ML19262A407
| ML19262A407 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 12/31/1977 |
| From: | Arnold R METROPOLITAN EDISON CO. |
| To: | Reid R Office of Nuclear Reactor Regulation |
| References | |
| GQL-1792, NUDOCS 7910290488 | |
| Download: ML19262A407 (5) | |
Text
U.S. NUCLE AR REGULAT COM' 'SION COCKET NUMBER NRC F<sRu 19; fC -- Z 8 9 (2-?s t NRC DISTRIBUTION FOR PART 50 DOCKET MATERI AL TO:
FROM: yet 3, DATE OF COCUMENT R. W. Ee b Rea ing, Pennsylvania 19603 12-31-76 R. C. Arnola DATE RECEIVED 1-3-77 ILETTER O NOTORizE o PROP (NPUT FORM NUMBER OF COPIES RECE4VE D IORIGIN AL QlJ NC LASSIFI E D OC "
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DESCRIPTION ENCLOSURE Ltr Ref our 11-29-76 ltr. Trans The Preliminarf Sur:rarf Calculations of Following:
Pin Pressure & Tenrerature for TMI-1...With attache 4 tables 1 & 2...
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~XCKNOWLEDGED DONOT R20V3 rtANr NAME: ranzE xItE 1SI n D UNIT +
14R4 113 SAFETY
,4 ArTION/INFORMATION wNtrTun JCM l-7-77 3
ASSIGNED AD:
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BRANCH CHIEFr
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PROJECT MANAGER:
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SYSTEMS SAFETY PLANT SYSTEMS SITE SAFETY & ' ~ '
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/ GOSSICK & STAFF ENGINEERING IPPOLITO ENVIRO TECH.
MIPC MACARRY KIRKWOOD ERNST CASE KNIGHT BALLARD HANAUER SIHWEIL OPERATING REACTORS I SPANGLER HARLESS PAWLICKI STELLO SITE TECH.
PROJECT MANAGEMENT REACTOR SAFETY OPERATING TECH.
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Aw&E9 FCA /*(MAf33 METROPOLI FAN EDISON COMPANY POST OFFICE BOX 542 RE ADING, PENNSYLVAN! A 19603 TELEPHONE 215 - 929-0601 December 31, 1976 P,eguktorf Rchet File
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Directer of :Iuclear Reactor Regulation 6
3 Attn:
R. W. Reid, Chief k g' g
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Docket :Io. 50-289 Operating License :lo. DFR-50 Attached is cur preliminary su==ary of 3&W's response to the infcrmaticn requested in your inter, cf :Icvember 29, 1976. Although we have nct ccepleted cur reviev ve have identified several questions concerning this information, which we are currently discussing with B&W.
We anticipate that we vill Oc=plete cur review, satisfactor.17 resolve cur questicns, and forward to ycu any resultant =cdifications of sne attached by January 31, 1977.
Sincerely,
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R. C. Arncli Vice President fx.z
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It has been determined that Three Mile Island nuclear Station, Unit 1 (TMI-1) has already reached a burnup exposure of 20,000 megswatt-days per metric ten of uranium.
Calculations of pin pressure and temperature have been made for TMI-l as requested in ycur letter. These calculations were based cn utilizing the TAFY Ccde with and without the NEC FGR equation.
The TAFY analysis without the MEC FGR equatien was taken frc the Cecnee II Cycle I licensin6 analysis. The input parameters and TAFY URO restrictions are listed in Table 1.
This input is representative of current operating plants.
The input to the analysis using the URC FGR odel was identical, thus the differences in pin FGR =cdels.
The fuel temperatures remain the same because of the URC restriction that 3CL temperatures be used for accident analysis.
Since the internal fuel rod pressure does not exceed the nominal system pressure nc discussien is given for Operating with fuel cladding in tensicn.
The TAFY Ccde without the NRC FGR equation is the code used in current LCCA and safety analyses. Use of the NEC FGR equation in TAFY vill impact 1GCA analysis as the worst pin pressures vould nov occur at earlier burnups. Since initial inside and outside cladding surface oxide layers would be thinner at earlier burnups, the circalcy-vater (retal-water) reaction vould be larger than that previcusly calculated. The increased energy generation in the cladding vculd raise the peak cladding te=peratures and vculd probably result in the present LCCA limits violating the criteria of ICCFR50.h6. Requalification of LCCA limits vould be necessary, ultimately resulting in the issuance of revised Technical Specifications.
A survey of the non 1CCA related safety analysis accidents was performed and it was concluded that the NRC FGR =cdel would not affect the results. The average fuel te ieratures used in the transient cladding temperature calculaticns is not changed by the new URC FGR =cdel.
Further, the higher pin pressures at ECL vould not result in cladding rupture during these transients. Rence, the safety analysis accidents are not impacted by the NRC FGR =cdel.
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TA3LE 2 CCO:iEE 1, 2 & 3 ; TMI-1, A'iG-1 DATA Peak Rod Burnup TAFY B&W FGR Mcdel TAFY IRC FGR Model (MWD /MTU)
Pin Press.
Avg. Fuel Pin Press. Avg. Fuel (psi)
Temp.* at (psi)
Temp.
- at 17 KW/FT (cF) 17 EW/FT (cF) 20,000 1210 2990 1210 2990 22,000 1235 2990 12ho 2990 25,000 1295 2990 1320 2990 27,000 13ho 2990 1h10 2990 30,000 1h00 2990 1550 2990 32,000 1k30 2990 1615 2990 35,C00 1h70 2990 17h5 2990 37,000 1510 2990 1865 2990 38,000 1525 2990 1925 2990
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