ML18227A424
| ML18227A424 | |
| Person / Time | |
|---|---|
| Site: | Turkey Point  |
| Issue date: | 05/02/1977 |
| From: | Schmidt A Florida Power & Light Co |
| To: | Moseley N Office of Nuclear Reactor Regulation |
| References | |
| PRN-LI-77-130 | |
| Download: ML18227A424 (8) | |
Text
U.S. NUCLEAR REGULATORY CO ION NRC FoRM 195 I2 76)
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. NRC DISTRIBUTION FOR PART 50 DOCKET MATERIAL OQCggENUMBER~
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FILE NUMBER INCIDENT REPORT TO:
Mr. Norman CD Moseley FROM:
Florida Power
& Light Company Miami, Florida A~ D, Schmidt DATE OF DOCUMENT 5/2/77 DATE RECEIVED 5/27/77 ETTE R SfoR IGINAL QCOP Y DESCRIPTION 0 NOTOR I2E 0 fgUNCI.ASS IF I E D PROP INPUT FORM ENCLOSURE NUMBER OF COPIES RECEIVED g/6'6 Licensee Event Report (RO 50-251/77-3).on 4/18/77 concerning discussions with their vendor showing that Tech Spec Fig 2'-1,
".Reactor Core Thermal
& HydraUlic Safety Limits", requires revisioni ~ ~ ~ ~ ~
PLANT NAME:
Tuikey Point Unit No.,4 RJL (1-P)
(2-P)
NOTE ~
IF PERSONNEL EXPOSURE 1S INVOLVED SEND DIRECTLY TO KREGER/J, COLLINS BRANCH CHIEF:
W 3 CYS FOR ACTION LIC~ ASST~
W CYS ACRS CYS Z NG S NT FOR ACTION/INFORMATION ppvrS G FIL HRC PDR ICE (2 MIPC SCHROEDER/IPPOLITO HOUSTON NOVAK/CHECK GRIMES INTERNALDISTRIBUTION BUTLER HANAUER TEDESCO/MACCARY EISENHUT BAER SIIAO VOLLMER/BUNCH KREGER/J COLLINS LPDR:
TIC EXTERNAL DISTRIBUTION C NTROL NUMBER NRC FORM 196 {2.76)
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FLORIDAPOWER & LIGHT COMPANY
$p+Rblff May 2, 1977 PRN-LI-77-130 Mr. Norman C. Moseley, Director, Region II Office of Inspection and Enforcement U.
S. Nuclear Regulatory Commission 230 Peachtree
- Street, N. W., Suite 1217 Atlanta, Georgia 30303 WAYSV~977 Q.~k 14INEAR RKGULATO><
lOII COhNI SS IMI~l A
Dear Mr. Moseley:
REPORTABLE OCCURRENCE 251-77-3 TURKEY POINT UNIT 4 DATE OF OCCURRENCE:
APRIL 18, 1977 NOMINAL POWER The attached Licensee Event: Report is being submitted in accordance with Technical Specification 6.9.2 to provide prompt notification of the subject occurrence.
Very truly yours,
~
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D. Schmidt Vice President Power Resources MAS/cpc Attachment cc:
Robert Lowenstein, Esquire Director, Office of Inspection and Enforcement (40)
Director, Office of Management Information and Program Control (3) 7714701g9 PEOPLE... SERVING PEOPLE
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LICENSEE EVENT REPORT COiVTROL BLOCK:
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0 0 14 15 LICENSE NUMBEA 0
0 0
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0 UCENSE TYPc EVENT TYPc 0
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~01 25 26 30 31 32 C'TEGQAY
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~ EVENT OATE REPORT OATE 0
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69 74 75 80 EVENT OESCRIPTION
~02 is sions with our NSSS vendor have shown that Technical S ecification Fi ure 2.1-1 as
[OD~]
"Reactor Core Thermal and H draulic Safet Limits" re uires revision.
.An analvsis 7
8 9 tOD4]
performed at 2200 Mwt has ielded a corrected Fi ure 2.1-1 a
licable to both Units 3 80 80 a9 80 Qg and 4.
No chan es to set 'pints o eratin limits or o eratin ractices are re uired 7
8 9 80 QO 6 as a result of this correction.
Neither unit has ever o crated nonconservativel as a
7 8 9 SYSTEM CAUSE COOK CQOE COMPONENT COOK tao>J ~zI z
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7 8 9 10 11 12 CAUSE OESCRIPTION Tisourewcauedbv PILe Ic COMPONENT SUFFUKA CCMFONKNT MApi4rACAJFKA Z
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9 17 43 44 47 a difference betw en t e values 80 VCLAnON N
48 of "nominal ower" a lied I 7
6 9
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'~IIC 80 v
i i s and 4 bv FPL a d the NSSS vendor.
The vendor erforms anal ses 60 wer of 2200 Mwt our icensed stead state reactor core thermal OTHER STATUS NA AMOUNT QF ACTIVITY NA P+QQ~LE CONSEQUENCES 7
89 FACIUTY STATUS
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9 10 12 13 FOAM QF ACTIVITY COATKNT AP~SEO OF RK'&SE HI lzJ hJ 7
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10 11 PERSONNEL EXPOSURES NUMBGEA TYPc OESCA~iN fg] ~oo o
~z 89 11 12 13 PERSONNEL INJUR'S NUMBEA
" CESCAIPTION
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7 89 11 12 MEiHQO OF OISCOVEAY 45'5 44 45 NA OISCOVERY OESCAIPTIQN NSSS Vendor Discussions LOCATION OF AELEASc NA 80 80 7
8 9 LOSS OR OAMAGE TO FACIUTV TYPc OKSCA,'PTI N
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89 10 PUSL:CITV 7
89 NA NA 80 AOOITIONAL FACTORS
[116]
See Pa e Tso for continuation. of Event Descri cion and Cause Description 7,8 9 80 89 NAME M. A. Scho man PHONE. 305/552-3802 80
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REPORTABLE OCCURRENCE 251-77-3 LICENSEE EVENT REPORT PAGE TWO Event Descri tion (continued) result of the minor discrepancy in Figure 2.1-1.
This was the first reportable occurrence at Turkey Point Unit 4 involving a nonconservative specification on reactor core thermal and hydraulic safety limits.
A similar report (250-77-5) will be submitted for Unit 3.
(251-77-3)
Cause Descri tion (continued) power output.
However, they recently advised us that the specific number being used for analytical work was 2192 Hwt.
An analysis performed at 2200 Mwt showed that, in a single case, it would be necessary to make a minor administrative change to a Technical Specification (Figure 2.1-1).
Figure 2.1-1 has been adjusted downward the necessary amount (approximately 1.5'F) to compensate for the correction.
A proposed Technical Specification amendment will be submitted to the NRC to request a formal change correcting Figure 2.1-1.
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