differential temperature and the strese differential te=rerature for the high temperature conditions experienced at TMI-l is less than 70*F With regard to the ccnponents within uhe R3,a detaEbd analysis has been ccnducted and all safety related cc=penents anf ccatings have been determined to be ec=patible with the operating temperatures to which they are subjected.
Analyses have also been conducted to determine the effect, if any, that the increased operating temperature has en post LCCA contaianent conditions.
It was determined that none of the maximun post LCCA temperature er pressure conditicns specified in the FSA3 would be exceeded.
In su= nary the probability and consequences of an accident are not increased and the pcssibility of an accident not previcusly analyced has not been created by this increased RB temperature.
3.
Corre tive Actions In addition to the analyses mentioned abcVe, an investigation is in prcgress to locate any insulatien conditions that may be centributing to a RB heat load greater than design. Further, efforts have been and are being made to improve the industrial cccler performance.
Until such time as the abcVe actions are cc=pleted and the appropriate corrective acticns are taken the following actica vill be taken if RE average temperatures rise above 130"? abcVe 320' or 120 F belev 320',
R3 emergency ecoling will be initiated as necessary to reduce temperatures belcw these quoted abcve.
It shculd be noted that the conditions described abcve do not now exist in that outside ambient temperatures are well belcw 75 F vet bulb. Further, a FSAR change is being prepared to address these increased temperatures.
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