ML17219A334

Forwards Addl Reasons for Determination of NSHC for Std 1 of 10CFR50.92 Submitted in 861202 Application to Amend License NPF-16,revising Tech Specs Re Low Steam Generator Level Reactor Trip & Auxiliary Feedwater Actuating Sys Setpoints

ML17219A334
Person / Time
Site:	Saint Lucie
Issue date:	02/03/1987
From:	Woody C FLORIDA POWER & LIGHT CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
L-87-39, NUDOCS 8702090086
Download: ML17219A334 (6)

Text

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ACCESSION NBR: 8702090086 DOC. DATE: 87/02/03 NOTARIZED: NO DOCKET FACIL:.50-389 St. Lucie Plant~ Unit 2i Florida Potoer '5 Light Co. 05000389 AUTH. NANE AUTHOR AFFILIATION MOODY> C. O. Florida Power Zc Light Co.

REC IP. NAME RECIPIENT AFFILIATION Document Control Branch (Document Cbntrol Desk )

SUBJECT:

Forwards amp.lif ication of determination of NSHC for Std 1 o' 10CFR50. 92 submitted in 861202 application to amend License NPF-16'evising Tech Specs re low steam generator level reactor trip h auxiliary feedeater actuating sos setpoints.

DISTRlBUTION CODE: A001D COPIES RECEIVED: LTR ENCL SIZE:

TITLE: ORSuhmittal: General Distribution NOTES:

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ia P. o. 14000, JUNO BEACH, FL 33408-0420 I\ ~ pyhll/yi FEBRUARY 3 >987 L-87-39 U. S. Nuclear Regulatory Commission Document- Control Desk Washington, D. C. 20555 Gentlemen:

Pe: St. Lucie Unit 2, Docket No. 50-389 Proposed License Amendment Low Steam Generator Level Reactor Tri and AFAS Set pints By letter L-86-452, dated December 2, l986, Florida Power & Light Company (FPL) proposed to revise Technical Specification 2.2. I, Reactor Protection Instrumentation (RPS) and 3/4.3.2, Engineered Safety Features Actuation System (ESFAS) Instrumentation and the associated Bases of the St. Lucie Unit 2 Technical Specifications.

This proposed amendment was discussed with the NRC Staff in a conference call on, January l4, l987. During the conference call the Staff requested that FPL amplify the no significant hazards consideration for standard (I) of IO CFR 50.92 submitted in FPL's December 2, l 986 letter. Attached is FPL's amplification of the no significant hazards consideration for this standard.

During the same conference call, the Staff asked for the normal operating level band for steam generator level, and whether FPL intended to change the normal operating band following approval of the subject proposed amendment. FPL does not intend to change the normal, narrow range steam generator level operating band from the current value of 65% (narrow range).

If additional information is required about this submittal, please contact us.

Very truly yours, C. O. Woo Group Vi President Nuclear Energy COW/E JW/gp Attachment cc: Dr. J. Nelson Grace, Region II, USNRC USNRC Resident Inspector, St. Lucie Plant Mr. Alan Schubert, Florida Dept. of Health and Rehabilitative Services ook 2090086 870P0g PDR ADOCN 0g000389 PDR an FPL Group company E J W4/003/ I

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",1, It~'4 W A <<'f ttlll I P'i<<IP . 111 I',>>, 4 ~ P PP, I II I 4 P, C. V 1 ~ ~ ' 4 1 I'p ft 4 I 1 II p ~ 1 4 cll II 4 li 4 ~ lt, >> ( II ~" I -rt ~ ~ "..ll II 1 I ~ AI ','I ~ 4 4 I ~ 11.4<<4.4<<A>> ~ >> 4 ,44 <<>>I 'I II ~ 'f I I I f',-' 4 'll, IVI 1 14 ~ ~4 1 4 4 4 4 ~ I lI tt ~ 4 <<JC' 1 ~ II 4>> ~ ~, >>p>> =k ~ t(C, I 'f' ~ V 1 1 ' 4 1 I ~ ~ -1 ~ II 1 P ~, 1<< 44 ATTACHMENT Amplification of the Determination of No Significant Hazards Consideration, Standard (I) for Florida Power & Light Company, St. Lucie Unit 2 Proposed License Amendment Low Steam Generator Level Reactor Tri and AFAS Set pints (I) Operation of the facility in accordance with the proposed amendment would not involve a signficant increase in the probability or consequences of an accident previously evaluated. The proposed change lowers the possibility of an unplanned reactor trip or auxiliary feedwater actuation system initiation on low steam generator level, thus lowering the number of unnecessary challenges to reactor systems and instrumentation. The proposed amendment to the low steam generator level setpoints does not impact the results presented in the Reload Safety Evaluation for Cycles 2 and 3 but only reflects changes to the determination of the instrument setpoint consistent with the Cycles 2 and 3 Reactor Protection System (RPS) and Engineered Safety Features Actuation System (ESFAS) analysis assumptions. As a result, there is no increase in the possibility or consequences of an accident previously evaluated. E J W4/003/2 IIII ~ It 'I . 4't A 4 4. ~ ' Ul 4 all A 4, ~ ' II It 4 4 ~ I 4 QU ~ tt IU I tll 4 44 Jf lf -,f:4, . ILJ. lf ~ r JIIUIU II ,4 U I IJ 4&II f. !4 I I 4 ~f 4 I It Ill. ( ~ 4 t 4 II ~ 4 I I 4 U.II U 4 4 4